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Sample records for consequence code system

Accident consequence assessment code development

International Nuclear Information System (INIS)

Homma, T.; Togawa, O.

1991-01-01

This paper describes the new computer code system, OSCAAR developed for off-site consequence assessment of a potential nuclear accident. OSCAAR consists of several modules which have modeling capabilities in atmospheric transport, foodchain transport, dosimetry, emergency response and radiological health effects. The major modules of the consequence assessment code are described, highlighting the validation and verification of the models. (author)
Study on the code system for the off-site consequences assessment of severe nuclear accident

Energy Technology Data Exchange (ETDEWEB)

Kim, Sora; Mn, Byung Il; Park, Ki Hyun; Yang, Byung Mo; Suh, Kyung Suk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-12-15

The importance of severe nuclear accidents and probabilistic safety assessment (PSA) were brought to international attention with the occurrence of severe nuclear accidents caused by the extreme natural disaster at Fukushima Daiichi nuclear power plant in Japan. In Korea, studies on level 3 PSA had made little progress until recently. The code systems of level 3 PSA, MACCS2 (MELCORE Accident Consequence Code System 2, US), COSYMA (COde SYstem from MAria, EU) and OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents, JAPAN), were reviewed in this study, and the disadvantages and limitations of MACCS2 were also analyzed. Experts from Korea and abroad pointed out that the limitations of MACCS2 include the following: MACCS2 cannot simulate multi-unit accidents/release from spent fuel pools, and its atmospheric dispersion is based on a simple Gaussian plume model. Some of these limitations have been improved in the updated versions of MACCS2. The absence of a marine and aquatic dispersion model and the limited simulating range of food-chain and economic models are also important aspects that need to be improved. This paper is expected to be utilized as basic research material for developing a Korean code system for assessing off-site consequences of severe nuclear accidents.
Study on the code system for the off-site consequences assessment of severe nuclear accident

International Nuclear Information System (INIS)

Kim, Sora; Mn, Byung Il; Park, Ki Hyun; Yang, Byung Mo; Suh, Kyung Suk

2016-01-01

The importance of severe nuclear accidents and probabilistic safety assessment (PSA) were brought to international attention with the occurrence of severe nuclear accidents caused by the extreme natural disaster at Fukushima Daiichi nuclear power plant in Japan. In Korea, studies on level 3 PSA had made little progress until recently. The code systems of level 3 PSA, MACCS2 (MELCORE Accident Consequence Code System 2, US), COSYMA (COde SYstem from MAria, EU) and OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents, JAPAN), were reviewed in this study, and the disadvantages and limitations of MACCS2 were also analyzed. Experts from Korea and abroad pointed out that the limitations of MACCS2 include the following: MACCS2 cannot simulate multi-unit accidents/release from spent fuel pools, and its atmospheric dispersion is based on a simple Gaussian plume model. Some of these limitations have been improved in the updated versions of MACCS2. The absence of a marine and aquatic dispersion model and the limited simulating range of food-chain and economic models are also important aspects that need to be improved. This paper is expected to be utilized as basic research material for developing a Korean code system for assessing off-site consequences of severe nuclear accidents
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
A parametric study of MELCOR Accident Consequence Code System 2 (MACCS2) Input Values for the Predicted Health Effect

International Nuclear Information System (INIS)

Kim, So Ra; Min, Byung Il; Park, Ki Hyun; Yang, Byung Mo; Suh, Kyung Suk

2016-01-01

The MELCOR Accident Consequence Code System 2, MACCS2, has been the most widely used through the world among the off-site consequence analysis codes. MACCS2 code is used to estimate the radionuclide concentrations, radiological doses, health effects, and economic consequences that could result from the hypothetical nuclear accidents. Most of the MACCS model parameter values are defined by the user and those input parameters can make a significant impact on the output. A limited parametric study was performed to identify the relative importance of the values of each input parameters in determining the predicted early and latent health effects in MACCS2. These results would not be applicable to every case of the nuclear accidents, because only the limited calculation was performed with Kori-specific data. The endpoints of the assessment were early- and latent cancer-risk in the exposed population, therefore it might produce the different results with the parametric studies for other endpoints, such as contamination level, absorbed dose, and economic cost. Accident consequence assessment is important for decision making to minimize the health effect from radiation exposure, accordingly the sufficient parametric studies are required for the various endpoints and input parameters in further research
MELCOR Accident Consequence Code System (MACCS)

International Nuclear Information System (INIS)

Rollstin, J.A.; Chanin, D.I.; Jow, H.N.

1990-02-01

This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projections, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management
MELCOR Accident Consequence Code System (MACCS)

Energy Technology Data Exchange (ETDEWEB)

Jow, H.N.; Sprung, J.L.; Ritchie, L.T. (Sandia National Labs., Albuquerque, NM (USA)); Rollstin, J.A. (GRAM, Inc., Albuquerque, NM (USA)); Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA))

1990-02-01

This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management. 59 refs., 14 figs., 15 tabs.
MELCOR Accident Consequence Code System (MACCS)

International Nuclear Information System (INIS)

Jow, H.N.; Sprung, J.L.; Ritchie, L.T.; Rollstin, J.A.; Chanin, D.I.

1990-02-01

This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management. 59 refs., 14 figs., 15 tabs
MELCOR Accident Consequence Code System (MACCS)

International Nuclear Information System (INIS)

Chanin, D.I.; Sprung, J.L.; Ritchie, L.T.; Jow, Hong-Nian

1990-02-01

This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previous CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. This document, Volume 1, the Users's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems
MELCOR Accident Consequence Code System (MACCS)

Energy Technology Data Exchange (ETDEWEB)

Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA)); Sprung, J.L.; Ritchie, L.T.; Jow, Hong-Nian (Sandia National Labs., Albuquerque, NM (USA))

1990-02-01

This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previous CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. This document, Volume 1, the Users's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems.
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
Evaluation of severe accident risks: Quantification of major input parameters: MAACS [MELCOR Accident Consequence Code System] input

International Nuclear Information System (INIS)

Sprung, J.L.; Jow, H-N; Rollstin, J.A.; Helton, J.C.

1990-12-01

Estimation of offsite accident consequences is the customary final step in a probabilistic assessment of the risks of severe nuclear reactor accidents. Recently, the Nuclear Regulatory Commission reassessed the risks of severe accidents at five US power reactors (NUREG-1150). Offsite accident consequences for NUREG-1150 source terms were estimated using the MELCOR Accident Consequence Code System (MACCS). Before these calculations were performed, most MACCS input parameters were reviewed, and for each parameter reviewed, a best-estimate value was recommended. This report presents the results of these reviews. Specifically, recommended values and the basis for their selection are presented for MACCS atmospheric and biospheric transport, emergency response, food pathway, and economic input parameters. Dose conversion factors and health effect parameters are not reviewed in this report. 134 refs., 15 figs., 110 tabs
Health effects estimation code development for accident consequence analysis

International Nuclear Information System (INIS)

Togawa, O.; Homma, T.

1992-01-01

As part of a computer code system for nuclear reactor accident consequence analysis, two computer codes have been developed for estimating health effects expected to occur following an accident. Health effects models used in the codes are based on the models of NUREG/CR-4214 and are revised for the Japanese population on the basis of the data from the reassessment of the radiation dosimetry and information derived from epidemiological studies on atomic bomb survivors of Hiroshima and Nagasaki. The health effects models include early and continuing effects, late somatic effects and genetic effects. The values of some model parameters are revised for early mortality. The models are modified for predicting late somatic effects such as leukemia and various kinds of cancers. The models for genetic effects are the same as those of NUREG. In order to test the performance of one of these codes, it is applied to the U.S. and Japanese populations. This paper provides descriptions of health effects models used in the two codes and gives comparisons of the mortality risks from each type of cancer for the two populations. (author)
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
PROBCON-HDW: A probability and consequence system of codes for long-term analysis of Hanford defense wastes

International Nuclear Information System (INIS)

Piepho, M.G.; Nguyen, T.H.

1988-12-01

The PROBCON-HDW (PROBability and CONsequence analysis for Hanford defense waste) computer code system calculates the long-term cumulative releases of radionuclides from the Hanford defense wastes (HDW) to the accessible environment and compares the releases to environmental release limits as defined in 40 CFR 191. PROBCON-HDW takes into account the variability of input parameter values used in models to calculate HDW release and transport in the vadose zone to the accessible environment (taken here as groundwater). A human intrusion scenario, which consists of drilling boreholes into the waste beneath the waste sites and bringing waste to the surface, is also included in PROBCON-HDW. PROBCON-HDW also includes the capability to combine the cumulative releases according to various long-term (10,000 year) scenarios into a composite risk curve or complementary cumulative distribution function (CCDF). The system structure of the PROBCON-HDW codes, the mathematical models in PROBCON-HDW, the input files, the input formats, the command files, and the graphical output results of several HDW release scenarios are described in the report. 3 refs., 7 figs., 9 tabs
Methods and codes for assessing the off-site Consequences of nuclear accidents. Volume 2

International Nuclear Information System (INIS)

Kelly, G.N.; Luykx, F.

1991-01-01

The Commission of the European Communities, within the framework of its 1980-84 radiation protection research programme, initiated a two-year project in 1983 entitled methods for assessing the radiological impact of accidents (Maria). This project was continued in a substantially enlarged form within the 1985-89 research programme. The main objectives of the project were, firstly, to develop a new probabilistic accident consequence code that was modular, incorporated the best features of those codes already in use, could be readily modified to take account of new data and model developments and would be broadly applicable within the EC; secondly, to acquire a better understanding of the limitations of current models and to develop more rigorous approaches where necessary; and, thirdly, to quantify the uncertainties associated with the model predictions. This research led to the development of the accident consequence code Cosyma (COde System from MAria), which will be made generally available later in 1990. The numerous and diverse studies that have been undertaken in support of this development are summarized in this paper, together with indications of where further effort might be most profitably directed. Consideration is also given to related research directed towards the development of real-time decision support systems for use in off-site emergency management
Confidence level in the calculations of HCDA consequences using large codes

International Nuclear Information System (INIS)

Nguyen, D.H.; Wilburn, N.P.

1979-01-01

The probabilistic approach to nuclear reactor safety is playing an increasingly significant role. For the liquid-metal fast breeder reactor (LMFBR) in particular, the ultimate application of this approach could be to determine the probability of achieving the goal of a specific line-of-assurance (LOA). Meanwhile a more pressing problem is one of quantifying the uncertainty in a calculated consequence for hypothetical core disruptive accident (HCDA) using large codes. Such uncertainty arises from imperfect modeling of phenomenology and/or from inaccuracy in input data. A method is presented to determine the confidence level in consequences calculated by a large computer code due to the known uncertainties in input invariables. A particular application was made to the initial time of pin failure in a transient overpower HCDA calculated by the code MELT-IIIA in order to demonstrate the method. A probability distribution function (pdf) for the time of failure was first constructed, then the confidence level for predicting this failure parameter within a desired range was determined
Experience with COSYMA in an international intercomparison of probabilistic accident consequence assessment codes

International Nuclear Information System (INIS)

Hasemann, I.; Jones, J.A.; Steen, J. van der; Wonderen, E. van

1996-01-01

The Commission of the European Communities and the Nuclear Energy Agency of the OECD have organized an international exercise to compare the predictions of accident consequence assessment codes, and to identify those features of the models which lead to differences in the predicted results. Alongside this, a further exercise was undertaken in which the COSYMA code was used independently by several different organizations. Some of the findings of the COSYMA users' exercise are described that have general applications to accident consequence assessments. A number of areas are identified in which further work on accident consequence models may be justified. These areas, which are also of interest for codes other than COSYMA, are (a) the calculation and averaging of doses and risks to people sheltered in different types of buildings, particularly with respect to the evaluation of early health effects; (b) the modeling of long-duration releases and their description as a series of shorter releases; (c) meteorological sampling for results at a certain location, specifically for use with trajectory models of atmospheric dispersion; and (d) aspects of calculating probabilities of consequences at a point
Parameterization of the driving time in the evacuation or fast relocation model of an accident consequence code

International Nuclear Information System (INIS)

Pfeffer, W.; Hofer, E.; Nowak, E.; Schnadt, H.

1988-01-01

The model of protective measures in the accident consequence code system UFOMOD of the German Risk Study, Phase B, requires the driving times of the population to be evacuated for the evaluation of the dose received during the evacuation. The parameter values are derived from evacuation simulations carried out with the code EVAS for 36 sectors from various sites. The simulations indicated that the driving time strongly depends on the population density, whereas other influences are less important. It was decided to use different driving times in the consequence code for each of four population density classes as well as for each of three or four fractions of the population in a sector. The variability between sectors of a class was estimated from the 36 sectors, in order to derive subjective probability distributions that are to model the uncertainty in the parameter value to be used for any of the fractions in a particular sector for which an EVAS simulation has not yet been performed. To this end also the impact of the uncertainties in the parameters and modelling assumptions of EVAS on the simulated times was quantified using expert judgement. The distributions permit the derivation of a set of driving times to be used as so-called ''best estimate'' or reference values in the accident consequence code. Additionally they are directly applicable in an uncertainty and sensitivity analysis
Review of the chronic exposure pathways models in MACCS [MELCOR Accident Consequence Code System] and several other well-known probabilistic risk assessment models

International Nuclear Information System (INIS)

Tveten, U.

1990-06-01

The purpose of this report is to document the results of the work performed by the author in connection with the following task, performed for US Nuclear Regulatory Commission, (USNRC) Office of Nuclear Regulatory Research, Division of Systems Research: MACCS Chronic Exposure Pathway Models: Review the chronic exposure pathway models implemented in the MELCOR Accident Consequence Code System (MACCS) and compare those models to the chronic exposure pathway models implemented in similar codes developed in countries that are members of the OECD. The chronic exposures concerned are via: the terrestrial food pathways, the water pathways, the long-term groundshine pathway, and the inhalation of resuspended radionuclides pathway. The USNRC has indicated during discussions of the task that the major effort should be spent on the terrestrial food pathways. There is one chapter for each of the categories of chronic exposure pathways listed above

Evaluation system for the analysis of the radiological consequences (RASCAL)

International Nuclear Information System (INIS)

Miguel Martinez, M. J. de

2011-01-01

The code currently employed by the Spanish Nuclear power is the IRDAM (Interactive Rapid Dose Assessment Model). This code will now be replaced by the RASCAL (Radiological Assessment System for Consequence Analysis) version 4.0. This is a significant improvement in dose calculation for escapes to the atmosphere for radioactive emergencies. The main objective is to highlight the most significant changes introduced in the RASCAL when performing these dose estimates.
Proceedings of the Seminar on Methods and Codes for Assessing the off-site consequences of nuclear accidents. Volume 1

International Nuclear Information System (INIS)

Kelly, G.N.; Luykx, F.

1991-01-01

The Commission of the European Communities, within the framework of its 1980-84 radiation protection research programme, initiated a two-year project in 1983 entitled 'methods for assessing the radiological impact of accidents' (Maria). This project was continued in a substantially enlarged form within the 1985-89 research programme. The main objectives of the project were, firstly, to develop a new probabilistic accident consequence code that was modular, incorporated the best features of those codes already in use, could be readily modified to take account of new data and model developments and would be broadly applicable within the EC; secondly, to acquire a better understanding of the limitations of current models and to develop more rigorous approaches where necessary; and, thirdly, to quantify the uncertainties associated with the model predictions. This research led to the development of the accident consequence code Cosyma (COde System from MAria), which will be made generally available later in 1990. The numerous and diverse studies that have been undertaken in support of this development are summarized in this paper, together with indications of where further effort might be most profitably directed. Consideration is also given to related research directed towards the development of real-time decision support systems for use in off-site emergency management
[Quality management and strategic consequences of assessing documentation and coding under the German Diagnostic Related Groups system].

Science.gov (United States)

Schnabel, M; Mann, D; Efe, T; Schrappe, M; V Garrel, T; Gotzen, L; Schaeg, M

2004-10-01

The introduction of the German Diagnostic Related Groups (D-DRG) system requires redesigning administrative patient management strategies. Wrong coding leads to inaccurate grouping and endangers the reimbursement of treatment costs. This situation emphasizes the roles of documentation and coding as factors of economical success. The aims of this study were to assess the quantity and quality of initial documentation and coding (ICD-10 and OPS-301) and find operative strategies to improve efficiency and strategic means to ensure optimal documentation and coding quality. In a prospective study, documentation and coding quality were evaluated in a standardized way by weekly assessment. Clinical data from 1385 inpatients were processed for initial correctness and quality of documentation and coding. Principal diagnoses were found to be accurate in 82.7% of cases, inexact in 7.1%, and wrong in 10.1%. Effects on financial returns occurred in 16%. Based on these findings, an optimized, interdisciplinary, and multiprofessional workflow on medical documentation, coding, and data control was developed. Workflow incorporating regular assessment of documentation and coding quality is required by the DRG system to ensure efficient accounting of hospital services. Interdisciplinary and multiprofessional cooperation is recognized to be an important factor in establishing an efficient workflow in medical documentation and coding.
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.
Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide

International Nuclear Information System (INIS)

Ritchie, L.T.; Johnson, J.D.; Blond, R.M.

1983-02-01

The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems
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)
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
EDPUFF- a Gaussian dispersion code for consequence analysis

International Nuclear Information System (INIS)

Oza, R.B.; Bapat, V.N.; Nair, R.N.; Hukkoo, R.K.; Krishnamoorthy, T.M.

1995-01-01

EDPUFF- Equi Distance Puff is a Gaussian dispersion code in FORTRAN language to model atmospheric dispersion of instantaneous or continuous point source releases. It is designed to incorporate the effect of changing meteorological conditions and source release rates on the spatial distribution profiles and its consequences. Effects of variation of parameters like puff spacing, puff packing, averaging schemes are discussed and the choice of the best values for minimum errors and minimum computer CPU time are identified. The code calculates the doses to individual receptors as well as average doses for population zones from internal and external routes over the area of interest. Internal dose computations are made for inhalation and ingestion pathways while the doses from external route consists of cloud doses and doses from surface deposited activity. It computes inhalation and ingestion dose (milk route only) for critical group (1 yr old child). In case of population zones it finds out maximum possible doses in a given area along with the average doses discussed above. Report gives the doses from various pathways for unit release of fixed duration. (author). 7 refs., figs., 7 appendixes
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
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)
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
A review of the Melcor Accident Consequence Code System (MACCS): Capabilities and applications

International Nuclear Information System (INIS)

Young, M.

1995-01-01

MACCS was developed at Sandia National Laboratories (SNL) under U.S. Nuclear Regulatory Commission (NRC) sponsorship to estimate the offsite consequences of potential severe accidents at nuclear power plants (NPPs). MACCS was publicly released in 1990. MACCS was developed to support the NRC's probabilistic safety assessment (PSA) efforts. PSA techniques can provide a measure of the risk of reactor operation. PSAs are generally divided into three levels. Level one efforts identify potential plant damage states that lead to core damage and the associated probabilities, level two models damage progression and containment strength for establishing fission-product release categories, and level three efforts evaluate potential off-site consequences of radiological releases and the probabilities associated with the consequences. MACCS was designed as a tool for level three PSA analysis. MACCS performs probabilistic health and economic consequence assessments of hypothetical accidental releases of radioactive material from NPPs. MACCS includes models for atmospheric dispersion and transport, wet and dry deposition, the probabilistic treatment of meteorology, environmental transfer, countermeasure strategies, dosimetry, health effects, and economic impacts. The computer systems MACCS is designed to run on are the 386/486 PC, VAX/VMS, E3M RISC S/6000, Sun SPARC, and Cray UNICOS. This paper provides an overview of MACCS, reviews some of the applications of MACCS, international collaborations which have involved MACCS, current developmental efforts, and future directions
Evaluation of radiological dispersion/consequence codes supporting DOE nuclear facility SARs

International Nuclear Information System (INIS)

O'Kula, K.R.; Paik, I.K.; Chung, D.Y.

1996-01-01

Since the early 1990s, the authorization basis documentation of many U.S. Department of Energy (DOE) nuclear facilities has been upgraded to comply with DOE orders and standards. In this process, many safety analyses have been revised. Unfortunately, there has been nonuniform application of software, and the most appropriate computer and engineering methodologies often are not applied. A DOE Accident Phenomenology and Consequence (APAC) Methodology Evaluation Program was originated at the request of DOE Defense Programs to evaluate the safety analysis methodologies used in nuclear facility authorization basis documentation and to define future cost-effective support and development initiatives. Six areas, including source term development (fire, spills, and explosion analysis), in-facility transport, and dispersion/ consequence analysis (chemical and radiological) are contained in the APAC program. The evaluation process, codes considered, key results, and recommendations for future model and software development of the Radiological Dispersion/Consequence Working Group are summarized in this paper
RASCAL 4.l: evaluation system for the analysis of the radiological consequences

International Nuclear Information System (INIS)

Miguel, I. de; Gomez-Arguello, B.

2012-01-01

Recently the Spanish Nuclear Safety Council (CSN) has promoted the replacement of the code used in the Spanish Nuclear Power Plants to estimate potential and actual doses due to atmospheric releases. IRDAM Code (interactive Rapid Dose Assessment Model, 1983) by RASCAL Code version 4.1 (Radiological Assessment System for consequence Analysis). This code was developed by the NRC (Nuclear Regulatory Commission) to evaluate releases from nuclear power plants, spent fuel storage pools and casks, fuel cycle facilities and radioactive material handling facilities. RASCAL was officially requested to the Spanish nuclear power plants in February 1st, 2012, and it can be used for external dose assessment and to evaluate the recommended protective action for the public in a specific situation (sheltering, prophylaxis, evacuation, etc.). (Author)
The Therapy Process Observational Coding System for Child Psychotherapy Strategies Scale

Science.gov (United States)

McLeod, Bryce D.; Weisz, John R.

2010-01-01

Most everyday child and adolescent psychotherapy does not follow manuals that document the procedures. Consequently, usual clinical care has remained poorly understood and rarely studied. The Therapy Process Observational Coding System for Child Psychotherapy-Strategies scale (TPOCS-S) is an observational measure of youth psychotherapy procedures…
The integrated code system CASCADE-3D for advanced core design and safety analysis

International Nuclear Information System (INIS)

Neufert, A.; Van de Velde, A.

1999-01-01

The new program system CASCADE-3D (Core Analysis and Safety Codes for Advanced Design Evaluation) links some of Siemens advanced code packages for in-core fuel management and accident analysis: SAV95, PANBOX/COBRA and RELAP5. Consequently by using CASCADE-3D the potential of modern fuel assemblies and in-core fuel management strategies can be much better utilized because safety margins which had been reduced due to conservative methods are now predicted more accurately. By this innovative code system the customers can now take full advantage of the recent progress in fuel assembly design and in-core fuel management.(author)
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
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
Development of Integrated Code for Risk Assessment (INCORIA) for Physical Protection System

International Nuclear Information System (INIS)

Jang, Sung Soon; Seo, Hyung Min; Yoo, Ho Sik

2010-01-01

A physical protection system (PPS) integrates people, procedures and equipment for the protection of assets or facilities against theft, sabotage or other malevolent human attacks. Among critical facilities, nuclear facilities and nuclear weapon sites require the highest level of PPS. After the September 11, 2001 terrorist attacks, international communities, including the IAEA, have made substantial efforts to protect nuclear material and nuclear facilities. The international flow on nuclear security is using the concept or risk assessment. The concept of risk assessment is firstly devised by nuclear safety people. They considered nuclear safety including its possible risk, which is the frequency of failure and possible consequence. Nuclear security people also considers security risk, which is the frequency of threat action, vulnerability, and consequences. The concept means that we should protect more when the credible threat exists and the possible radiological consequence is high. Even if there are several risk assessment methods of nuclear security, the application needs the help of tools because of a lot of calculation. It's also hard to find tools for whole phase of risk assessment. Several codes exist for the part of risk assessment. SAVI are used for vulnerability of PPS. Vital area identification code is used for consequence analysis. We are developing Integrated Code for Risk Assessment (INCORIA) to apply risk assessment methods for nuclear facilities. INCORIA evaluates PP-KINAC measures and generation tools for threat scenario. PP-KINAC is risk assessment measures for physical protection system developed by Hosik Yoo and is easy to apply. A threat scenario tool is used to generate threat scenario, which is used as one of input value to PP-KINAC measures
Analyses of PWR boron dilution consequences with the Arrotta code

International Nuclear Information System (INIS)

Johanson, E.; Cheng, H.W.; Sehgal, B.R.

1998-03-01

During the past few years, major attention has been paid to analyzing the issue of reactivity initiated accidents (RIAs), of which the boron dilution event is of very special interest to the countries having pressurized water reactors (PWRs) in their nuclear power delivery systems. The scenario considered is that if an inadvertent accumulation of boron free water in one loop during reactor startup operations of a PWR and the inadvertent startup of the reactor coolant pump (RCP) in the loop. This could then lead to a rapid boron dilution in the core, which can in turn give rise to a power excursion. This report is devoted to studying the potential physical and thermal hydraulic consequences of a slug of diluted coolant entering the core after one RCP start under a couple of postulated cases. The severity of the consequences of such a scenario is primarily determined by the amount of positive reactivity insertion, and they are also related to the reactivity insertion rate. Therefore, in the report, detailed calculations and analyses have been carried out from case to case by using the well-known space-time kinetics code, ARROTTA. As a result, the spatial distribution for nodal power, fuel enthalpy, fuel temperature and clad outside temperature as well as the change in core reactivity, total core power and peak fuel temperature can be provided. In general, the maximum fuel enthalpy, peak fuel temperature, and clad outside temperature, for all the cases considered in the report, do not exceed their respective routine safety limitations because of the strong Doppler effect and moderator temperature feedback, except if the safety limitations on fuel enthalpy addition for high burnup fuel are drastically reduced

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.
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
Development of sub-channel/system coupled code and its application to a supercritical water-cooled test loop

International Nuclear Information System (INIS)

Liu, X.J.; Yang, T.; Cheng, X.

2014-01-01

To analyze the local thermal-hydraulic parameters in the supercritical water reactor-fuel qualification test (SCWR-FQT) fuel bundle with a flow blockage, a coupled sub-channel and system code system is developed in this paper. Both of the sub-channel code and system code are adapted to transient analysis of SCWR. Two codes are coupled by data transfer and data adaptation at the interface. In the 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. Sensitivity analysis are carried out respectively in ATHLET-SC and COBRA-SC code, to identify the appropriate models for description of the flow blockage phenomenon in the test loop. Some measures to mitigate the accident consequence are also trialed to demonstrate their effectiveness. 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 assembly can be reduced effectively by the safety measures of SCWR-FQT. (author)
The applicability of ALPHA/PHOENIX/ANC nuclear design code system on Korean standard PWR's

International Nuclear Information System (INIS)

Lee, Kookjong; Choi, Kie-Yong; Lee, Hae-Chan; Roh, Eun-Rae

1996-01-01

For the Korean Standard Nuclear Power Plant (KSNPP) designed based on Combustion Engineering (CE) System 80, the Westinghouse nuclear design code system ALPHA/PHOENIX/ANC was applied to the follow-up design of initial and reload core of KSNPP. The follow-up design results of Yonggwang Unit 3 Cycle 1, 2 and Yonggwang Unit 4 Cycle 1 have shown good agreements with the measured data. The assemblywise power distributions have shown less than 2% average differences and critical boron concentrations have shown less than 20 ppm differences. All the low power physics test parameters are in good agreement. Consequently, APA design code system can be applied to KNSPP cores. (author)
Athermalization of infrared dual field optical system based on wavefront coding

Science.gov (United States)

Jiang, Kai; Jiang, Bo; Liu, Kai; Yan, Peipei; Duan, Jing; Shan, Qiu-sha

2017-02-01

Wavefront coding is a technology which combination of the optical design and digital image processing. By inserting a phase mask closed to the pupil plane of the optical system the wavefront of the system is re-modulated. And the depth of focus is extended consequently. In reality the idea is same as the athermalization theory of infrared optical system. In this paper, an uncooled infrared dual field optical system with effective focal as 38mm/19mm, F number as 1.2 of both focal length, operating wavelength varying from 8μm to 12μm was designed. A cubic phase mask was used at the pupil plane to re-modulate the wavefront. Then the performance of the infrared system was simulated with CODEV as the environment temperature varying from -40° to 60°. MTF curve of the optical system with phase mask are compared with the outcome before using phase mask. The result show that wavefront coding technology can make the system not sensitive to thermal defocus, and then realize the athermal design of the infrared optical system.
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
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
Support of Maritime Education and Training Systems for the Implementation of the ISM Code

Directory of Open Access Journals (Sweden)

Elena Maggi

2012-10-01

Full Text Available The problem of the safety improvement and pollution preventionin maritime tra11Sp011 has become more and more criticaland urgent to solve. In fact, the number of accidents at seahas increased very quickly over time. Concern is growing aboutpoor qualification on safety and poor management standardsin shipping industry. The ISM Code (Intenzational Safety ManagementCode aims to provide intemational standards for safemanagement of ship operations and for pollution prevention.The paper, which presents a part of the work done by the Universityof Trieste - ISTIEE within the framework of theMETHAR project, aims to identify the expected supp01t fromthe MaJitime Education and Training (MET systems to implemmttheISM Code, consequently improving safety and preventingpollution. The paper desCiibes first the 01igin, the objectivesof the ISM Code and the standard requirements on METidentified by the Code. Secondly, it summarises the opinions ofthe operators collected through questionnaires. Finally, it identifiesthe possible enrichment of the MET system in order tobel/.er optimise the implementation of the Code.
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.
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
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
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.
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...
The program system UFOMOD for assessing the consequences of nuclear accidents

International Nuclear Information System (INIS)

Ehrhardt, J.; Burkart, K.; Hasemann, I.; Matzerath, C.; Panitz, H.J.; Steinhauer, C.

1988-10-01

The programm system UFOMOD is a completely new accident consequence assessment (ACA) code. Its structure and modelling is based on the experience gained from applications of the old UFOMOD code during and after the German Risk Study - Phase A, the results of scientific investigations performed within the ongoing Phase B and the CEC-project MARIA, and the requirements resulting from the extended use of ACAs to help in decision-making. One of the most important improvements is the introduction of different trajecotry models for describing atmospheric dispersion in the near range and at larger distances. Emergency actions and countermeasures modelling takes into account recommendations of international commissions. The dosimetric models contain completely new age-, sex- and time-dependent data of dose-conversion factors for external and internal radiation; the ingestion pathway is modelled to consider seasonal dependencies. New dose-risk-relationships for stochastic and non-stochastic health effects are implemented; a special algorithm developed for ACA codes allows individual and collective leukemia and cancer risks to be presented as a function of time after the accident. According to the modular structure of the new program system UFOMOD, an easy access to parameter values and the results of the various submodels exists what facilitates sensitivity and uncertainty analyses. (orig.) [de
Study on a new meteorological sampling scheme developed for the OSCAAR code system

International Nuclear Information System (INIS)

Liu Xinhe; Tomita, Kenichi; Homma, Toshimitsu

2002-03-01

One important step in Level-3 Probabilistic Safety Assessment is meteorological sequence sampling, on which the previous studies were mainly related to code systems using the straight-line plume model and more efforts are needed for those using the trajectory puff model such as the OSCAAR code system. This report describes the development of a new meteorological sampling scheme for the OSCAAR code system that explicitly considers population distribution. A group of principles set for the development of this new sampling scheme includes completeness, appropriate stratification, optimum allocation, practicability and so on. In this report, discussions are made about the procedures of the new sampling scheme and its application. The calculation results illustrate that although it is quite difficult to optimize stratification of meteorological sequences based on a few environmental parameters the new scheme do gather the most inverse conditions in a single subset of meteorological sequences. The size of this subset may be as small as a few dozens, so that the tail of a complementary cumulative distribution function is possible to remain relatively static in different trials of the probabilistic consequence assessment code. (author)
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)
Delay Estimation in Long-Code Asynchronous DS/CDMA Systems Using Multiple Antennas

Directory of Open Access Journals (Sweden)

Sirbu Marius

2004-01-01

Full Text Available The problem of propagation delay estimation in asynchronous long-code DS-CDMA multiuser systems is addressed. Almost all the methods proposed so far in the literature for propagation delay estimation are derived for short codes and the knowledge of the codes is exploited by the estimators. In long-code CDMA, the spreading code is aperiodic and the methods developed for short codes may not be used or may increase the complexity significantly. For example, in the subspace-based estimators, the aperiodic nature of the code may require subspace tracking. In this paper we propose a novel method for simultaneous estimation of the propagation delays of several active users. A specific multiple-input multiple-output (MIMO system model is constructed in a multiuser scenario. In such model the channel matrix contains information about both the users propagation delays and channel impulse responses. Consequently, estimates of the delays are obtained as a by-product of the channel estimation task. The channel matrix has a special structure that is exploited in estimating the delays. The proposed delay estimation method lends itself to an adaptive implementation. Thus, it may be applied to joint channel and delay estimation in uplink DS-CDMA analogously to the method presented by the authors in 2003. The performance of the proposed method is studied in simulation using realistic time-varying channel model and different SNR levels in the face of near-far effects, and using low spreading factor (high data rates.
Performance and Complexity Evaluation of Iterative Receiver for Coded MIMO-OFDM Systems

Directory of Open Access Journals (Sweden)

Rida El Chall

2016-01-01

Full Text Available Multiple-input multiple-output (MIMO technology in combination with channel coding technique is a promising solution for reliable high data rate transmission in future wireless communication systems. However, these technologies pose significant challenges for the design of an iterative receiver. In this paper, an efficient receiver combining soft-input soft-output (SISO detection based on low-complexity K-Best (LC-K-Best decoder with various forward error correction codes, namely, LTE turbo decoder and LDPC decoder, is investigated. We first investigate the convergence behaviors of the iterative MIMO receivers to determine the required inner and outer iterations. Consequently, the performance of LC-K-Best based receiver is evaluated in various LTE channel environments and compared with other MIMO detection schemes. Moreover, the computational complexity of the iterative receiver with different channel coding techniques is evaluated and compared with different modulation orders and coding rates. Simulation results show that LC-K-Best based receiver achieves satisfactory performance-complexity trade-offs.
EquiFACS: The Equine Facial Action Coding System.

Directory of Open Access Journals (Sweden)

Jen Wathan

Full Text Available Although previous studies of horses have investigated their facial expressions in specific contexts, e.g. pain, until now there has been no methodology available that documents all the possible facial movements of the horse and provides a way to record all potential facial configurations. This is essential for an objective description of horse facial expressions across a range of contexts that reflect different emotional states. Facial Action Coding Systems (FACS provide a systematic methodology of identifying and coding facial expressions on the basis of underlying facial musculature and muscle movement. FACS are anatomically based and document all possible facial movements rather than a configuration of movements associated with a particular situation. Consequently, FACS can be applied as a tool for a wide range of research questions. We developed FACS for the domestic horse (Equus caballus through anatomical investigation of the underlying musculature and subsequent analysis of naturally occurring behaviour captured on high quality video. Discrete facial movements were identified and described in terms of the underlying muscle contractions, in correspondence with previous FACS systems. The reliability of others to be able to learn this system (EquiFACS and consistently code behavioural sequences was high--and this included people with no previous experience of horses. A wide range of facial movements were identified, including many that are also seen in primates and other domestic animals (dogs and cats. EquiFACS provides a method that can now be used to document the facial movements associated with different social contexts and thus to address questions relevant to understanding social cognition and comparative psychology, as well as informing current veterinary and animal welfare practices.
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.

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
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.
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)
New quantum codes derived from a family of antiprimitive BCH codes

Science.gov (United States)

Liu, Yang; Li, Ruihu; Lü, Liangdong; Guo, Luobin

The Bose-Chaudhuri-Hocquenghem (BCH) codes have been studied for more than 57 years and have found wide application in classical communication system and quantum information theory. In this paper, we study the construction of quantum codes from a family of q2-ary BCH codes with length n=q2m+1 (also called antiprimitive BCH codes in the literature), where q≥4 is a power of 2 and m≥2. By a detailed analysis of some useful properties about q2-ary cyclotomic cosets modulo n, Hermitian dual-containing conditions for a family of non-narrow-sense antiprimitive BCH codes are presented, which are similar to those of q2-ary primitive BCH codes. Consequently, via Hermitian Construction, a family of new quantum codes can be derived from these dual-containing BCH codes. Some of these new antiprimitive quantum BCH codes are comparable with those derived from primitive BCH codes.
General Logic-Systems and Consequence Operators

OpenAIRE

Herrmann, Robert A.

2005-01-01

In this paper, general logic-systems are investigated. It is shown that there are infinitely many finite consequence operators defined on a fixed language L that cannot be generated from a finite logic-system. It is shown that a set map is a finite consequence operator iff it is defined by a general logic-system.
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)
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.).
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)
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.)
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.
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.
Status of reactor core design code system in COSINE code package

Energy Technology Data Exchange (ETDEWEB)

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

2014-07-01

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

International Nuclear Information System (INIS)

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

2014-01-01

For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)
Comparison of MACCS users calculations for the international comparison exercise on probabilistic accident consequence assessment code, October 1989--June 1993

International Nuclear Information System (INIS)

Neymotin, L.

1994-04-01

Over the past several years, the OECD/NEA and CEC sponsored an international program intercomparing a group of six probabilistic consequence assessment (PCA) codes designed to simulate health and economic consequences of radioactive releases into atmosphere of radioactive materials following severe accidents at nuclear power plants (NPPs): ARANO (Finland), CONDOR (UK), COSYMA (CEC), LENA (Sweden), MACCS (USA), and OSCAAR (Japan). In parallel with this effort, two separate groups performed similar calculations using the MACCS and COSYMA codes. Results produced in the MACCS Users Group (Greece, Italy, Spain, and USA) calculations and their comparison are contained in the present report. Version 1.5.11.1 of the MACCS code was used for the calculations. Good agreement between the results produced in the four participating calculations has been reached, with the exception of the results related to the ingestion pathway dose predictions. The main reason for the scatter in those particular results is attributed to the lack of a straightforward implementation of the specifications for agricultural production and counter-measures criteria provided for the exercise. A significantly smaller scatter in predictions of other consequences was successfully explained by differences in meteorological files and weather sampling, grids, rain distance intervals, dispersion model options, and population distributions
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)
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
AUS98 - The 1998 version of the AUS modular neutronic code system

Energy Technology Data Exchange (ETDEWEB)

Robinson, G.S.; Harrington, B.V

1998-07-01

AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energy deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous AUS publications are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM main-frame computers to UNIX workstations This report gives details of all system aspects of AUS and all modules except the POW3D multi-dimensional diffusion module refs., tabs.
AUS98 - The 1998 version of the AUS modular neutronic code system

International Nuclear Information System (INIS)

Robinson, G.S.; Harrington, B.V.

1998-07-01

AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energy deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous AUS publications are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM main-frame computers to UNIX workstations This report gives details of all system aspects of AUS and all modules except the POW3D multi-dimensional diffusion module
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
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.

Participation in the international comparison of probabilistic consequence assessment codes organized by OECD/NEA and CEC. Final Report

International Nuclear Information System (INIS)

Rossi, J.

1994-02-01

Probabilistic Consequence Assessment (PCA) methods are exploited not only in risk evaluation but also to study alternative design features, reactor siting recommendations and to obtain acceptable dose criteria by the radiation safety authorities. The models are programmed into computer codes for these kind of assessment. To investigate the quality and competence of different models, OECD/NEA and CEC organized the international code comparison exercise, which was participated by the organizations from 15 countries. There were seven codes participating in the exercise. The objectives of the code comparison exercise were to compare the results by the codes, to contribute to PCA code quality assurance, to harmonize the codes, to provide a forum for discussion on various approaches and to produce the report on the exercise. The project started in 1991 and the results of the calculations were completed in autumn 1992. The international report consists of two parts: the Overview Report for decision makers and the supporting detailed Technical Report. The results of the project are reviewed as an user of the ARANO-programme of VTT and trends of it's further development are indicated in this report. (orig.) (11 refs., 13 figs., 4 tabs.)
Validation and application of the system code ATHLET-CD for BWR severe accident analyses

Energy Technology Data Exchange (ETDEWEB)

Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Imke, Uwe; Sanchez, Victor

2016-10-15

Highlights: • We present the application of the system code ATHLET-CD code for BWR safety analyses. • Validation of core in-vessel models is performed based on KIT CORA experiments. • A SB-LOCA scenario is simulated on a generic German BWR plant up to vessel failure. • Different core reflooding possibilities are investigated to mitigate the accident consequences. • ATHLET-CD modelling features reflect the current state of the art of severe accident codes. - Abstract: This paper is aimed at the validation and application of the system code ATHLET-CD for the simulation of severe accident phenomena in Boiling Water Reactors (BWR). The corresponding models for core degradation behaviour e.g., oxidation, melting and relocation of core structural components are validated against experimental data available from the CORA-16 and -17 bundle tests. Model weaknesses are discussed along with needs for further code improvements. With the validated ATHLET-CD code, calculations are performed to assess the code capabilities for the prediction of in-vessel late phase core behaviour and reflooding of damaged fuel rods. For this purpose, a small break LOCA scenario for a generic German BWR with postulated multiple failures of the safety systems was selected. In the analysis, accident management measures represented by cold water injection into the damaged reactor core are addressed to investigate the efficacy in avoiding or delaying the failure of the reactor pressure vessel. Results show that ATHLET-CD is applicable to the description of BWR plant behaviour with reliable physical models and numerical methods adopted for the description of key in-vessel phenomena.
AUTOET code (a code for automatically constructing event trees and displaying subsystem interdependencies)

International Nuclear Information System (INIS)

Wilson, J.R.; Burdick, G.R.

1977-06-01

This is a user's manual for AUTOET I and II. AUTOET I is a computer code for automatic event tree construction. It is designed to incorporate and display subsystem interdependencies and common or key component dependencies in the event tree format. The code is written in FORTRAN IV for the CDC Cyber 76 using the Integrated Graphics System (IGS). AUTOET II incorporates consequence and risk calculations, in addition to some other refinements. 5 figures
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)
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).
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...
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
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
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...
Assessing economic consequences of radiation accidents

International Nuclear Information System (INIS)

Rowe, M.D.; Lee, J.C.; Grimshaw, C.A.; Kalb, P.D.

1987-01-01

A recent review of existing models and methods for assessing potential consequences of accidents in the high-level radioactive waste (HLW) disposal system identifies economic consequence assessment methods as a weak point. Existing methods have mostly been designed to assess economic consequences of reactor accidents, the possible scale of which can be several orders of magnitude greater than anything possible in the HLW disposal system. There is therefore some question about the applicability of these methods, their assumptions, and their level of detail to assessments of smaller accidents. The US Dept. of Energy funded this study to determine needs for code modifications or model development for assessing economic costs of accidents in the HLW disposal system. The objectives of the study were as follows: (1) review the literature on economic consequences of accidents to determine the availability of assessment methods and data and their applicability to the HLW disposal system before closure. (2) Determine needs for expansion, revision, or adaptation of methods and data for modeling economic consequences of accidents of the scale projected for the disposal system. (3) Gather data that might be useful for the needed revisions for modeling economic impacts on this scale
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)
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)
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.
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.
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)
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.
Ex-plant consequence assessment for NUREG-1150: models, typical results, uncertainties

International Nuclear Information System (INIS)

Sprung, J.L.

1988-01-01

The assessment of ex-plant consequences for NUREG-1150 source terms was performed using the MELCOR Accident Consequence Code System (MACCS). This paper briefly discusses the following elements of MACCS consequence calculations: input data, phenomena modeled, computational framework, typical results, controlling phenomena, and uncertainties. Wherever possible, NUREG-1150 results will be used to illustrate the discussion. 28 references
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.
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.
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...

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)
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)
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).
Improvement of calculation method for temperature coefficient of HTTR by neutronics calculation code based on diffusion theory. Analysis for temperature coefficient by SRAC code system

International Nuclear Information System (INIS)

Goto, Minoru; Takamatsu, Kuniyoshi

2007-03-01

The HTTR temperature coefficients required for the core dynamics calculations had been calculated from the HTTR core calculation results by the diffusion code with which the corrections had been performed using the core calculation results by the Monte-Carlo code MVP. This calculation method for the temperature coefficients was considered to have some issues to be improved. Then, the calculation method was improved to obtain the temperature coefficients in which the corrections by the Monte-Carlo code were not required. Specifically, from the point of view of neutron spectrum calculated by lattice calculations, the lattice model was revised which had been used for the calculations of the temperature coefficients. The HTTR core calculations were performed by the diffusion code with the group constants which were generated by the lattice calculations with the improved lattice model. The core calculations and the lattice calculations were performed by the SRAC code system. The HTTR core dynamics calculation was performed with the temperature coefficient obtained from the core calculation results. In consequence, the core dynamics calculation result showed good agreement with the experimental data and the valid temperature coefficient could be calculated only by the diffusion code without the corrections by Monte-Carlo code. (author)
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)
Improvement of MSLB transient analysis for VVER by the coupled code system KIKO3D/ATHLET

International Nuclear Information System (INIS)

Hegyi, Gy.; Kereszturi, A.; Trosztel, I.

2001-01-01

An overview is given on the investigations of the Main Steam Line Break transient in a VVER- 440 NPP by using the KIKO3D/ATHLET 1.2.A coupled code system. Special attention was paid for the influence of modeling the outcore detector signals and the malfunctioning of the emergency control system (scram with stuck rod). The conservatism of the calculations was assured even in the case of application of the 3D best estimate KIKO3D code. The consequence of MSLB accident is investigated at the end of cycle (EOC), at full power (FP) and shut down initial conditions. Even if very strong conservative assumptions were applied, dangerous hot spots were not found in the supposed scenarios.(author)
Non-binary unitary error bases and quantum codes

Energy Technology Data Exchange (ETDEWEB)

Knill, E.

1996-06-01

Error operator bases for systems of any dimension are defined and natural generalizations of the bit-flip/ sign-change error basis for qubits are given. These bases allow generalizing the construction of quantum codes based on eigenspaces of Abelian groups. As a consequence, quantum codes can be constructed form linear codes over {ital Z}{sub {ital n}} for any {ital n}. The generalization of the punctured code construction leads to many codes which permit transversal (i.e. fault tolerant) implementations of certain operations compatible with the error basis.
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
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.
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+.
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.
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.
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)
Development and Application of a Code for Internal Exposure (CINEX) based on the CINDY code

International Nuclear Information System (INIS)

Kravchik, T.; Duchan, N.; Sarah, R.; Gabay, Y.; Kol, R.

2004-01-01

Internal exposure to radioactive materials at the NRCN is evaluated using the CINDY (Code for Internal Dosimetry) Package. The code was developed by the Pacific Northwest Laboratory to assist the interpretation of bioassay data, provide bioassay projections and evaluate committed and calendar-year doses from intake or bioassay measurement data. It provides capabilities to calculate organ dose and effective dose equivalents using the International Commission on Radiological Protection (ICRP) 30 approach. The CINDY code operates under DOS operating system and consequently its operation needs a relatively long procedure which also includes a lot of manual typing that can lead to personal human mistakes. A new code has been developed at the NRCN, the CINEX (Code for Internal Exposure), which is an Excel application and leads to a significant reduction in calculation time (in the order of 5-10 times) and in the risk of personal human mistakes. The code uses a database containing tables which were constructed by the CINDY and contain the bioassay values predicted by the ICRP30 model after an intake of an activity unit of each isotope. Using the database, the code than calculates the appropriate intake and consequently the committed effective dose and organ dose. Calculations with the CINEX code were compared to similar calculations with the CINDY code. The discrepancies were less than 5%, which is the rounding error of the CINDY code. Attached is a table which compares parameters calculated with the CINEX and the CINDY codes (for a class Y uranium). The CINEX is now used at the NRCN to calculate occupational intakes and doses to workers with radioactive materials
Abstract feature codes: The building blocks of the implicit learning system.

Science.gov (United States)

Eberhardt, Katharina; Esser, Sarah; Haider, Hilde

2017-07-01

According to the Theory of Event Coding (TEC; Hommel, Müsseler, Aschersleben, & Prinz, 2001), action and perception are represented in a shared format in the cognitive system by means of feature codes. In implicit sequence learning research, it is still common to make a conceptual difference between independent motor and perceptual sequences. This supposedly independent learning takes place in encapsulated modules (Keele, Ivry, Mayr, Hazeltine, & Heuer 2003) that process information along single dimensions. These dimensions have remained underspecified so far. It is especially not clear whether stimulus and response characteristics are processed in separate modules. Here, we suggest that feature dimensions as they are described in the TEC should be viewed as the basic content of modules of implicit learning. This means that the modules process all stimulus and response information related to certain feature dimensions of the perceptual environment. In 3 experiments, we investigated by means of a serial reaction time task the nature of the basic units of implicit learning. As a test case, we used stimulus location sequence learning. The results show that a stimulus location sequence and a response location sequence cannot be learned without interference (Experiment 2) unless one of the sequences can be coded via an alternative, nonspatial dimension (Experiment 3). These results support the notion that spatial location is one module of the implicit learning system and, consequently, that there are no separate processing units for stimulus versus response locations. (PsycINFO Database Record (c) 2017 APA, all rights reserved).
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
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
Calculation code for erosion-corrosion induced wall thinning in piping systems

International Nuclear Information System (INIS)

Henzel, N.; Kastner, W.; Stellwag, B.; Erve, M.

1988-01-01

There was great material erosion mainly in consequence of an extremely unfavourable geometry at the damaged place in Surry-2. The pipeline sections affected in Trojan were in the area of action of great sources of turbulence, i.e.: less than 10 pipe diameters from junctions, elbows etc. Because of the many parameters which determine the amount of material removal by erosion-corrosion, the analysis of such damage is only possible using a computer program. The main purpose of such a PC code called WATHEC developed by Siemens/KWU is not the subsequent confirmation of damage which has occurred, but its application for preventive diagnosis in pipeline systems. (orig./DG) [de
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)
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.

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
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
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.)
Apply Functional Modelling to Consequence Analysis in Supervision Systems

DEFF Research Database (Denmark)

Zhang, Xinxin; Lind, Morten; Gola, Giulio

2013-01-01

This paper will first present the purpose and goals of applying functional modelling approach to consequence analysis by adopting Multilevel Flow Modelling (MFM). MFM Models describe a complex system in multiple abstraction levels in both means-end dimension and whole-part dimension. It contains...... consequence analysis to practical or online applications in supervision systems. It will also suggest a multiagent solution as the integration architecture for developing tools to facilitate the utilization results of functional consequence analysis. Finally a prototype of the multiagent reasoning system...... causal relations between functions and goals. A rule base system can be developed to trace the causal relations and perform consequence propagations. This paper will illustrate how to use MFM for consequence reasoning by using rule base technology and describe the challenges for integrating functional...
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
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
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)
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.
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
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.
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
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.
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.
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
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
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)
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
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)
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)
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.

CAT: a computer code for the automated construction of fault trees

International Nuclear Information System (INIS)

Apostolakis, G.E.; Salem, S.L.; Wu, J.S.

1978-03-01

A computer code, CAT (Computer Automated Tree, is presented which applies decision table methods to model the behavior of components for systematic construction of fault trees. The decision tables for some commonly encountered mechanical and electrical components are developed; two nuclear subsystems, a Containment Spray Recirculation System and a Consequence Limiting Control System, are analyzed to demonstrate the applications of CAT code
Estimation of the Radiological Consequences of Fukushima Dai-ichi Nuclear Power Plant Accident using MACCS2

Energy Technology Data Exchange (ETDEWEB)

Kim, Sora; Min, Byung-Il; Park, Kihyun; Yang, Byung-Mo; Suh, Kyung-suk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

Three of them have undergone fuel melting and hydrogen explosions. A significant amount of radioactive material was released into the atmosphere from FDNPP and dispersed all over the world. In this study, we assessed the offsite consequences of Fukushima disaster in the region within a 30-km radius of FDNPP using the MELCOR Accident Consequence Code Systems 2(MACCS2) code, which is the Nuclear Regulatory Commission's (NRC's) code. The reflection of the realistic regional characteristics, such as long-term meteorological data, site- and population-specific data, and radiation safety regulatory, is essential to accurately analyze the off-site consequences. The assessment that reflects regional characteristics would contribute to identify main causes of exposure doses and to find the effective countermeasures for minimizing the accidental off-site consequences.
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
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.
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.
Dose calculations for a simplified Mammosite system with the Monte Carlo Penelope and MCNPX simulation codes

International Nuclear Information System (INIS)

Rojas C, E.L.; Varon T, C.F.; Pedraza N, R.

2007-01-01

The treatment of the breast cancer at early stages is of vital importance. For that, most of the investigations are dedicated to the early detection of the suffering and their treatment. As investigation consequence and clinical practice, in 2002 it was developed in U.S.A. an irradiation system of high dose rate known as Mammosite. In this work we carry out dose calculations for a simplified Mammosite system with the Monte Carlo Penelope simulation code and MCNPX, varying the concentration of the contrast material that it is used in the one. (Author)
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
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.
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
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...
Toward a Systemic Notion of Information: Practical Consequences

Directory of Open Access Journals (Sweden)

Nagib Callaos

2002-01-01

Full Text Available Our main purpose in this paper is to start a process of a systemic definition of the notion of information and to provide some initial practical consequences of it. We will try to do that providing: 1 a conceptual definition, following Ackoff's (1962 description and method of such a kind of definition, and 2 following Peirce's (1931-5,1958 conception of "meaning", where the practical consequences should be included. To our knowledge, no attempt has been done up to the present neither to find a Peircean meaning to the notion of information, nor to start a process of describing a systemic notion of information. Consequently, we will try to integrate the different definitions made on information. But to integrate we should first differentiate what is to be integrated. Thus, we will typify information conceptions in subjective and objective, providing brief description and analysis of each type, integrating them in the context of a systemic notion of information, and drawing the respective pragmatic consequences, as required by Peirce, for any meaning description, and by a pragmatic-teleological systemic epistemology (Churchmann, 1971
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.
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
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
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
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)
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.
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)
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
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.

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
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...
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.
Code Coupling for Multi-Dimensional Core Transient Analysis

International Nuclear Information System (INIS)

Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il

2015-01-01

After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident
Code Coupling for Multi-Dimensional Core Transient Analysis

Energy Technology Data Exchange (ETDEWEB)

Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il [KEPCO NF, Daejeon (Korea, Republic of)

2015-05-15

After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident.
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
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)
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.
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)
The EGS5 Code System

Energy Technology Data Exchange (ETDEWEB)

Hirayama, Hideo; Namito, Yoshihito; /KEK, Tsukuba; Bielajew, Alex F.; Wilderman, Scott J.; U., Michigan; Nelson, Walter R.; /SLAC

2005-12-20

In the nineteen years since EGS4 was released, it has been used in a wide variety of applications, particularly in medical physics, radiation measurement studies, and industrial development. Every new user and every new application bring new challenges for Monte Carlo code designers, and code refinements and bug fixes eventually result in a code that becomes difficult to maintain. Several of the code modifications represented significant advances in electron and photon transport physics, and required a more substantial invocation than code patching. Moreover, the arcane MORTRAN3[48] computer language of EGS4, was highest on the complaint list of the users of EGS4. The size of the EGS4 user base is difficult to measure, as there never existed a formal user registration process. However, some idea of the numbers may be gleaned from the number of EGS4 manuals that were produced and distributed at SLAC: almost three thousand. Consequently, the EGS5 project was undertaken. It was decided to employ the FORTRAN 77 compiler, yet include as much as possible, the structural beauty and power of MORTRAN3. This report consists of four chapters and several appendices. Chapter 1 is an introduction to EGS5 and to this report in general. We suggest that you read it. Chapter 2 is a major update of similar chapters in the old EGS4 report[126] (SLAC-265) and the old EGS3 report[61] (SLAC-210), in which all the details of the old physics (i.e., models which were carried over from EGS4) and the new physics are gathered together. The descriptions of the new physics are extensive, and not for the faint of heart. Detailed knowledge of the contents of Chapter 2 is not essential in order to use EGS, but sophisticated users should be aware of its contents. In particular, details of the restrictions on the range of applicability of EGS are dispersed throughout the chapter. First-time users of EGS should skip Chapter 2 and come back to it later if necessary. With the release of the EGS4 version
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)
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
Severe accident consequence mitigation by filtered containment venting at Canadian nuclear power plants

International Nuclear Information System (INIS)

Lebel, Luke S.; Morreale, Andrew C.; Korolevych, Volodymyr; Brown, Morgan J.; Gyepi-Garbrah, Sam

2017-01-01

Highlights: • Use of filtered containment venting during a severe accident assessed. • Severe accident simulations performed using MAAP-CANDU and ADDAM. • Flow capacity, initiation protocols, efficiency, mass and thermal loading evaluated. • Efficient, robust system drastically reduces accident consequences. - Abstract: Having the capability to use filtered containment venting during a severe nuclear accident can significantly reduce its overall consequences. This study employs the MAAP-CANDU severe accident analysis code and the ADDAM atmospheric dispersion code to study the progression of: an unmitigated station blackout accident at a generic pressurized heavy water reactor, the release of radioactive material into the environment, the subsequent dispersion of the fission products through the atmosphere and the subsequent consequences (evacuation radius). The goal is to evaluate the application of filtered venting as an accident mitigation technology. Several aspects of filtered containment venting system design, like flow capacity, initiation protocols, filter efficiency, mass loading, and thermal loading are considered. An efficient and robust filtered containment venting system can reduce the amount of radiological materials emitted during an accident by 25 times or more, and as a result considerably reduce the off-site consequences of an accident.
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
A Critical Appraisal of the Juvenile Justice System under Cameroon's 2005 Criminal Procedure Code: Emerging Challenges

Directory of Open Access Journals (Sweden)

S Tabe

2012-03-01

Full Text Available The objective of this article is to examine the changes introduced by the 2005 Cameroonian Criminal Procedure Code on matters of juvenile justice, considering that before this Code, juvenile justice in Cameroon was governed by extra-national laws. In undertaking this analysis, the article highlights the evolution of the administration of juvenile justice 50 years after independence of Cameroon. It also points out the various difficulties and shortcomings in the treatment of juvenile offenders in Cameroon since the enactment of the new Criminal Procedure Code. The article reveals that the 2005 Code is an amalgamation of all hitherto existing laws in the country that pertained to juvenile justice, and that despite the considerable amount of criticism it has received, the Code is clearly an improvement of the system of juvenile justice in Cameroon, since it represents a balance of the due process rights of young people, the protection of society and the special needs of young offenders. This is so because the drafters of the Code took a broad view of the old laws on juvenile justice. Also a wide range of groups were consulted, including criminal justice professionals, children’s service organisations, victims, parents, young offenders, educators, advocacy groups and social-policy analysts. However, to address the challenges that beset the juvenile justice system of Cameroon, the strategy of the government should be focussed on three areas: the prevention of youth crime, the provision of meaningful consequences for the actions of young people, and the rehabilitation and reintegration of young offenders. Cameroonian law should seek educative solutions rather than to impose prison sentences or other repressive measures on young offenders. Special courts to deal with young offenders should be established outside the regular penal system and should be provided with resources that are adequate for and appropriate to fostering their understanding of
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.
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.
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)
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.
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

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.
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.
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.
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
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
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)
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
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
Application of the cause-consequence diagram method to static systems

International Nuclear Information System (INIS)

Andrews, J.D.; Ridley, L.M.

2002-01-01

In the last 30 years, various mathematical models have been used to identify the effect of component failures on the performance of a system. The most frequently used technique for system reliability assessment is Fault Tree Analysis (FTA) and a large proportion of its popularity can be attributed to the fact that it provides a very good documentation of the way that the system failure logic was developed. Exact quantification of the fault tree, however, can be problematic for very large systems and in such situations, approximations can be used. Alternatively, an exact result can be obtained via the conversion of the fault tree into a binary decision diagram (BDD). The BDD, however, loses all failure logic documentation during the conversion process. This paper outlines the use of the cause-consequence diagram method as a tool for system risk and reliability analysis. As with the FTA method, the cause-consequence diagram documents the failure logic of the system. In addition to this the cause-consequence diagram produces the exact failure probability in a very efficient calculation procedure. The cause-consequence diagram technique has been applied to a static system and shown to yield the same result as those produced by the solution of the equivalent fault tree and BDD. On the basis of this general rules have been devised for the correct construction of the cause-consequence diagram given a static system. The use of the cause-consequence method in this manner has significant implications in terms of efficiency of the reliability analysis and can be shown to have benefits for static systems
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...
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
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.
Modern Nuclear Data Evaluation with the TALYS Code System

Science.gov (United States)

Koning, A. J.; Rochman, D.

2012-12-01

This paper presents a general overview of nuclear data evaluation and its applications as developed at NRG, Petten. Based on concepts such as robustness, reproducibility and automation, modern calculation tools are exploited to produce original nuclear data libraries that meet the current demands on quality and completeness. This requires a system which comprises differential measurements, theory development, nuclear model codes, resonance analysis, evaluation, ENDF formatting, data processing and integral validation in one integrated approach. Software, built around the TALYS code, will be presented in which all these essential nuclear data components are seamlessly integrated. Besides the quality of the basic data and its extensive format testing, a second goal lies in the diversity of processing for different type of users. The implications of this scheme are unprecedented. The most important are: 1. Complete ENDF-6 nuclear data files, in the form of the TENDL library, including covariance matrices, for many isotopes, particles, energies, reaction channels and derived quantities. All isotopic data files are mutually consistent and are supposed to rival those of the major world libraries. 2. More exact uncertainty propagation from basic nuclear physics to applied (reactor) calculations based on a Monte Carlo approach: "Total" Monte Carlo (TMC), using random nuclear data libraries. 3. Automatic optimization in the form of systematic feedback from integral measurements back to the basic data. This method of work also opens a new way of approaching the analysis of nuclear applications, with consequences in both applied nuclear physics and safety of nuclear installations, and several examples are given here. This applied experience and feedback is integrated in a final step to improve the quality of the nuclear data, to change the users vision and finally to orchestrate their integration into simulation codes.
Modern Nuclear Data Evaluation with the TALYS Code System

International Nuclear Information System (INIS)

Koning, A.J.; Rochman, D.

2012-01-01

This paper presents a general overview of nuclear data evaluation and its applications as developed at NRG, Petten. Based on concepts such as robustness, reproducibility and automation, modern calculation tools are exploited to produce original nuclear data libraries that meet the current demands on quality and completeness. This requires a system which comprises differential measurements, theory development, nuclear model codes, resonance analysis, evaluation, ENDF formatting, data processing and integral validation in one integrated approach. Software, built around the TALYS code, will be presented in which all these essential nuclear data components are seamlessly integrated. Besides the quality of the basic data and its extensive format testing, a second goal lies in the diversity of processing for different type of users. The implications of this scheme are unprecedented. The most important are: 1. Complete ENDF-6 nuclear data files, in the form of the TENDL library, including covariance matrices, for many isotopes, particles, energies, reaction channels and derived quantities. All isotopic data files are mutually consistent and are supposed to rival those of the major world libraries. 2. More exact uncertainty propagation from basic nuclear physics to applied (reactor) calculations based on a Monte Carlo approach: “Total” Monte Carlo (TMC), using random nuclear data libraries. 3. Automatic optimization in the form of systematic feedback from integral measurements back to the basic data. This method of work also opens a new way of approaching the analysis of nuclear applications, with consequences in both applied nuclear physics and safety of nuclear installations, and several examples are given here. This applied experience and feedback is integrated in a final step to improve the quality of the nuclear data, to change the users vision and finally to orchestrate their integration into simulation codes.
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
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.
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
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.
Results of the Czech-Austrian calculations of BDBA radiological consequences

International Nuclear Information System (INIS)

Carny, P.; Hohenberg, J.-K.

2003-01-01

participates in proceeded comparisons of ingestion doses and procedures for ingestion doses calculations performed in the frame of this Czech-Austrian cooperation. For calculations of the long range radiological consequences under real meteorological conditions there are used by the Czech side and participate in these comparisons codes PTM, RODOS/MATCH and MEDIA (CHMI). Code PTM (Puff Trajectory Model, Lagrangian dispersion model) can be used as part of the code RTARC for modeling of dispersion in medium and large distances. MATCH dispersion model participates in this comparisons as part of RODOS system. Long range dispersion model MEDIA (CHMI) participates as it is basic support instrument for the Czech nuclear regulatory body. An Austrian cooperation consisting of Austrian Research Center Seibersdorf, ZAMG (Austrian Central Institute for Meteorology and Geodynamics), and the Radiation Protection Division in the Ministry of Agriculture, Forestry, Environment and Water Management performed their calculations with the following codes: TAMOS long range dispersion model using ECMWF data and OECOSYS the Austrian version of ECOSYS for dose assessment. In the step I there were performed 3 various exercises: deterministic and probabilistic evaluation of consequences of hypothetical accident source term; extended set of deterministic calculations with Austrian source term 'VVER-1000'; comparison of results of codes HERALD and HAVAR with the code PC Cosyma based an 'LBLOCA scenario'. Results and conclusions are presented in the paper. In the step II and III these have been performed another exercises: calculation of radiological situation after hypothetical accident in Temelin. These calculations have been performed with long range models and real sequence of meteorological situation (wind field) has been chosen for calculations; calculations of relevant radiological parameters at 2 sites an the Czech and at 2 sites an the Austrian territory and comparison of radiation protection
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.

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.
Evaluation of dose calculation models for inhabited areas applicable in nuclear accident consequence assessment codes

International Nuclear Information System (INIS)

Katalin Eged; Zoltan Kis; Natalia Semioschkina; Gabriele Voigt

2004-01-01

One of the objectives of the EC project EVANET-TERRA is to provide suitable inputs to the RODOS system. This study gives an overview on urban dose calculation models with special emphasis on the RECLAIM-EDEM2M and TEMAS-urban codes. The TEMAS-urban code is more complex compared to the RECLAIM-EDEM2M code although both models use similar and some times even same model parameters. The database and the way of its data collection as used in RECLAIM-EDEM2M is recommended as a preferred option because it contains many data from local and regional measurements. However in a decision situation the outputs of the TEMASurban model may better help stake holders by providing a ranking of the surfaces to be decontaminated. (author)
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.
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.)
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.)
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.)
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)
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...
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)
Thermo-hydraulic consequence of pressure suppression containment vessel during blowdown, 2

International Nuclear Information System (INIS)

Aya, Izuo; Nariai, Hideki; Kobayashi, Michiyuki

1980-01-01

As a part of the safety research works for the integral-type marine reactor, an analytical code SUPPAC-2V was developed to simulate the thermo-hydraulic consequence of a pressure suppression containment system during blowdown and the code was applied to the Model Experimental Facility of the Safety of Integral Type Marine Reactors (explained already in Part 1). SUPPAC-2V is much different from existing codes in the following points. A nonhomogeneous model for the gaseous region in the drywell, a new correlation for condensing heat transfer coefficient at drywell wall based on existing data and approximation of air bubbles in wetwell water by one dimensional bubble rising model are adopted in this code. In comparing calculational results with experimental results, values of predominant input parameters were evaluated and discussed. Moreover, the new code was applied also to the NSR-7 marine reactor, conceptually designed at the Shipbuilding Research Association in Japan, of which suppression system had been already analysed by CONTEMPT-PS. (author)
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.
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
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
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
Development of a Methodology for VHTR Accident Consequence Assessment

Energy Technology Data Exchange (ETDEWEB)

Lee, Joeun; Kim, Jintae; Jae, Moosung [Hanyang University, Seoul (Korea, Republic of)

2016-05-15

The substitution of the VHTR for burning fossil fuels conserves these hydrocarbon resources for other uses and eliminates the emissions of greenhouse. In Korea, for these reasons, constructing the VHTR plan for hydrogen production is in progress. In this study, the consequence analysis for the off-site releases of radioactive materials during severe accidents has been performed using the level 3 PRA technology. The offsite consequence analysis for a VHTR using the MACCS code has been performed. Since the passive system such as the RCCS(Reactor Cavity Cooling System) are equipped, the frequency of occurrence of accidents has been evaluated to be very low. For further study, the assessment for characteristic of VHTR safety system and precise quantification of its accident scenarios is expected to conduct more certain consequence analysis. This methodology shown in this study might contribute to enhancing the safety of VHTR design by utilizing the results having far lower effect on the environment than the LWRs.
[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
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...
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)
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...
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.

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)
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).
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
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)
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)
A Cause-Consequence Chart of a Redundant Protection System

DEFF Research Database (Denmark)

Nielsen, Dan Sandvik; Platz, O.; Runge, B.

1975-01-01

A cause-consequence chart is applied in analysing failures of a redundant protection system (a core spray system in a nuclear power plant). It is shown how the diagram provides a basis for calculating two probability measures for malfunctioning of the protection system. The test policy of compone...... of components is taken into account. The possibility of using parameter variation as a basis for the choice of test policy is indicated.......A cause-consequence chart is applied in analysing failures of a redundant protection system (a core spray system in a nuclear power plant). It is shown how the diagram provides a basis for calculating two probability measures for malfunctioning of the protection system. The test policy...
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
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
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
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
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
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)
Modeling of pipe break accident in a district heating system using RELAP5 computer code

International Nuclear Information System (INIS)

Kaliatka, A.; Valinčius, M.

2012-01-01

Reliability of a district heat supply system is a very important factor. However, accidents are inevitable and they occur due to various reasons, therefore it is necessary to have possibility to evaluate the consequences of possible accidents. This paper demonstrated the capabilities of developed district heating network model (for RELAP5 code) to analyze dynamic processes taking place in the network. A pipe break in a water supply line accident scenario in Kaunas city (Lithuania) heating network is presented in this paper. The results of this case study were used to demonstrate a possibility of the break location identification by pressure decrease propagation in the network. -- Highlights: ► Nuclear reactor accident analysis code RELAP5 was applied for accident analysis in a district heating network. ► Pipe break accident scenario in Kaunas city (Lithuania) district heating network has been analyzed. ► An innovative method of pipe break location identification by pressure-time data is proposed.
Development and validation of sodium fire analysis code ASSCOPS

International Nuclear Information System (INIS)

Ohno, Shuji

2001-01-01

A version 2.1 of the ASSCOPS sodium fire analysis code was developed to evaluate the thermal consequences of a sodium leak and consequent fire in LMFBRs. This report describes the computational models and the validation studies using the code. The ASSCOPS calculates sodium droplet and pool fire, and consequential heat/mass transfer behavior. Analyses of sodium pool or spray fire experiments confirmed that this code and parameters used in the validation studies gave valid results on the thermal consequences of sodium leaks and fires. (author)
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
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
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.
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.
The feasibility of QR-code prescription in Taiwan.

Science.gov (United States)

Lin, C-H; Tsai, F-Y; Tsai, W-L; Wen, H-W; Hu, M-L

2012-12-01

An ideal Health Care Service is a service system that focuses on patients. Patients in Taiwan have the freedom to fill their prescriptions at any pharmacies contracted with National Health Insurance. Each of these pharmacies uses its own computer system. So far, there are at least ten different systems on the market in Taiwan. To transmit the prescription information from the hospital to the pharmacy accurately and efficiently presents a great issue. This study consisted of two-dimensional applications using a QR-code to capture Patient's identification and prescription information from the hospitals as well as using a webcam to read the QR-code and transfer all data to the pharmacy computer system. Two hospitals and 85 community pharmacies participated in the study. During the trial, all participant pharmacies appraised highly of the accurate transmission of the prescription information. The contents in QR-code prescriptions from Taipei area were picked up efficiently and accurately in pharmacies at Taichung area (middle Taiwan) without software system limit and area limitation. The QR-code device received a patent (No. M376844, March 2010) from Intellectual Property Office Ministry of Economic Affair, China. Our trial has proven that QR-code prescription can provide community pharmacists an efficient, accurate and inexpensive device to digitalize the prescription contents. Consequently, pharmacists can offer better quality of pharmacy service to patients. © 2012 Blackwell Publishing Ltd.
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

A radiological accident consequence assessment system for Hong Kong

International Nuclear Information System (INIS)

Wong, M.C.; Lam, H.K.

1993-01-01

An account is given of the Hong Kong Radiological Accident Consequence Assessment System which would be used to assess the potential consequences of an emergency situation involving atmospheric release of radioactive material. The system has the capability to acquire real-time meteorological information from the Observatory's network of automatic stations, synoptic stations in the nearby region as well as forecast data from numerical prediction models. The system makes use of these data to simulate the transport and dispersion of the released radioactive material. The effectiveness of protective action on the local population is also modeled. The system serves as a powerful aid in the protective action recommendation processes
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.
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.
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
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.
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.
Evaluation of consequences and risks in Slovenia

International Nuclear Information System (INIS)

Susnik, J.

1996-01-01

The paper describes the evaluation of nuclear power plant accident consequences and risks using probabilistic safety codes during the last 12 years at the J. Stefan Institute. They cover classic individual and population risk studies due to assumed potential severe accident scenarios, prediction and estimation of Chernobyl accident consequences, the optimization of emergency countermeasures at the Krsko site, where the 632 MWe Westinghouse PWR NPP went into commercial operation on January 1983, and the ranking of population risk within the public debate in connection with the civil initiative to close the NPP Krsko. We report on the initial use of the CRAC2 code in 1984 and later, when it was first applied for the study of population risk in the area of the second planned Slovenian-Croatian NPP for the Prevlaka site. The study was completed a few weeks before the Chernobyl accident in April 1986. Risk evaluation was also included in the analysis of nuclear safety at the NPP Krsko during the war for Slovenia's independence in 1991. We report on the (CRAC2) analyses of the Chernobyl accident: on initial estimation of the maximal potentially expected consequences in Slovenia, on the effect of the radioactive cloud rise on the consequences relatively close to the NPP; on the further research after the detailed information on the radioactivity release and on the air masses movement were published; then the cloud activity which moved towards Slovenia was assessed and the expected consequences along its path were calculated. As the calculated integral individual exposure to the I 131 inhalation and the ground Cs 137 contamination matched with the measurements in Ljubljana and with the UNSCEAR 1988 data, our reliance on the CRAC2 code and on its ancestors is high. We report on the analyses, performed by the CRAC2 code and since 1993 also by the PC COSYMA code, related to the countermeasure effects. The consequences studied were extended to late health effects. We analyzed
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.
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
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
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.
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.
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
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
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.
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
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.
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
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.
Simulation for photon detection in spectrometric system of high purity (HPGe) using MCNPX code

International Nuclear Information System (INIS)

Correa, Guilherme Jorge de Souza

2013-01-01

The Brazilian National Commission of Nuclear Energy defines parameters for classification and management of radioactive waste in accordance with the activity of materials. The efficiency of a detection system is crucial to determine the real activity of a radioactive source. When it's possible, the system's calibration should be performed using a standard source. Unfortunately, there are only a few cases that it can be done this way, considering the difficulty of obtaining appropriate standard sources for each type of measurement. So, computer simulations can be performed to assist in calculating of the efficiency of the system and, consequently, also auxiliary the classification of radioactive waste. This study aims to model a high purity germanium (HPGe) detector with MCNPX code, approaching the spectral values computationally obtained of the values experimentally obtained for the photopeak of 137 Cs. The approach will be made through changes in outer dead layer of the germanium crystal modeled. (author)

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.
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.
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.
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.
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).
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)
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...
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.
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)
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.
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...
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.
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.
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.)
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
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.
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.
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.)
Selected systems engineering process deficiencies and their consequences

Science.gov (United States)

Thomas, L. Dale

2007-06-01

The systems engineering process is well established and well understood. While this statement could be argued in the light of the many systems engineering guidelines and that have been developed, comparative review of these respective descriptions reveal that they differ primarily in the number of discrete steps or other nuances, and are at their core essentially common. Likewise, the systems engineering textbooks differ primarily in the context for application of systems engineering or in the utilization of evolved tools and techniques, not in the basic method. Thus, failures in systems engineering cannot credibly be attributed to implementation of the wrong systems engineering process among alternatives. However, numerous system failures can be attributed to deficient implementation of the systems engineering process. What may clearly be perceived as a systems engineering deficiency in retrospect can appear to be a well considered system engineering efficiency in real time—an efficiency taken to reduce cost or meet a schedule, or more often both. Typically these efficiencies are grounded on apparently solid rationale, such as reuse of heritage hardware or software. Over time, unintended consequences of a systems engineering process deficiency may begin to be realized, and unfortunately often the consequence is systems failure. This paper describes several actual cases of system failures that resulted from deficiencies in their systems engineering process implementation, including the Ariane 5 and the Hubble Space Telescope.
Improved atmospheric dispersion modelling in the new program system UFOMOD for accident consequence assessments

International Nuclear Information System (INIS)

Panitz, H.J.

1988-01-01

An essential aim of the improvements of the new program system UFOMOD for Accident Consequence Assessments (ACAs) was to substitute the straightline Gaussian plume model conventionally used in ACA models by more realistic atmospheric dispersion models. To identify improved models which can be applied in ACA codes and to quantify the implications of different concepts of dispersion modelling on the results of an ACA, probabilistic comparative calculations with different atmospheric dispersion models have been carried out. The study showed that there are trajectory models available which can be applied in ACAs and that these trajectory models provide more realistic results of ACAs than straight-line Gaussian models. This led to a completly novel concept of atmospheric dispersion modelling which distinguish between two different distance ranges of validity: the near range ( 50 km). The two ranges are assigned to respective trajectory models

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
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
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
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
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.
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.)
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
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.
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
Validation of thermalhydraulic codes

International Nuclear Information System (INIS)

Wilkie, D.

1992-01-01

Thermalhydraulic codes require to be validated against experimental data collected over a wide range of situations if they are to be relied upon. A good example is provided by the nuclear industry where codes are used for safety studies and for determining operating conditions. Errors in the codes could lead to financial penalties, to the incorrect estimation of the consequences of accidents and even to the accidents themselves. Comparison between prediction and experiment is often described qualitatively or in approximate terms, e.g. ''agreement is within 10%''. A quantitative method is preferable, especially when several competing codes are available. The codes can then be ranked in order of merit. Such a method is described. (Author)
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)
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
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.
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.
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.
Validation of the U.S. NRC coupled code system TRITON/TRACE/PARCS with the special power excursion reactor test III (SPERT III)

Energy Technology Data Exchange (ETDEWEB)

Wang, R. C.; Xu, Y.; Downar, T. [Dept. of Nuclear Engineering and Radiological Sciences, Univ. of Michigan, Ann Arbor, MI 48104 (United States); Hudson, N. [RES Div., U.S. NRC, Rockville, MD (United States)

2012-07-01

The Special Power Excursion Reactor Test III (SPERT III) was a series of reactivity insertion experiments conducted in the 1950's. This paper describes the validation of the U.S. NRC Coupled Code system TRITON/PARCS/TRACE to simulate reactivity insertion accidents (RIA) by using several of the SPERT III tests. The work here used the SPERT III E-core configuration tests in which the RIA was initiated by ejecting a control rod. The resulting super-prompt reactivity excursion and negative reactivity feedback produced the familiar bell shaped power increase and decrease. The energy deposition during such a power peak has important safety consequences and provides validation basis for core coupled multi-physics codes. The transients of five separate tests are used to benchmark the PARCS/TRACE coupled code. The models were thoroughly validated using the original experiment documentation. (authors)
Modeling the consequences of hypothetical accidents for the Titan II system

International Nuclear Information System (INIS)

Greenly, G.D.; Sullivan, T.J.

1981-11-01

Calculations have been made with the Atmospheric Release Advisory Capability (ARAC) suite of three-dimensional transport and diffusion codes MATHEW/ADPIC to assess the consequences of severe, hypothetical accident scenarios. One set of calculations develops the integrated dose and surface deposition patterns for a non-nuclear, high explosive detonation and dispersal of material. A second set of calculations depicts the time integrated dose and instantaneous concentration patterns for a substantial, continuous leak of the missile fuel oxidizer converted to nitrogen dioxide (NO 2 ). The areas affected and some of the implications for emergency response management are discussed
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.
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.)
Development of an accident consequence assessment code for evaluating site suitability of light- and heavy-water reactors based on the Korean Technical standards

Energy Technology Data Exchange (ETDEWEB)

Hwang, Won Tae; Jeong, Hae Sung; Jeong, Hyo Joon; Kil, A Reum; Kim, Eun Han; Han, Moon Hee [Nuclear Environment Safety Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-12-15

Methodologies for a series of radiological consequence assessments show a distinctive difference according to the design principles of the original nuclear suppliers and their technical standards to be imposed. This is due to the uncertainties of the accidental source term, radionuclide behavior in the environment, and subsequent radiological dose. Both types of PWR and PHWR are operated in Korea. However, technical standards for evaluating atmospheric dispersion have been enacted based on the U.S. NRC's positions regardless of the reactor types. For this reason, it might cause a controversy between the licensor and licensee of a nuclear power plant. It was modelled under the framework of the NRC Regulatory Guide 1.145 for light-water reactors, reflecting the features of heavy-water reactors as specified in the Canadian National Standard and the modelling features in MACCS2, such as atmospheric diffusion coefficient, ground deposition, surface roughness, radioactive plume depletion, and exposure from ground deposition. An integrated accident consequence assessment code, ACCESS (Accident Consequence Assessment Code for Evaluating Site Suitability), was developed by taking into account the unique regulatory positions for reactor types under the framework of the current Korean technical standards. Field tracer experiments and hand calculations have been carried out for validation and verification of the models. The modelling approaches of ACCESS and its features are introduced, and its applicative results for a hypothetical accidental scenario are comprehensively discussed. In an applicative study, the predicted results by the light-water reactor assessment model were higher than those by other models in terms of total doses.

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.
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.
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
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.
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.
AECL's advanced code program

Energy Technology Data Exchange (ETDEWEB)

McGee, G.; Ball, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

2012-07-01

This paper discusses the advanced code project at AECL.Current suite of Analytical, Scientific and Design (ASD) computer codes in use by Canadian Nuclear Power Industry is mostly developed 20 or more years ago. It is increasingly difficult to develop and maintain. It consist of many independent tools and integrated analysis is difficult, time consuming and error-prone. The objectives of this project is to demonstrate that nuclear facility systems, structures and components meet their design objectives in terms of function, cost, and safety; demonstrate that the nuclear facility meets licensing requirements in terms of consequences of off-normal events; dose to public, workers, impact on environment and demonstrate that the nuclear facility meets operational requirements with respect to on-power fuelling and outage management.
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.
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)
Development of web-based Off-Site Consequence Analysis Program (OSCAP) for extending ILRT intervals and its application

International Nuclear Information System (INIS)

Jeon, Ho-Jun; Hwang, Seok-Won; Oh, Ji-Yong

2012-01-01

Highlights: ► We develop web-based offsite consequence analysis program based on MACCS II code. ► The program has an automatic processing module to make the main input data. ► It is effective in conducting risk assessments according to extending ILRT intervals. ► Even a beginner can perform offsite consequence analysis with the program. - Abstract: For an offsite consequence analysis, MELCOR Accident Consequence Code System (MACCS) II code is widely used as a tool. In this study, the algorithm of web-based Off-Site Consequence Analysis Program (OSCAP) using the MACCS II code was developed for an integrated leak rate test (ILRT) interval extension and Level 3 probabilistic safety assessment (PSA), and verification and validation (V and V) of the program was performed. The main input data of the MACCS II code are meteorological data, population distribution data and source term data. However, it requires lots of time and efforts to generate the main input data for an offsite consequence analysis using the MACCS II code. For example, the meteorological data are collected from each nuclear power site in real time, but the formats of the raw data collected are different from each other as a site. To reduce efforts and time for risk assessments, the web-based OSCAP has an automatic processing module which converts the format of the raw data collected from each site in Korea to the input data format of the MACCS II code. The program also provides an automatic function of converting the latest population data from Statistics Korea, the National Statistical Office, to the population distribution input data format of the MACCS II code. In case of the source term data, the program includes the release fraction of each source term category resulting from Modular Accident Analysis Program (MAAP) code analysis and the core inventory data from ORIGEN code analysis. These analysis results of each plant in Korea are stored in a database module of the web-based OSCAP, so a
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
Deterministic sensitivity and uncertainty methodology for best estimate system codes applied in nuclear technology

International Nuclear Information System (INIS)

Petruzzi, A.; D'Auria, F.; Cacuci, D.G.

2009-01-01

Nuclear Power Plant (NPP) technology has been developed based on the traditional defense in depth philosophy supported by deterministic and overly conservative methods for safety analysis. In the 1970s [1], conservative hypotheses were introduced for safety analyses to address existing uncertainties. Since then, intensive thermal-hydraulic experimental research has resulted in a considerable increase in knowledge and consequently in the development of best-estimate codes able to provide more realistic information about the physical behaviour and to identify the most relevant safety issues allowing the evaluation of the existing actual margins between the results of the calculations and the acceptance criteria. However, the best-estimate calculation results from complex thermal-hydraulic system codes (like Relap5, Cathare, Athlet, Trace, etc..) are affected by unavoidable approximations that are un-predictable without the use of computational tools that account for the various sources of uncertainty. Therefore the use of best-estimate codes (BE) within the reactor technology, either for design or safety purposes, implies understanding and accepting the limitations and the deficiencies of those codes. Taking into consideration the above framework, a comprehensive approach for utilizing quantified uncertainties arising from Integral Test Facilities (ITFs, [2]) and Separate Effect Test Facilities (SETFs, [3]) in the process of calibrating complex computer models for the application to NPP transient scenarios has been developed. The methodology proposed is capable of accommodating multiple SETFs and ITFs to learn as much as possible about uncertain parameters, allowing for the improvement of the computer model predictions based on the available experimental evidences. The proposed methodology constitutes a major step forward with respect to the generally used expert judgment and statistical methods as it permits a) to establish the uncertainties of any parameter
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
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)
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)
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)
Probabilistic accident consequence uncertainty analysis -- Late health effects uncertain assessment. Volume 2: Appendices

Energy Technology Data Exchange (ETDEWEB)

Little, M.P.; Muirhead, C.R. [National Radiological Protection Board (United Kingdom); Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Harper, F.T. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States)

1997-12-01

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA late health effects models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the expert panel on late health effects, (4) short biographies of the experts, and (5) the aggregated results of their responses.
Probabilistic accident consequence uncertainty analysis -- Early health effects uncertainty assessment. Volume 2: Appendices

Energy Technology Data Exchange (ETDEWEB)

Haskin, F.E. [Univ. of New Mexico, Albuquerque, NM (United States); Harper, F.T. [Sandia National Labs., Albuquerque, NM (United States); Goossens, L.H.J.; Kraan, B.C.P. [Delft Univ. of Technology (Netherlands)

1997-12-01

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA early health effects models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the panel on early health effects, (4) short biographies of the experts, and (5) the aggregated results of their responses.
Probabilistic accident consequence uncertainty analysis -- Uncertainty assessment for internal dosimetry. Volume 2: Appendices

Energy Technology Data Exchange (ETDEWEB)

Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Harrison, J.D. [National Radiological Protection Board (United Kingdom); Harper, F.T. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States)

1998-04-01

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA internal dosimetry models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the panel on internal dosimetry, (4) short biographies of the experts, and (5) the aggregated results of their responses.
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
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.

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
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.
Modelling fog in probabilistic consequence assessment

International Nuclear Information System (INIS)

Underwood, B.Y.

1993-02-01

Earlier work examined the potential influence of foggy weather conditions on the probabilistic assessment of the consequences of accidental releases of radioactive material to the atmosphere (PCA), in particular the impact of a fraction of the released aerosol becoming incorporated into droplets. A major uncertainty emerging from the initial scoping study concerned estimation of the fraction of the released material that would be taken up into droplets. An objective is to construct a method for handling in a PCA context the effect of fog on deposition, basing the method on the experience gained from prior investigations. There are two aspects to explicitly including the effect of fog in PCA: estimating the probability of occurrence of various types of foggy condition and calculating the impact on the conventional end-points of consequence assessment. For the first, a brief outline is given of the use of meteorological data by PCA computer codes, followed by a discussion of some routinely-recorded meteorological parameters that are pertinent to fog, such as the presentweather code and horizontal visibility. Four stylized scenarios are defined to cover a wide range of situations in which particle growth by uptake of water may have an important impact on deposition. A description is then given of the way in which routine meteorological data could be used to flag the presence of each of these conditions in the meteorological data file used by the PCA code. The approach developed to calculate the impact on deposition is pitched at a level of complexity appropriate to the PCA context and reflects the physical constraints of the system and accounts for the specific characteristics of the released aerosol. (Author)
Modeling atmospheric dispersion for reactor accident consequence evaluation

International Nuclear Information System (INIS)

Alpert, D.J.; Gudiksen, P.H.; Woodard, K.

1982-01-01

Atmospheric dispersion models are a central part of computer codes for the evaluation of potential reactor accident consequences. A variety of ways of treating to varying degrees the many physical processes that can have an impact on the predicted consequences exists. The currently available models are reviewed and their capabilities and limitations, as applied to reactor accident consequence analyses, are discussed
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.
Cracking the code: the accuracy of coding shoulder procedures and the repercussions.

Science.gov (United States)

Clement, N D; Murray, I R; Nie, Y X; McBirnie, J M

2013-05-01

Coding of patients' diagnosis and surgical procedures is subject to error levels of up to 40% with consequences on distribution of resources and financial recompense. Our aim was to explore and address reasons behind coding errors of shoulder diagnosis and surgical procedures and to evaluate a potential solution. A retrospective review of 100 patients who had undergone surgery was carried out. Coding errors were identified and the reasons explored. A coding proforma was designed to address these errors and was prospectively evaluated for 100 patients. The financial implications were also considered. Retrospective analysis revealed the correct primary diagnosis was assigned in 54 patients (54%) had an entirely correct diagnosis, and only 7 (7%) patients had a correct procedure code assigned. Coders identified indistinct clinical notes and poor clarity of procedure codes as reasons for errors. The proforma was significantly more likely to assign the correct diagnosis (odds ratio 18.2, p code (odds ratio 310.0, p coding department. High error levels for coding are due to misinterpretation of notes and ambiguity of procedure codes. This can be addressed by allowing surgeons to assign the diagnosis and procedure using a simplified list that is passed directly to coding.
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.
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)
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.
Classification Systems, their Digitization and Consequences for Data-Driven Decision Making

DEFF Research Database (Denmark)

Stein, Mari-Klara; Newell, Sue; Galliers, Robert D.

2013-01-01

Classification systems are foundational in many standardized software tools. This digitization of classification systems gives them a new ‘materiality’ that, jointly with the social practices of information producers/consumers, has significant consequences on the representational quality of such ...... and the foundational role of representational quality in understanding the success and consequences of data-driven decision-making.......-narration and meta-narration), and three different information production/consumption situations. We contribute to the relational theorization of representational quality and extend classification systems research by drawing explicit attention to the importance of ‘materialization’ of classification systems...
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
Probabilistic accident consequence uncertainty analysis: Dispersion and deposition uncertainty assessment, main report

International Nuclear Information System (INIS)

Harper, F.T.; Young, M.L.; Miller, L.A.; Hora, S.C.; Lui, C.H.; Goossens, L.H.J.; Cooke, R.M.; Paesler-Sauer, J.; Helton, J.C.

1995-01-01

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the risks presented by nuclear installations based on postulated frequencies and magnitudes of potential accidents. In 1991, the US Nuclear Regulatory Commission (NRC) and the Commission of the European Communities (CEC) began a joint uncertainty analysis of the two codes. The ultimate objective of the joint effort was to develop credible and traceable uncertainty distributions for the input variables of the codes. Expert elicitation was identified as the best technology available for developing a library of uncertainty distributions for the selected consequence parameters. The study was formulated jointly and was limited to the current code models and to physical quantities that could be measured in experiments. Experts developed their distributions independently. To validate the distributions generated for the wet deposition input variables, samples were taken from these distributions and propagated through the wet deposition code model. Resulting distributions closely replicated the aggregated elicited wet deposition distributions. To validate the distributions generated for the dispersion code input variables, samples from the distributions and propagated through the Gaussian plume model (GPM) implemented in the MACCS and COSYMA codes. Project teams from the NRC and CEC cooperated successfully to develop and implement a unified process for the elaboration of uncertainty distributions on consequence code input parameters. Formal expert judgment elicitation proved valuable for synthesizing the best available information. Distributions on measurable atmospheric dispersion and deposition parameters were successfully elicited from experts involved in the many phenomenological areas of consequence analysis. This volume is the first of a three-volume document describing the project
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
Validation of thermal hydraulic computer codes for advanced light water reactor

International Nuclear Information System (INIS)

Macek, J.

2001-01-01

The Czech Republic operates 4 WWER-440 units, two WWER-1000 units are being finalised (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppressure system are modelled with RALOC and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems. An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. The paper provides a concise information on these activities of the NRI and its Thermal-hydraulics Department. A detailed example of the system code validation and the consequent utilisation of the results for a real NPP purposes is included. (author)
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
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.
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)
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
The consequences of multiplexing and limited view angle in coded-aperture imaging

International Nuclear Information System (INIS)

Smith, W.E.; Barrett, H.H.; Paxman, R.G.

1984-01-01

Coded-aperture imaging (CAI) is a method for reconstructing distributions of radionuclide tracers that offers advantages over ECT and PET; namely, many views can be taken simultaneously without detector motion, and large numbers of photons are utilized since collimators are not required. However, because of this type of data acquisition, the coded image suffers from multiplexing; i.e., more than one object point may be mapped to each detector in the coded image. To investigate the dependence of the reconstruction on multiplexing, the authors reconstruct a simulated two-dimensional circular object from multiplexed one-dimensional coded-image data, then perform the reconstruction from un-multiplexed data. Each of these reconstructions are produced both from noise-free and noisy simulated data. To investigate the dependence on view angle, the authors reconstruct two simulated three-dimensional objects; a spherical phantom, and a series of point-like objects arranged nearly in a plane. Each of these reconstructions are from multiplexed two-dimensional coded-image data, first using two orthogonal views, and then a single viewing direction. The two-dimensional reconstructions demonstrate that, in the noise-free case, the multiplexing of the data does not seriously affect the reconstruction equality and that in the noisy-data case, the multiplexing helps, due to the fact that more photons are collected. Also, for point-like objects confined to a near-planar region of space, the authors show that restricted views can give satisfactory results, but that, for a large, three-dimensional object, a more complete viewing geometry is required
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.

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)
International assessment of PCA codes

International Nuclear Information System (INIS)

Neymotin, L.; Lui, C.; Glynn, J.; Archarya, S.

1993-11-01

Over the past three years (1991-1993), an extensive international exercise for intercomparison of a group of six Probabilistic Consequence Assessment (PCA) codes was undertaken. The exercise was jointly sponsored by the Commission of European Communities (CEC) and OECD Nuclear Energy Agency. This exercise was a logical continuation of a similar effort undertaken by OECD/NEA/CSNI in 1979-1981. The PCA codes are currently used by different countries for predicting radiological health and economic consequences of severe accidents at nuclear power plants (and certain types of non-reactor nuclear facilities) resulting in releases of radioactive materials into the atmosphere. The codes participating in the exercise were: ARANO (Finland), CONDOR (UK), COSYMA (CEC), LENA (Sweden), MACCS (USA), and OSCAAR (Japan). In parallel with this inter-code comparison effort, two separate groups performed a similar set of calculations using two of the participating codes, MACCS and COSYMA. Results of the intercode and inter-MACCS comparisons are presented in this paper. The MACCS group included four participants: GREECE: Institute of Nuclear Technology and Radiation Protection, NCSR Demokritos; ITALY: ENEL, ENEA/DISP, and ENEA/NUC-RIN; SPAIN: Universidad Politecnica de Madrid (UPM) and Consejo de Seguridad Nuclear; USA: Brookhaven National Laboratory, US NRC and DOE
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
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.
NACOM - a code for sodium spray fire analysis

International Nuclear Information System (INIS)

Rao, P.M.; Kannan, S.E.

2002-01-01

Full text: In liquid metal fast breeder reactors (LMFBR), leakage of sodium can result in a spray fire. Because of higher burning rates in droplet form combustion of sodium in spray fire, thermal consequences are more severe than that in a sodium pool fire. The code NACOM was developed for the analysis of sodium spray fires in LMFBRs facilities. The code uses the validated model for estimating the falling droplet burning rates in pre-ignition and vapour phase combustion stages. It uses a distribution system to generate the droplet groups of different diameters that represent the spray. The code requires about 20 input parameters like sodium leak rates, sodium temperature, initial cell conditions like oxygen concentration, temperature and dimensions. NACOM is a validated code based on experiments with sodium inventory up to 650 kg in 0 to 21 % O 2 atmospheres. The paper brings out the salient features of the code along with the sensitivity analysis of the main input parameters like spray volume mean diameter, oxygen concentration etc. based on the results obtained. The limitations of the code and the confidence margins applicable to results obtained are also brought out
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.
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
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.
Safety Analysis in Large Volume Vacuum Systems Like Tokamak: Experiments and Numerical Simulation to Analyze Vacuum Ruptures Consequences

Directory of Open Access Journals (Sweden)

A. Malizia

2014-01-01

Full Text Available The large volume vacuum systems are used in many industrial operations and research laboratories. Accidents in these systems should have a relevant economical and safety impact. A loss of vacuum accident (LOVA due to a failure of the main vacuum vessel can result in a fast pressurization of the vessel and consequent mobilization dispersion of hazardous internal material through the braches. It is clear that the influence of flow fields, consequence of accidents like LOVA, on dust resuspension is a key safety issue. In order to develop this analysis an experimental facility is been developed: STARDUST. This last facility has been used to improve the knowledge about LOVA to replicate a condition more similar to appropriate operative condition like to kamaks. By the experimental data the boundary conditions have been extrapolated to give the proper input for the 2D thermofluid-dynamics numerical simulations, developed by the commercial CFD numerical code. The benchmark of numerical simulation results with the experimental ones has been used to validate and tune the 2D thermofluid-dynamics numerical model that has been developed by the authors to replicate the LOVA conditions inside STARDUST. In present work, the facility, materials, numerical model, and relevant results will be presented.
RELAP5/MOD3 code manual: Summaries and reviews of independent code assessment reports. Volume 7, Revision 1

International Nuclear Information System (INIS)

Moore, R.L.; Sloan, S.M.; Schultz, R.R.; Wilson, G.E.

1996-10-01

Summaries of RELAP5/MOD3 code assessments, a listing of the assessment matrix, and a chronology of the various versions of the code are given. Results from these code assessments have been used to formulate a compilation of some of the strengths and weaknesses of the code. These results are documented in the report. Volume 7 was designed to be updated periodically and to include the results of the latest code assessments as they become available. Consequently, users of Volume 7 should ensure that they have the latest revision available
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.
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
[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.
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)
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.
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.
Validation of containment thermal hydraulic computer codes for VVER reactor

Energy Technology Data Exchange (ETDEWEB)

Jiri Macek; Lubomir Denk [Nuclear Research Institute Rez plc Thermal-Hydraulic Analyses Department CZ 250 68 Husinec-Rez (Czech Republic)

2005-07-01

Full text of publication follows: The Czech Republic operates 4 VVER-440 units, two VVER-1000 units are being finalized (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppression system are modelled with COCOSYS and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems.An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. One of the important components of the VVER 440/213 NPP is its containment with pressure suppression system (bubble condenser). For safety analyses of this system, computer codes of the type MELCOR and COCOSYS are used in the Czech Republic. These codes were developed for containments of classic PWRs or BWRs. In order to apply these codes for VVER 440 systems, their validation on experimental facilities must be performed.The paper provides concise information on these activities of the NRI and its Thermal-hydraulics Department. The containment system of the VVER 440/213, its functions and approaches to solution of its safety is described with definition of acceptance criteria. A detailed example of the containment code validation on EREC Test facility (LOCA and MSLB) and the consequent utilisation of the results for a real NPP purposes is included. An approach to
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.
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.)
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.

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
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
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).
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)
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.
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.
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
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.
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.
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.
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.
The Vulnerability Assessment Code for Physical Protection System

International Nuclear Information System (INIS)

Jang, Sung Soon; Yoo, Ho Sik

2007-01-01

To neutralize the increasing terror threats, nuclear facilities have strong physical protection system (PPS). PPS includes detectors, door locks, fences, regular guard patrols, and a hot line to a nearest military force. To design an efficient PPS and to fully operate it, vulnerability assessment process is required. Evaluating PPS of a nuclear facility is complicate process and, hence, several assessment codes have been developed. The estimation of adversary sequence interruption (EASI) code analyzes vulnerability along a single intrusion path. To evaluate many paths to a valuable asset in an actual facility, the systematic analysis of vulnerability to intrusion (SAVI) code was developed. KAERI improved SAVI and made the Korean analysis of vulnerability to intrusion (KAVI) code. Existing codes (SAVI and KAVI) have limitations in representing the distance of a facility because they use the simplified model of a PPS called adversary sequence diagram. In adversary sequence diagram the position of doors, sensors and fences is described just as the locating area. Thus, the distance between elements is inaccurate and we cannot reflect the range effect of sensors. In this abstract, we suggest accurate and intuitive vulnerability assessment based on raster map modeling of PPS. The raster map of PPS accurately represents the relative position of elements and, thus, the range effect of sensor can be easily incorporable. Most importantly, the raster map is easy to understand
Modeling of a planning system in radiotherapy and Nuclear Medicine using the MCNP6 code

International Nuclear Information System (INIS)

Massicano, Felipe

2015-01-01

Cancer therapy has many branches and one of them is the use of radiation sources as treatment leading method. Radiotherapy and nuclear medicine are examples of these treatment types. For using the ionization radiation as main tool for the therapy, there is the need of crafting many treatment simulation in order to maximum the tumoral tissue dose without surpass the dose limit in health tissue surrounding. Treatment planning systems (TPS) are systems which have the purpose of simulating these therapy types. Nuclear medicine and radiotherapy have many distinct features linked to the therapy mode and consequently they have different TPS destined for each. The radiotherapy TPS is more developed than the nuclear medicine TPS and by that reason the development of a TPS that was similar to the radiotherapy TPS, but enough generic for include other therapy types, it will contribute with significant advances in nuclear medicine and in others therapy types with radiation. Based on this, the goal of work was to model a TPS that utilizes the Monte Carlo N-Particle Transport code (MCNP6) in order to simulate radiotherapy therapy, nuclear medicine therapy and with potential for simulating other therapy types too. The result of this work was the creation of a Framework in Java language, object oriented, named IBMC which will assist in the development of new TPS with MCNP6 code. The IBMC allowed to develop rapidly and easily TPS for radiotherapy and nuclear medicine and the results were validated with systems already consolidated. The IBMC showed high potential for developing TPS by new therapy types. (author)
Simulation of LOCA power transients of CANDU6 by SCAN/RELAP-CANDU coupled code system

International Nuclear Information System (INIS)

Hong, In Seob; Kim, Chang Hyo; Hwang, Su Hyun; Kim, Man Woong; Chung, Bub Dong

2004-01-01

As can be seen in the standalone application of RELAP-CANDU for LOCA analysis of CANDU-PHWR, the system thermal-hydraulic code alone cannot predict the transient behavior accurately. Therefore, best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. The purpose of this research is to develop and test a coupled neutronics and thermal-hydraulics analysis code, SCAN (SNU CANDU-PHWR Neutronics) and RELAP-CANDU, for transient analysis of CANDU-PHWR's. For this purpose, a spatial kinetics calculation module of SCAN, a 3-D CANDU-PHWR neutronics design and analysis code, is dynamically coupled with RELAP-CANDU, the system thermal-hydraulic code for CANDU-PHWR. The performance of the coupled code system is examined by simulation of reactor power transients caused by a hypothetical Loss Of Coolant Accident (LOCA) in Wolsong units, which involves the insertion of positive void reactivity into the core in the course of transients. Specifically, a 40% Reactor Inlet Header (RIH) break LOCA was assumed for the test of the SCAN/RELAP-CANDU coupled code system analysis
Feasibility study on the rod ejection accident analysis with RETRAN-MASTER code system

International Nuclear Information System (INIS)

Kim, Y. H.; Lee, C. S.

2003-01-01

KEPRI has been developed the in-house methodology for non-LOCA safety analyses based on the codes and methodologies of vendors and EPRI. Using the methodology, the rod ejection accident, which is classified into the generic accident analysis category of reactivity insertion accident in primary system, has been analyzed with RETRAN-MASTER code system. And the feasibility of the coupled code system has been verified by the review of the results. Furthermore, to assess the important parameters to the accident, the sensitivity analyses have been carried out over some parameters
Architectural and Algorithmic Requirements for a Next-Generation System Analysis Code

Energy Technology Data Exchange (ETDEWEB)

V.A. Mousseau

2010-05-01

This document presents high-level architectural and system requirements for a next-generation system analysis code (NGSAC) to support reactor safety decision-making by plant operators and others, especially in the context of light water reactor plant life extension. The capabilities of NGSAC will be different from those of current-generation codes, not only because computers have evolved significantly in the generations since the current paradigm was first implemented, but because the decision-making processes that need the support of next-generation codes are very different from the decision-making processes that drove the licensing and design of the current fleet of commercial nuclear power reactors. The implications of these newer decision-making processes for NGSAC requirements are discussed, and resulting top-level goals for the NGSAC are formulated. From these goals, the general architectural and system requirements for the NGSAC are derived.
Systemization of burnup sensitivity analysis code. 2

International Nuclear Information System (INIS)

Tatsumi, Masahiro; Hyoudou, Hideaki

2005-02-01

Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint 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 the development of adjusted nuclear library using the cross-section adjustment method, in which the results of criticality 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 '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, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users 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 functions in the existing large system. It is not sufficient to unify each computational component for the following reasons; the 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
Coding response to a case-mix measurement system based on multiple diagnoses.

Science.gov (United States)

Preyra, Colin

2004-08-01

To examine the hospital coding response to a payment model using a case-mix measurement system based on multiple diagnoses and the resulting impact on a hospital cost model. Financial, clinical, and supplementary data for all Ontario short stay hospitals from years 1997 to 2002. Disaggregated trends in hospital case-mix growth are examined for five years following the adoption of an inpatient classification system making extensive use of combinations of secondary diagnoses. Hospital case mix is decomposed into base and complexity components. The longitudinal effects of coding variation on a standard hospital payment model are examined in terms of payment accuracy and impact on adjustment factors. Introduction of the refined case-mix system provided incentives for hospitals to increase reporting of secondary diagnoses and resulted in growth in highest complexity cases that were not matched by increased resource use over time. Despite a pronounced coding response on the part of hospitals, the increase in measured complexity and case mix did not reduce the unexplained variation in hospital unit cost nor did it reduce the reliance on the teaching adjustment factor, a potential proxy for case mix. The main implication was changes in the size and distribution of predicted hospital operating costs. Jurisdictions introducing extensive refinements to standard diagnostic related group (DRG)-type payment systems should consider the effects of induced changes to hospital coding practices. Assessing model performance should include analysis of the robustness of classification systems to hospital-level variation in coding practices. Unanticipated coding effects imply that case-mix models hypothesized to perform well ex ante may not meet expectations ex post.
Coding Response to a Case-Mix Measurement System Based on Multiple Diagnoses

Science.gov (United States)

Preyra, Colin

2004-01-01

Objective To examine the hospital coding response to a payment model using a case-mix measurement system based on multiple diagnoses and the resulting impact on a hospital cost model. Data Sources Financial, clinical, and supplementary data for all Ontario short stay hospitals from years 1997 to 2002. Study Design Disaggregated trends in hospital case-mix growth are examined for five years following the adoption of an inpatient classification system making extensive use of combinations of secondary diagnoses. Hospital case mix is decomposed into base and complexity components. The longitudinal effects of coding variation on a standard hospital payment model are examined in terms of payment accuracy and impact on adjustment factors. Principal Findings Introduction of the refined case-mix system provided incentives for hospitals to increase reporting of secondary diagnoses and resulted in growth in highest complexity cases that were not matched by increased resource use over time. Despite a pronounced coding response on the part of hospitals, the increase in measured complexity and case mix did not reduce the unexplained variation in hospital unit cost nor did it reduce the reliance on the teaching adjustment factor, a potential proxy for case mix. The main implication was changes in the size and distribution of predicted hospital operating costs. Conclusions Jurisdictions introducing extensive refinements to standard diagnostic related group (DRG)-type payment systems should consider the effects of induced changes to hospital coding practices. Assessing model performance should include analysis of the robustness of classification systems to hospital-level variation in coding practices. Unanticipated coding effects imply that case-mix models hypothesized to perform well ex ante may not meet expectations ex post. PMID:15230940
Systems guide to MCNP (Monte Carlo Neutron and Photon Transport Code)

International Nuclear Information System (INIS)

Kirk, B.L.; West, J.T.

1984-06-01

The subject of this report is the implementation of the Los Alamos National Laboratory Monte Carlo Neutron and Photon Transport Code - Version 3 (MCNP) on the different types of computer systems, especially the IBM MVS system. The report supplements the documentation of the RSIC computer code package CCC-200/MCNP. Details of the procedure to follow in executing MCNP on the IBM computers, either in batch mode or interactive mode, are provided

Computer-aided software understanding systems to enhance confidence of scientific codes

International Nuclear Information System (INIS)

Sheng, G.; Oeren, T.I.

1991-01-01

A unique characteristic of nuclear waste disposal is the very long time span over which the combined engineered and natural containment system must remain effective: hundreds of thousands of years. Since there is no precedent in human history for such an endeavour, simulation with the use of computers is the only means we have of forecasting possible future outcomes quantitatively. The need for reliable models and software to make such forecasts so far into the future is obvious. One of the critical elements necessary to ensure reliability is the degree of reviewability of the computer program. Among others, there are two very important reasons for this. Firstly, if there is to be any chance at all of validating the conceptual models as implemented by the computer code, peer reviewers must be able to see and understand what the program is doing. It is all but impossible to achieve this understanding by just looking at the code due to possible unfamiliarity with the language and often due as well to the length and complexity of the code. Secondly, a thorough understanding of the code is also necessary to carry out code maintenance activities which include among others, error detection, error correction and code modification for purposes of enhancing its performance, functionality or to adapt it to a changed environment. The emerging concepts of computer-aided software understanding and reverse engineering can answer precisely these needs. This paper will discuss the role they can play in enhancing the confidence one has on computer codes and several examples will be provided. Finally a brief discussion of combining state-of-art forward engineering systems with reverse engineering systems will show how powerfully they can contribute to the overall quality assurance of a computer program. (13 refs., 7 figs.)
Code conversion for system design and safety analysis of NSSS

Energy Technology Data Exchange (ETDEWEB)

Lee, Hae Cho; Kim, Young Tae; Choi, Young Gil; Kim, Hee Kyung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-01-01

This report describes overall project works related to conversion, installation and validation of computer codes which are used in NSSS design and safety analysis of nuclear power plants. Domain/os computer codes for system safety analysis are installed and validated on Apollo DN10000, and then Apollo version are converted and installed again on HP9000/700 series with appropriate validation. Also, COOLII and COAST which are cyber version computer codes are converted into versions of Apollo DN10000 and HP9000/700, and installed with validation. This report details whole processes of work involved in the computer code conversion and installation, as well as software verification and validation results which are attached to this report. 12 refs., 8 figs. (author)
A code system to generate multigroup cross-sections using basic data

International Nuclear Information System (INIS)

Garg, S.B.; Kumar, Ashok

1978-01-01

For the neutronic studies of nuclear reactors, multigroup cross-sections derived from the basic energy point data are needed. In order to carry out the design based studies, these cross-sections should also incorporate the temperature and fuel concentration effects. To meet these requirements, a code system comprising of RESRES, UNRES, FIGERO, INSCAT, FUNMO, AVER1 and BGPONE codes has been adopted. The function of each of these codes is discussed. (author)
CODE SWITCHING AND THE DEVELOPMENT OF LINGUISTIC SYSTEM OF SIMULTANEOUS BILINGUAL CHILDREN

Directory of Open Access Journals (Sweden)

Leni Amelia Suek

2017-11-01

Full Text Available Code switching and code mixing are the phenomena commonly seen done by a bilingual. This behavior is influenced by several aspects such as the linguistic system, sociolinguistics, pragmatics, and language competence of the bilingual. If children are able to distinguish two different languages since early age, they will be considered simultaneous bilinguals. They show that they develop multiple, rather than single, linguistic systems. However, it was understood that code switching and code mixing were due to the failure in using proper words, language features, and sociolinguistic competence. Yet, recent studies have shown that bilingual children are able to use both languages proficiently with no signs of confusion or failure in language use. This ability also does not hinder their cognitive development.
Development of non-orthogonal and 2-dimensional numerical code TFC2D-BFC for fluid flow

International Nuclear Information System (INIS)

Park, Ju Yeop; In, Wang Kee; Chun, Tae Hyun; Oh, Dong Seok

2000-09-01

The development of algorithm for three dimensional non-orthogonal coordinate system has been made. The algorithm adopts a non-staggered grid system, Cartesian velocity components for independent variables of momentum equations and a SIMPLER algorithm for a pressure correction equation. Except the pressure correction method, the selected grid system and the selected independent variables for momentum equations have been widely used in a commercial code. It is well known that the SIMPLER is superior to the SIMPLE algorithm in the view of convergence rate. Using this algorithm, a two dimensional non-orthogonal numerical code has been completed. The code adopts a structured single square block in a computational domain with a uniform mesh interval. Consequently, any solid body existing in a flow field can be implemented in the numerical code through a blocked-off method which was devised by Patankar
A QR code identification technology in package auto-sorting system

Science.gov (United States)

di, Yi-Juan; Shi, Jian-Ping; Mao, Guo-Yong

2017-07-01

Traditional manual sorting operation is not suitable for the development of Chinese logistics. For better sorting packages, a QR code recognition technology is proposed to identify the QR code label on the packages in package auto-sorting system. The experimental results compared with other algorithms in literatures demonstrate that the proposed method is valid and its performance is superior to other algorithms.
The SWAN-SCALE code for the optimization of critical systems

International Nuclear Information System (INIS)

Greenspan, E.; Karni, Y.; Regev, D.; Petrie, L.M.

1999-01-01

The SWAN optimization code was recently developed to identify the maximum value of k eff for a given mass of fissile material when in combination with other specified materials. The optimization process is iterative; in each iteration SWAN varies the zone-dependent concentration of the system constituents. This change is guided by the equal volume replacement effectiveness functions (EVREF) that SWAN generates using first-order perturbation theory. Previously, SWAN did not have provisions to account for the effect of the composition changes on neutron cross-section resonance self-shielding; it used the cross sections corresponding to the initial system composition. In support of the US Department of Energy Nuclear Criticality Safety Program, the authors recently removed the limitation on resonance self-shielding by coupling SWAN with the SCALE code package. The purpose of this paper is to briefly describe the resulting SWAN-SCALE code and to illustrate the effect that neutron cross-section self-shielding could have on the maximum k eff and on the corresponding system composition
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
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
Manual for IRS Coding. Joint IAEA/NEA International Reporting System for Operating Experience

International Nuclear Information System (INIS)

2011-01-01

The International Reporting System for Operating Experience (IRS) is jointly operated by the International Atomic Energy Agency (IAEA) and the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (OECD/NEA). In early 2010, the IAEA and OECD/NEA jointly issued the IRS Guidelines, which described the reporting system and process and gave users the necessary elements to enable them to produce IRS reports to a high standard of quality while retaining the effectiveness of the system expected by all Member States operating nuclear power plants. The purpose of the present Manual for IRS Coding is to provide supplementary guidance specifically on the coding element of IRS reports to ensure uniform coding of events that are reported through IRS. This Coding Manual does not supersede the IRS Guidelines, but rather, supports users and preparers in achieving a consistent and high level of quality in their IRS reports. Consistency and high quality in the IRS reports allow stakeholders to search and retrieve specific event information with ease. In addition, well-structured reports also enhance the efficient management of the IRS database. This Coding Manual will give specific guidance on the application of each section of the IRS codes, with examples where necessary, of when and how these codes are to be applied. As this reporting system is owned by the Member States, this manual has been developed and approved by the IRS National Coordinators with the assistance of the IAEA and NEA secretariats
CORESAFE: A Formal Approach against Code Replacement Attacks on Cyber Physical Systems

Science.gov (United States)

2018-04-19

AFRL-AFOSR-JP-TR-2018-0035 CORESAFE:A Formal Approach against Code Replacement Attacks on Cyber Physical Systems Sandeep Shukla INDIAN INSTITUTE OF...Formal Approach against Code Replacement Attacks on Cyber Physical Systems 5a. CONTRACT NUMBER 5b. GRANT NUMBER FA2386-16-1-4099 5c. PROGRAM ELEMENT...SUPPLEMENTARY NOTES 14. ABSTRACT Industrial Control Systems (ICS) used in manufacturing, power generators and other critical infrastructure monitoring and
Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

International Nuclear Information System (INIS)

Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E.; Tills, J.

1997-12-01

The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions
Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

Energy Technology Data Exchange (ETDEWEB)

Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E. [Sandia National Labs., Albuquerque, NM (United States); Tills, J. [J. Tills and Associates, Inc., Sandia Park, NM (United States)

1997-12-01

The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.
Fra massemedier til mediesystem - om kodediskussionen i systemteoretisk medieforskning [From mass media to media system - code discussions in systems theoretical media research

Directory of Open Access Journals (Sweden)

Mikkel Fugl Eskjær

2010-11-01

Full Text Available Systems theoretical media research raises the question whether the mass media constitute a unified institution, or whether the media, due to their internal differences, should be considered individually and independent of each other. By inscribing the media in a general social theory, systems theory conceptualises the media as an autonomous functional system. This intention is most clearly illustrated by the efforts to identify a shared code for the entire media system. Based on the media theory of Niklas Luhmann, this paper offers a critical presentation of the code discussion within systems theoretical media research. The first part of the paper briefly introduces the systems theoretical notion of a code as well as Luhmann’s definition of the media system as organised and regulated by the code of information. The second part presents a number of alternative suggestions and definitions of the media system’s code, which both indicate the scope of systems theoretical media research, but also point to some of the limitations in the systems theoretical approach. In the last part, the paper takes a critical look at the systems theoretical code discussion by arguing that a too narrow focus on code definitions is blocking a more productive investigation of the conditions, evolution, and autonomy of the media system.
Results from the Coded Aperture Neutron Imaging System (CANIS)

International Nuclear Information System (INIS)

Brubaker, Erik; Steele, John T.; Brennan, James S.; Hilton, Nathan R.; 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.
Burnup calculation code system COMRAD96

International Nuclear Information System (INIS)

Suyama, Kenya; Masukawa, Fumihiro; Ido, Masaru; Enomoto, Masaki; Takyu, Shuiti; Hara, Toshiharu.

1997-06-01

COMRAD was one of the burnup code system developed by JAERI. COMRAD96 is a transfered version of COMRAD to Engineering Work Station. It is divided to several functional modules, 'Cross Section Treatment', 'Generation and Depletion Calculation', and 'Post Process'. It enables us to analyze a burnup problem considering a change of neutron spectrum using UNITBURN. Also it can display the γ Spectrum on a terminal. This report is the general description and user's manual of COMRAD96. (author)
Power Optimization of Wireless Media Systems With Space-Time Block Codes

OpenAIRE

Yousefi'zadeh, Homayoun; Jafarkhani, Hamid; Moshfeghi, Mehran

2004-01-01

We present analytical and numerical solutions to the problem of power control in wireless media systems with multiple antennas. We formulate a set of optimization problems aimed at minimizing total power consumption of wireless media systems subject to a given level of QoS and an available bit rate. Our formulation takes in to consideration the power consumption related to source coding, channel coding, and transmission of multiple-transmit antennas. In our study, we consider Gauss-Markov and...
SCALE: A modular code system for performing standardized computer analyses for licensing evaluation

International Nuclear Information System (INIS)

1997-03-01

This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This manual covers an array of modules written for the SCALE package, consisting of drivers, system libraries, cross section and materials properties libraries, input/output routines, storage modules, and help files
SCALE: A modular code system for performing standardized computer analyses for licensing evaluation

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-03-01

This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This manual covers an array of modules written for the SCALE package, consisting of drivers, system libraries, cross section and materials properties libraries, input/output routines, storage modules, and help files.
An Adaptation of the HELIOS/MASTER Code System to the Analysis of VHTR Cores

International Nuclear Information System (INIS)

Noh, Jae Man; Lee, Hyun Chul; Kim, Kang Seog; Kim, Yong Hee

2006-01-01

KAERI is developing a new computer code system for an analysis of VHTR cores based on the existing HELIOS/MASTER code system which was originally developed for a LWR core analysis. In the VHTR reactor physics, there are several unique neutronic characteristics that cannot be handled easily by the conventional computer code system applied for the LWR core analysis. Typical examples of such characteristics are a double heterogeneity problem due to the particulate fuels, the effects of a spectrum shift and a thermal up-scattering due to the graphite moderator, and a strong fuel/reflector interaction, etc. In order to facilitate an easy treatment of such characteristics, we developed some methodologies for the HELIOS/MASTER code system and tested their applicability to the VHTR core analysis

Computer codes and methods for simulating accelerator driven systems

International Nuclear Information System (INIS)

Sartori, E.; Byung Chan Na

2003-01-01

A large set of computer codes and associated data libraries have been developed by nuclear research and industry over the past half century. A large number of them are in the public domain and can be obtained under agreed conditions from different Information Centres. The areas covered comprise: basic nuclear data and models, reactor spectra and cell calculations, static and dynamic reactor analysis, criticality, radiation shielding, dosimetry and material damage, fuel behaviour, safety and hazard analysis, heat conduction and fluid flow in reactor systems, spent fuel and waste management (handling, transportation, and storage), economics of fuel cycles, impact on the environment of nuclear activities etc. These codes and models have been developed mostly for critical systems used for research or power generation and other technological applications. Many of them have not been designed for accelerator driven systems (ADS), but with competent use, they can be used for studying such systems or can form the basis for adapting existing methods to the specific needs of ADS's. The present paper describes the types of methods, codes and associated data available and their role in the applications. It provides Web addresses for facilitating searches for such tools. Some indications are given on the effect of non appropriate or 'blind' use of existing tools to ADS. Reference is made to available experimental data that can be used for validating the methods use. Finally, some international activities linked to the different computational aspects are described briefly. (author)
PRELIMINARY STUDY ON APPLICATION OF MAX PLUS ALGEBRA IN DISTRIBUTED STORAGE SYSTEM THROUGH NETWORK CODING

Directory of Open Access Journals (Sweden)

Agus Maman Abadi

2016-04-01

Full Text Available The increasing need in techniques of storing big data presents a new challenge. One way to address this challenge is the use of distributed storage systems. One strategy that implemented in distributed data storage systems is the use of Erasure Code which applied to network coding. The code used in this technique is based on the algebraic structure which is called as vector space. Some studies have also been carried out to create code that is based on other algebraic structures such as module. In this study, we are going to try to set up a code based on the algebraic structure which is a generalization of the module that is semimodule by utilizing the max operations and sum operations at max plus algebra. The results of this study indicate that the max operation and the addition operation on max plus algebra cannot be used to establish a semimodule code, but by modifying the operation "+" as "min", we get a code based on semimodule. Keywords: code, distributed storage systems, network coding, semimodule, max plus algebra
Bilingual processing of ASL-English code-blends: The consequences of accessing two lexical representations simultaneously

OpenAIRE

Emmorey, Karen; Petrich, Jennifer; Gollan, Tamar H.

2012-01-01

Bilinguals who are fluent in American Sign Language (ASL) and English often produce code-blends - simultaneously articulating a sign and a word while conversing with other ASL-English bilinguals. To investigate the cognitive mechanisms underlying code-blend processing, we compared picture-naming times (Experiment 1) and semantic categorization times (Experiment 2) for code-blends versus ASL signs and English words produced alone. In production, code-blending did not slow lexical retrieval for...
Integration of QR codes into an anesthesia information management system for resident case log management.

Science.gov (United States)

Avidan, Alexander; Weissman, Charles; Levin, Phillip D

2015-04-01

Quick response (QR) codes containing anesthesia syllabus data were introduced into an anesthesia information management system. The code was generated automatically at the conclusion of each case and available for resident case logging using a smartphone or tablet. The goal of this study was to evaluate the use and usability/user-friendliness of such system. Resident case logging practices were assessed prior to introducing the QR codes. QR code use and satisfactions amongst residents was reassessed at three and six months. Before QR code introduction only 12/23 (52.2%) residents maintained a case log. Most of the remaining residents (9/23, 39.1%) expected to receive a case list from the anesthesia information management system database at the end of their residency. At three months and six months 17/26 (65.4%) and 15/25 (60.0%) residents, respectively, were using the QR codes. Satisfaction was rated as very good or good. QR codes for residents' case logging with smartphones or tablets were successfully introduced in an anesthesia information management system and used by most residents. QR codes can be successfully implemented into medical practice to support data transfer. Copyright © 2015 Elsevier Ireland Ltd. All rights reserved.
Field-based tests of geochemical modeling codes usign New Zealand hydrothermal systems

International Nuclear Information System (INIS)

Bruton, C.J.; Glassley, W.E.; Bourcier, W.L.

1994-06-01

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
A Coding System for Analysing a Spoken Text Database.

Science.gov (United States)

Cutting, Joan

1994-01-01

This paper describes a coding system devised to analyze conversations of graduate students in applied linguistics at Edinburgh University. The system was devised to test the hypothesis that as shared knowledge among conversation participants grows, the textual density of in-group members has more cues than that of strangers. The informal…
COMPARISON OF CONSEQUENCE ANALYSIS RESULTS FROM TWO METHODS OF PROCESSING SITE METEOROLOGICAL DATA

International Nuclear Information System (INIS)

, D

2007-01-01

Consequence analysis to support documented safety analysis requires the use of one or more years of representative meteorological data for atmospheric transport and dispersion calculations. At minimum, the needed meteorological data for most atmospheric transport and dispersion models consist of hourly samples of wind speed and atmospheric stability class. Atmospheric stability is inferred from measured and/or observed meteorological data. Several methods exist to convert measured and observed meteorological data into atmospheric stability class data. In this paper, one year of meteorological data from a western Department of Energy (DOE) site is processed to determine atmospheric stability class using two methods. The method that is prescribed by the U.S. Nuclear Regulatory Commission (NRC) for supporting licensing of nuclear power plants makes use of measurements of vertical temperature difference to determine atmospheric stability. Another method that is preferred by the U.S. Environmental Protection Agency (EPA) relies upon measurements of incoming solar radiation, vertical temperature gradient, and wind speed. Consequences are calculated and compared using the two sets of processed meteorological data from these two methods as input data into the MELCOR Accident Consequence Code System 2 (MACCS2) code
Economic aspects and models for building codes

DEFF Research Database (Denmark)

Bonke, Jens; Pedersen, Dan Ove; Johnsen, Kjeld

It is the purpose of this bulletin to present an economic model for estimating the consequence of new or changed building codes. The object is to allow comparative analysis in order to improve the basis for decisions in this field. The model is applied in a case study.......It is the purpose of this bulletin to present an economic model for estimating the consequence of new or changed building codes. The object is to allow comparative analysis in order to improve the basis for decisions in this field. The model is applied in a case study....
Performance analysis of wavelength/spatial coding system with fixed in-phase code matrices in OCDMA network

Science.gov (United States)

Tsai, Cheng-Mu; Liang, Tsair-Chun

2011-12-01

This paper proposes a wavelength/spatial (W/S) coding system with fixed in-phase code (FIPC) matrix in the optical code-division multiple-access (OCDMA) network. A scheme is presented to form the FIPC matrix which is applied to construct the W/S OCDMA network. The encoder/decoder in the W/S OCDMA network is fully able to eliminate the multiple-access-interference (MAI) at the balanced photo-detectors (PD), according to fixed in-phase cross correlation. The phase-induced intensity noise (PIIN) related to the power square is markedly suppressed in the receiver by spreading the received power into each PD while the net signal power is kept the same. Simulation results show that the W/S OCDMA network based on the FIPC matrices cannot only completely remove the MAI but effectively suppress the PIIN to upgrade the network performance.
A Perspective of Energy Codes and Regulations for the Buildings of the Future

Energy Technology Data Exchange (ETDEWEB)

Rosenberg, Michael [Pacific Northwest National Laboratory,2032 Todd Street,Eugene, OR 97405e-mail: michael.rosenberg@pnnl.gov; Jonlin, Duane [Seattle Department ofConstruction and Inspections,P.O. Box 34019,Seattle, WA 98124e-mail: duane.jonlin@seattle.gov; Nadel, Steven [American Council for anEnergy-Efficient Economy,529 14th Street NW #600,Washington, DC 20045e-mail: snadel@aceee.org

2016-10-13

Today’s building energy codes focus on prescriptive requirements for features of buildings that are directly controlled by the design and construction teams and verifiable by municipal inspectors. Although these code requirements have had a significant impact, they fail to influence a large slice of the building energy use pie – including not only miscellaneous plug loads, cooking equipment and commercial/industrial processes, but the maintenance and optimization of the code-mandated systems as well. Currently, code compliance is verified only through the end of construction, and there are no limits or consequences for the actual energy use in an occupied building. In the future, our suite of energy regulations will likely expand to include building efficiency, energy use or carbon emission budgets over their full life cycle. Intelligent building systems, extensive renewable energy, and a transition from fossil fuel to electric heating systems will likely be required to meet ultra-low-energy targets. This paper lays out the authors’ perspectives on how buildings may evolve over the course of the 21st century and the roles that codes and regulations will play in shaping those buildings of the future.
Surface acoustic wave coding for orthogonal frequency coded devices

Science.gov (United States)

Malocha, Donald (Inventor); Kozlovski, Nikolai (Inventor)

2011-01-01

Methods and systems for coding SAW OFC devices to mitigate code collisions in a wireless multi-tag system. Each device producing plural stepped frequencies as an OFC signal with a chip offset delay to increase code diversity. A method for assigning a different OCF to each device includes using a matrix based on the number of OFCs needed and the number chips per code, populating each matrix cell with OFC chip, and assigning the codes from the matrix to the devices. The asynchronous passive multi-tag system includes plural surface acoustic wave devices each producing a different OFC signal having the same number of chips and including a chip offset time delay, an algorithm for assigning OFCs to each device, and a transceiver to transmit an interrogation signal and receive OFC signals in response with minimal code collisions during transmission.
110 °C range athermalization of wavefront coding infrared imaging systems

Science.gov (United States)

Feng, Bin; Shi, Zelin; Chang, Zheng; Liu, Haizheng; Zhao, Yaohong

2017-09-01

110 °C range athermalization is significant but difficult for designing infrared imaging systems. Our wavefront coding athermalized infrared imaging system adopts an optical phase mask with less manufacturing errors and a decoding method based on shrinkage function. The qualitative experiments prove that our wavefront coding athermalized infrared imaging system has three prominent merits: (1) working well over a temperature range of 110 °C; (2) extending the focal depth up to 15.2 times; (3) achieving a decoded image being approximate to its corresponding in-focus infrared image, with a mean structural similarity index (MSSIM) value greater than 0.85.
Strength evaluation code STEP for brittle materials

International Nuclear Information System (INIS)

Ishihara, Masahiro; Futakawa, Masatoshi.

1997-12-01

In a structural design using brittle materials such as graphite and/or ceramics it is necessary to evaluate the strength of component under complex stress condition. The strength of ceramic materials is said to be influenced by the stress distribution. However, in the structural design criteria simplified stress limits had been adopted without taking account of the strength change with the stress distribution. It is, therefore, important to evaluate the strength of component on the basis of the fracture model for brittle material. Consequently, the strength evaluation program, STEP, on a brittle fracture of ceramic materials based on the competing risk theory had been developed. Two different brittle fracture modes, a surface layer fracture mode dominated by surface flaws and an internal fracture mode by internal flaws, are treated in the STEP code in order to evaluate the strength of brittle fracture. The STEP code uses stress calculation results including complex shape of structures analyzed by the generalized FEM stress analysis code, ABAQUS, so as to be possible to evaluate the strength of brittle fracture for the structures having complicate shapes. This code is, therefore, useful to evaluate the structural integrity of arbitrary shapes of components such as core graphite components in the HTTR, heat exchanger components made of ceramics materials etc. This paper describes the basic equations applying to the STEP code, code system with a combination of the STEP and the ABAQUS codes and the result of the verification analysis. (author)
Enhancement of safety analysis reliability for a CANDU-6 reactor using RELAP-CANDU/SCAN coupled code system

International Nuclear Information System (INIS)

Kim, Man Woong; Choi, Yong Seog; Sin, Chul; Kim, Hyun Koon; Kim, Hho Jung; Hwang, Su Hyun; Hong, In Seob; Kim, Chang Hyo

2005-01-01

In LOCA analysis of the CANDU reactor, the system thermal-hydraulic code, RELAP-CANDU, alone cannot predict the transient behavior accurately. Therefore, the best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. To perform on-line calculation of safety analysis for CANDU reactor, a coupled thermal hydraulics-neutronics code system was developed in such a way that the best-estimate thermal-hydraulic system code for CANDU reactor, RELAP-CANDU, is coupled with the full three-dimensional reactor core kinetic code
Finite mixture models for sensitivity analysis of thermal hydraulic codes for passive safety systems analysis

Energy Technology Data Exchange (ETDEWEB)

Di Maio, Francesco, E-mail: francesco.dimaio@polimi.it [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Nicola, Giancarlo [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Zio, Enrico [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Chair on System Science and Energetic Challenge Fondation EDF, Ecole Centrale Paris and Supelec, Paris (France); Yu, Yu [School of Nuclear Science and Engineering, North China Electric Power University, 102206 Beijing (China)

2015-08-15

Highlights: • Uncertainties of TH codes affect the system failure probability quantification. • We present Finite Mixture Models (FMMs) for sensitivity analysis of TH codes. • FMMs approximate the pdf of the output of a TH code with a limited number of simulations. • The approach is tested on a Passive Containment Cooling System of an AP1000 reactor. • The novel approach overcomes the results of a standard variance decomposition method. - Abstract: For safety analysis of Nuclear Power Plants (NPPs), Best Estimate (BE) Thermal Hydraulic (TH) codes are used to predict system response in normal and accidental conditions. The assessment of the uncertainties of TH codes is a critical issue for system failure probability quantification. In this paper, we consider passive safety systems of advanced NPPs and present a novel approach of Sensitivity Analysis (SA). The approach is based on Finite Mixture Models (FMMs) to approximate the probability density function (i.e., the uncertainty) of the output of the passive safety system TH code with a limited number of simulations. We propose a novel Sensitivity Analysis (SA) method for keeping the computational cost low: an Expectation Maximization (EM) algorithm is used to calculate the saliency of the TH code input variables for identifying those that most affect the system functional failure. The novel approach is compared with a standard variance decomposition method on a case study considering a Passive Containment Cooling System (PCCS) of an Advanced Pressurized reactor AP1000.
The radiological assessment system for consequence analysis - RASCAL

International Nuclear Information System (INIS)

Sjoreen, A.L.; Ramsdell, J.V.; Athey, G.F.

1996-01-01

The Radiological Assessment System for Consequence Analysis, Version 2.1 (RASCAL 2.1) has been developed for use during a response to radiological emergencies. The model estimates doses for comparison with U.S. Environmental Protection Agency (EPA) Protective Action Guides (PAGs) and thresholds for acute health effects. RASCAL was designed to be used by U.S. Nuclear Regulatory Commission (NRC) personnel who report to the site of a nuclear accident to conduct an independent evaluation of dose and consequence projections and personnel who conduct training and drills on emergency responses. It allows consideration of the dominant aspects of the source term, transport, dose, and consequences. RASCAL consists of three computational tools: ST-DOSE, FM-DOSE, and DECAY. ST-DOSE computes source term, atmospheric transport, and dose to man from accidental airborne releases of radionuclides. The source-term calculations are appropriate for accidents at U.S. power reactors. FM-DOSE computes doses from environmental concentrations of radionuclides in the air and on the ground. DECAY computes radiological decay and daughter in-growth. RASCAL 2.1 is a DOS application that can be run under Windows 3.1 and 95. RASCAL has been the starting point for other accident consequence models, notably INTERRAS, an international version of RASCAL, and HASCAL, an expansion of RASCAL that will model radiological, biological, and chemical accidents
SRAC: JAERI thermal reactor standard code system for reactor design and analysis

International Nuclear Information System (INIS)

Tsuchihashi, Keichiro; Takano, Hideki; Horikami, Kunihiko; Ishiguro, Yukio; Kaneko, Kunio; Hara, Toshiharu.

1983-01-01

The SRAC (Standard Reactor Analysis Code) is a code system for nuclear reactor analysis and design. It is composed of neutron cross section libraries and auxiliary processing codes, neutron spectrum routines, a variety of transport, 1-, 2- and 3-D diffusion routines, dynamic parameters and cell burn-up routines. By making the best use of the individual code function in the SRAC system, the user can select either the exact method for an accurate estimate of reactor characteristics or the economical method aiming at a shorter computer time, depending on the purpose of study. The user can select cell or core calculation; fixed source or eigenvalue problem; transport (collision probability or Sn) theory or diffusion theory. Moreover, smearing and collapsing of macroscopic cross sections are separately done by the user's selection. And a special attention is paid for double heterogeneity. Various techniques are employed to access the data storage and to optimize the internal data transfer. Benchmark calculations using the SRAC system have been made extensively for the Keff values of various types of critical assemblies (light water, heavy water and graphite moderated systems, and fast reactor systems). The calculated results show good prediction for the experimental Keff values. (author)
Safety analysis code SCTRAN development for SCWR and its application to CGNPC SCWR

International Nuclear Information System (INIS)

Wu, Pan; Gou, Junli; Shan, Jianqiang; Jiang, Yang; Yang, Jue; Zhang, Bo

2013-01-01

Highlights: ► A new safety analysis code named SCTRAN is developed for SCWRs. ► Capability of SCTRAN is verified by comparing with code APROS and RELAP5-3D. ► A new passive safety system is proposed for CGNPC SCWR and analyzed with SCTRAN. ► CGNPC SCWR is able to cope with two critical accidents for SCWRs, LOFA and LOCA. - Abstract: Design analysis is one of the main difficulties during the research and design of SCWRs. Currently, the development of safety analysis code for SCWR is still in its infancy all around the world, and very few computer codes could carry out the trans-critical calculations where significant changes in water properties would take place. In this paper, a safety analysis code SCTRAN for SCWRs has been developed based on code RETRAN-02, the best estimate code used for safety analysis of light water reactors. The ability of SCTRAN code to simulate transients where both supercritical and subcritical regimes are encountered has been verified by comparing with APROS and RELAP5-3D codes. Furthermore, the LOFA and LOCA transients for the CGNPC SCWR design were analyzed with SCTRAN code. The characteristics and performance of the passive safety systems applied to CGNPC SCWR were evaluated. The results show that: (1) The SCTRAN computer code developed in this study is capable to perform design analysis for SCWRs; (2) During LOFA and LOCA accidents in a CGNPC SCWR, the passive safety systems would significantly mitigate the consequences of these transients and enhance the inherent safety
Cost-effectiveness analysis of countermeasures using accident consequence assessment models

International Nuclear Information System (INIS)

Alonso, A.; Gallego, E.

1987-01-01

In the event of a large release of radionuclides from a nuclear power plant, protective actions for the population potentially affected must be implemented. Cost-effectiveness analysis will be useful to define the countermeasures and the criteria needed to implement them. This paper shows the application of Accident Consequence Assessment (ACA) models to cost-effectiveness analysis of emergency and long-term countermeasures, making use of the different relationships between dose, contamination levels, affected areas and population distribution, included in such a model. The procedure is illustrated with the new Melcor Accident Consequence Code System (MACCS 1.3), developed at Sandia National Laboratories (USA), for a fixed accident scenario. Different alternative actions are evaluated with regard to their radiological and economical impact, searching for an 'optimum' strategy. (author)
A Consistent System for Coding Laboratory Samples

Science.gov (United States)

Sih, John C.

1996-07-01

A formal laboratory coding system is presented to keep track of laboratory samples. Preliminary useful information regarding the sample (origin and history) is gained without consulting a research notebook. Since this system uses and retains the same research notebook page number for each new experiment (reaction), finding and distinguishing products (samples) of the same or different reactions becomes an easy task. Using this system multiple products generated from a single reaction can be identified and classified in a uniform fashion. Samples can be stored and filed according to stage and degree of purification, e.g. crude reaction mixtures, recrystallized samples, chromatographed or distilled products.

Burnup calculation code system COMRAD96

Energy Technology Data Exchange (ETDEWEB)

Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Masukawa, Fumihiro; Ido, Masaru; Enomoto, Masaki; Takyu, Shuiti; Hara, Toshiharu

1997-06-01

COMRAD was one of the burnup code system developed by JAERI. COMRAD96 is a transfered version of COMRAD to Engineering Work Station. It is divided to several functional modules, `Cross Section Treatment`, `Generation and Depletion Calculation`, and `Post Process`. It enables us to analyze a burnup problem considering a change of neutron spectrum using UNITBURN. Also it can display the {gamma} Spectrum on a terminal. This report is the general description and user`s manual of COMRAD96. (author)
Efficient coding and detection of ultra-long IDs for visible light positioning systems.

Science.gov (United States)

Zhang, Hualong; Yang, Chuanchuan

2018-05-14

Visible light positioning (VLP) is a promising technique to complement Global Navigation Satellite System (GNSS) such as Global positioning system (GPS) and BeiDou Navigation Satellite System (BDS) which features the advantage of low-cost and high accuracy. The situation becomes even more crucial for indoor environments, where satellite signals are weak or even unavailable. For large-scale application of VLP, there would be a considerable number of Light emitting diode (LED) IDs, which bring forward the demand of long LED ID detection. In particular, to provision indoor localization globally, a convenient way is to program a unique ID into each LED during manufacture. This poses a big challenge for image sensors, such as the CMOS camera in everybody's hands since the long ID covers the span of multiple frames. In this paper, we investigate the detection of ultra-long ID using rolling shutter cameras. By analyzing the pattern of data loss in each frame, we proposed a novel coding technique to improve the efficiency of LED ID detection. We studied the performance of Reed-Solomon (RS) code in this system and designed a new coding method which considered the trade-off between performance and decoding complexity. Coding technique decreases the number of frames needed in data processing, significantly reduces the detection time, and improves the accuracy of detection. Numerical and experimental results show that the detected LED ID can be much longer with the coding technique. Besides, our proposed coding method is proved to achieve a performance close to that of RS code while the decoding complexity is much lower.
ARC Code TI: Optimal Alarm System Design and Implementation

Data.gov (United States)

National Aeronautics and Space Administration — An optimal alarm system can robustly predict a level-crossing event that is specified over a fixed prediction horizon. The code contained in this packages provides...
Improving 3D-Turbo Code's BER Performance with a BICM System over Rayleigh Fading Channel

Directory of Open Access Journals (Sweden)

R. Yao

2016-12-01

Full Text Available Classical Turbo code suffers from high error floor due to its small Minimum Hamming Distance (MHD. Newly-proposed 3D-Turbo code can effectively increase the MHD and achieve a lower error floor by adding a rate-1 post encoder. In 3D-Turbo codes, part of the parity bits from the classical Turbo encoder are further encoded through the post encoder. In this paper, a novel Bit-Interleaved Coded Modulation (BICM system is proposed by combining rotated mapping Quadrature Amplitude Modulation (QAM and 3D-Turbo code to improve the Bit Error Rate (BER performance of 3D-Turbo code over Raleigh fading channel. A key-bit protection scheme and a Two-Dimension (2D iterative soft demodulating-decoding algorithm are developed for the proposed BICM system. Simulation results show that the proposed system can obtain about 0.8-1.0 dB gain at BER of 10^{-6}, compared with the existing BICM system with Gray mapping QAM.
Consequences of pipe ruptures in metal fueled, liquid metal cooled reactors

International Nuclear Information System (INIS)

Dunn, F.E.

1990-01-01

The capability to simulate pipe ruptures has recently been added to the SASSYS-1 LMR systems analysis code. Using this capability, the consequences of severe pipe ruptures in both loop-type and pool-type reactors using metal fuel were investigated. With metal fuel, if the control rods scram then either type of reactor can easily survive a complete double-ended break of a single pipe; although, as might be expected, the consequences are less severe for a pool-type reactor. A pool-type reactor can even survive a protected simultaneous breaking of all of its inlet pipes without boiling of the coolant or melting of the fuel or cladding. 2 refs., 16 figs., 1 tab
Assessing the impact of automated coding & grouping technology at St Vincent's Hospital, Sydney.

Science.gov (United States)

Howes, M H

1993-12-01

In 1992 the Hospital recognised that the existing casemix data reporting systems were too removed from individual patients to have any meaning for clinicians, analysis of the data was difficult and the processes involved in the DRG assignment were subject to considerable error. Consequently, the Hospital approved the purchase of technology that would facilitate the coding and grouping process. The impact of automated coding and grouping technology is assessed by three methods. Firstly, by looking at by-product information systems, secondly, through subjective responses by coders to a satisfaction questionnaire and, thirdly, by objectively measuring hospital activity and identified coding elements before and after implementation of the 3M technology. It was concluded that while the 3M Coding and Grouping software should not be viewed as a panacea to all coding and documentation ills, objective evidence and subjective comment from the coders indicated an improvement in data quality and more accurate DRG assignment. Development of an in-house casemix information system and a feedback mechanism between coder and clinician had been effected. The product had been used as a training tool for coders and had also proven to be a useful auditing tool. Finally, linkage with other systems and the generation of timely reports had been realised.
Network coding and its applications to satellite systems

DEFF Research Database (Denmark)

Vieira, Fausto; Roetter, Daniel Enrique Lucani

2015-01-01

Network coding has its roots in information theory where it was initially proposed as a way to improve a two-node communication using a (broadcasting) relay. For this theoretical construct, a satellite communications system was proposed as an illustrative example, where the relay node would be a ...
Development of particle and heavy ion transport code system

International Nuclear Information System (INIS)

Niita, Koji

2004-01-01

Particle and heavy ion transport code system (PHITS) is 3 dimension general purpose Monte Carlo simulation codes for description of transport and reaction of particle and heavy ion in materials. It is developed on the basis of NMTC/JAM for design and safety of J-PARC. What is PHITS, it's physical process, physical models and development process of PHITC code are described. For examples of application, evaluation of neutron optics, cancer treatment by heavy particle ray and cosmic radiation are stated. JAM and JQMD model are used as the physical model. Neutron motion in six polar magnetic field and gravitational field, PHITC simulation of trace of C 12 beam and secondary neutron track of small model of cancer treatment device in HIMAC and neutron flux in Space Shuttle are explained. (S.Y.)
Analysis of some antecipated transients without scram for PWR type reactors by coupling of the CORAN code to the ALMOD code system

International Nuclear Information System (INIS)

Carvalho, F. de A.T. de.

1985-01-01

This study investigates some antecipated transients without scram for a pressurized water cooled reactor, using coupling of the containment CORAN code to the ALMOD code system, under severe random conditions. This coupling has the objective of including containment model as part of an unified code system. These severe conditions include failure of reactor scram, following a station black-out and emergency power initiation for the burn-up status at the beginning and end of the cycle. Furthermore, for the burn-up status at the end of the cycle, a failure in the closure of the pressurizer relief valve was also investigated. (Author) [pt
A System Structure for a VHTR-SI Process Dynamic Simulation Code

International Nuclear Information System (INIS)

Chang, Jiwoon; Shin, Youngjoon; Kim, Jihwan; Lee, Kiyoung; Lee, Wonjae; Chang, Jonghwa; Youn, Cheung

2008-01-01

The VHTR-SI process dynamic simulation code embedded in a mathematical solution engine is an application software system that simulates the dynamic behavior of the VHTR-SI process. Also, the software system supports a user friendly graphical user interface (GUI) for user input/out. Structured analysis techniques were developed in the late 1970s by Yourdon, DeMarco, Gane and Sarson for applying a systematic approach to a systems analysis. It included the use of data flow diagrams and data modeling and fostered the use of an implementation-independent graphical notation for a documentation. In this paper, we present a system structure for a VHRT-SI process dynamic simulation code by using the methodologies of structured analysis
Channel estimation for physical layer network coding systems

CERN Document Server

Gao, Feifei; Wang, Gongpu

2014-01-01

This SpringerBrief presents channel estimation strategies for the physical later network coding (PLNC) systems. Along with a review of PLNC architectures, this brief examines new challenges brought by the special structure of bi-directional two-hop transmissions that are different from the traditional point-to-point systems and unidirectional relay systems. The authors discuss the channel estimation strategies over typical fading scenarios, including frequency flat fading, frequency selective fading and time selective fading, as well as future research directions. Chapters explore the performa
Construction and performance analysis of variable-weight optical orthogonal codes for asynchronous OCDMA systems

Science.gov (United States)

Li, Chuan-qi; Yang, Meng-jie; Zhang, Xiu-rong; Chen, Mei-juan; He, Dong-dong; Fan, Qing-bin

2014-07-01

A construction scheme of variable-weight optical orthogonal codes (VW-OOCs) for asynchronous optical code division multiple access (OCDMA) system is proposed. According to the actual situation, the code family can be obtained by programming in Matlab with the given code weight and corresponding capacity. The formula of bit error rate (BER) is derived by taking account of the effects of shot noise, avalanche photodiode (APD) bulk, thermal noise and surface leakage currents. The OCDMA system with the VW-OOCs is designed and improved. The study shows that the VW-OOCs have excellent performance of BER. Despite of coming from the same code family or not, the codes with larger weight have lower BER compared with the other codes in the same conditions. By taking simulation, the conclusion is consistent with the analysis of BER in theory. And the ideal eye diagrams are obtained by the optical hard limiter.
A Novel Criterion for Optimum MultilevelCoding Systems in Mobile Fading Channels

Institute of Scientific and Technical Information of China (English)

YUAN Dongfeng; WANG Chengxiang; YAO Qi; CAO Zhigang

2001-01-01

A novel criterion that is "capac-ity rule" and "mapping rule" for the design of op-timum MLC scheme over mobile fading channels isproposed.According to this theory,the performanceof multilevel coding with multistage decoding schemes(MLC/MSD) in mobile fading channels is investi-gated,in which BCH codes are chosen as componentcodes,and three mapping strategies with 8ASK mod-ulation are used.Numerical results indicate that whencode rates of component codes in MLC scheme are de-signed based on "capacity rule",the performance ofthe system with block partitioning (BP) is optimumfor Rayleigh fading channels,while the performance ofthe system with Ungerboeck partioning (UP) is bestfor AWGN channels.
Use of source term code package in the ELEBRA MX-850 system

International Nuclear Information System (INIS)

Guimaraes, A.C.F.; Goes, A.G.A.

1988-12-01

The implantation of source term code package in the ELEBRA-MX850 system is presented. The source term is formed when radioactive materials generated in nuclear fuel leakage toward containment and the external environment to reactor containment. The implantated version in the ELEBRA system are composed of five codes: MARCH 3, TRAPMELT 3, THCCA, VANESA and NAVA. The original example case was used. The example consists of a small loca accident in a PWR type reactor. A sensitivity study for the TRAPMELT 3 code was carried out, modifying the 'TIME STEP' to estimate the processing time of CPU for executing the original example case. (M.C.K.) [pt
Psacoin level 1A intercomparison probabilistic system assessment code (PSAC) user group

International Nuclear Information System (INIS)

Nies, A.; Laurens, J.M.; Galson, D.A.; Webster, S.

1990-01-01

This report describes an international code intercomparison exercise conducted by the NEA Probabilistic System Assessment Code (PSAC) User Group. The PSACOIN Level 1A exercise is the third of a series designed to contribute to the verification of probabilistic codes that may be used in assessing the safety of radioactive waste disposal systems or concepts. Level 1A is based on a more realistic system model than that used in the two previous exercises, and involves deep geological disposal concepts with a relatively complex structure of the repository vault. The report compares results and draws conclusions with regard to the use of different modelling approaches and the possible importance to safety of various processes within and around a deep geological repository. In particular, the relative significance of model uncertainty and data variability is discussed
Probabilistic accident consequence uncertainty analysis -- Uncertainty assessment for deposited material and external doses. Volume 2: Appendices

Energy Technology Data Exchange (ETDEWEB)

Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Boardman, J. [AEA Technology (United Kingdom); Jones, J.A. [National Radiological Protection Board (United Kingdom); Harper, F.T.; Young, M.L. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States)

1997-12-01

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA deposited material and external dose models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the panel on deposited material and external doses, (4) short biographies of the experts, and (5) the aggregated results of their responses.
Physical-layer network coding in coherent optical OFDM systems.

Science.gov (United States)

Guan, Xun; Chan, Chun-Kit

2015-04-20

We present the first experimental demonstration and characterization of the application of optical physical-layer network coding in coherent optical OFDM systems. It combines two optical OFDM frames to share the same link so as to enhance system throughput, while individual OFDM frames can be recovered with digital signal processing at the destined node.
Code development of the national hemovigilance system and expansion strategies for hospital blood banks

Directory of Open Access Journals (Sweden)

Kim Jeongeun

2012-01-01

Full Text Available Objectives : The aims of this study were to develop reportable event codes that are applicable to the national hemovigilance systems for hospital blood banks, and to present expansion strategies for the blood banks. Materials and Methods : The data were obtained from a literature review and expert consultation, followed by adding to and revising the established hemovigilance code system and guidelines to develop reportable event codes for hospital blood banks. The Medical Error Reporting System-Transfusion Medicine developed in the US and other codes of reportable events were added to the Korean version of the Biologic Products Deviation Report (BPDR developed by the Korean Red Cross Blood Safety Administration, then using these codes, mapping work was conducted. We deduced outcomes suitable for practice, referred to the results of the advisory councils, and conducted a survey with experts and blood banks practitioners. Results : We developed reportable event codes that were applicable to hospital blood banks and could cover blood safety - from blood product safety to blood transfusion safety - and also presented expansion strategies for hospital blood banks. Conclusion : It was necessary to add 10 major categories to the blood transfusion safety stage and 97 reportable event codes to the blood safety stage. Contextualized solutions were presented on 9 categories of expansion strategies of hemovigilance system for the hospital blood banks.
A new two-code keying scheme for SAC-OCDMA systems enabling bipolar encoding

Science.gov (United States)

Al-Khafaji, Hamza M. R.; Ngah, Razali; Aljunid, S. A.; Rahman, T. A.

2015-03-01

In this paper, we propose a new two-code keying scheme for enabling bipolar encoding in a high-rate spectral-amplitude coding optical code-division multiple-access (SAC-OCDMA) system. The mathematical formulations are derived for the signal-to-noise ratio and bit-error rate (BER) of SAC-OCDMA system based on the suggested scheme using multi-diagonal (MD) code. Performance analyses are assessed considering the effects of phase-induced intensity noise, as well as shot and thermal noises in photodetectors. The numerical results demonstrated that the proposed scheme exhibits an enhanced BER performance compared to the existing unipolar encoding with direct detection technique. Furthermore, the performance improvement afforded by this scheme is verified using simulation experiments.
Experimental demonstration of polar coded IM/DD optical OFDM for short reach system

Science.gov (United States)

Fang, Jiafei; Xiao, Shilin; Liu, Ling; Bi, Meihua; Zhang, Lu; Zhang, Yunhao; Hu, Weisheng

2017-11-01

In this paper, we propose a novel polar coded intensity modulation direct detection (IM/DD) optical orthogonal frequency division multiplexing (OFDM) system for short reach system. A method of evaluating the channel signal noise ratio (SNR) is proposed for soft-demodulation. The experimental results demonstrate that, compared to the conventional case, ∼9.5 dB net coding gain (NCG) at the bit error rate (BER) of 1E-3 can be achieved after 40-km standard single mode fiber (SSMF) transmission. Based on the experimental result, (512,256) polar code with low complexity and satisfactory BER performance meets the requirement of low latency in short reach system, which is a promising candidate for latency-stringent short reach optical system.

Development of environmental dose assessment system (EDAS) code of PC version

Energy Technology Data Exchange (ETDEWEB)

Taki, Mitsumasa; Kikuchi, Masamitsu; Kobayashi, Hideo; Yamaguchi, Takenori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2003-05-01

A computer code (EDAS) was developed to assess the public dose for the safety assessment to get the license of nuclear reactor operation. This code system is used for the safety analysis of public around the nuclear reactor in normal operation and severe accident. This code was revised and composed for personal computer user according to the Nuclear Safety Guidelines reflected the ICRP1990 recommendation. These guidelines are revised by Nuclear Safety Commission on March, 2001, which are 'Weather analysis guideline for the safety assessment of nuclear power reactor', 'Public dose around the facility assessment guideline corresponding to the objective value for nuclear power light water reactor' and 'Public dose assessment guideline for safety review of nuclear power light water reactor'. This code has been already opened for public user by JAERI, and English version code and user manual are also prepared. This English version code is helpful for international cooperation concerning the nuclear safety assessment with JAERI. (author)
Development of environmental dose assessment system (EDAS) code of PC version

CERN Document Server

Taki, M; Kobayashi, H; Yamaguchi, T

2003-01-01

A computer code (EDAS) was developed to assess the public dose for the safety assessment to get the license of nuclear reactor operation. This code system is used for the safety analysis of public around the nuclear reactor in normal operation and severe accident. This code was revised and composed for personal computer user according to the Nuclear Safety Guidelines reflected the ICRP1990 recommendation. These guidelines are revised by Nuclear Safety Commission on March, 2001, which are 'Weather analysis guideline for the safety assessment of nuclear power reactor', 'Public dose around the facility assessment guideline corresponding to the objective value for nuclear power light water reactor' and 'Public dose assessment guideline for safety review of nuclear power light water reactor'. This code has been already opened for public user by JAERI, and English version code and user manual are also prepared. This English version code is helpful for international cooperation concerning the nuclear safety assessme...
Simulation of water hammer phenomena using the system code ATHLET

Energy Technology Data Exchange (ETDEWEB)

Bratfisch, Christoph; Koch, Marco K. [Bochum Univ. (Germany). Reactor Simulation and Safety Group

2017-07-15

Water Hammer Phenomena can endanger the integrity of structures leading to a possible failure of pipes in nuclear power plants as well as in many industrial applications. These phenomena can arise in nuclear power plants in the course of transients and accidents induced by the start-up of auxiliary feed water systems or emergency core cooling systems in combination with rapid acting valves and pumps. To contribute to further development and validation of the code ATHLET (Analysis of Thermalhydraulics of Leaks and Transients), an experiment performed in the test facility Pilot Plant Pipework (PPP) at Fraunhofer UMSICHT is simulated using the code version ATHLET 3.0A.
Simulation of water hammer phenomena using the system code ATHLET

International Nuclear Information System (INIS)

Bratfisch, Christoph; Koch, Marco K.

2017-01-01

Water Hammer Phenomena can endanger the integrity of structures leading to a possible failure of pipes in nuclear power plants as well as in many industrial applications. These phenomena can arise in nuclear power plants in the course of transients and accidents induced by the start-up of auxiliary feed water systems or emergency core cooling systems in combination with rapid acting valves and pumps. To contribute to further development and validation of the code ATHLET (Analysis of Thermalhydraulics of Leaks and Transients), an experiment performed in the test facility Pilot Plant Pipework (PPP) at Fraunhofer UMSICHT is simulated using the code version ATHLET 3.0A.
Performance Analysis of Wavelength Multiplexed Sac Ocdma Codes in Beat Noise Mitigation in Sac Ocdma Systems

Science.gov (United States)

Alhassan, A. M.; Badruddin, N.; Saad, N. M.; Aljunid, S. A.

2013-07-01

In this paper we investigate the use of wavelength multiplexed spectral amplitude coding (WM SAC) codes in beat noise mitigation in coherent source SAC OCDMA systems. A WM SAC code is a low weight SAC code, where the whole code structure is repeated diagonally (once or more) in the wavelength domain to achieve the same cardinality as a higher weight SAC code. Results show that for highly populated networks, the WM SAC codes provide better performance than SAC codes. However, for small number of active users the situation is reversed. Apart from their promising improvement in performance, these codes are more flexible and impose less complexity on the system design than their SAC counterparts.
ARTEMIS: The core simulator of AREVA NP's next generation coupled neutronics/thermal-hydraulics code system ARCADIAR

International Nuclear Information System (INIS)

Hobson, Greg; Merk, Stephan; Bolloni, Hans-Wilhelm; Breith, Karl-Albert; Curca-Tivig, Florin; Van Geemert, Rene; Heinecke, Jochen; Hartmann, Bettina; Porsch, Dieter; Tiles, Viatcheslav; Dall'Osso, Aldo; Pothet, Baptiste

2008-01-01

AREVA NP has developed a next-generation coupled neutronics/thermal-hydraulics code system, ARCADIA R , to fulfil customer's current demands and even anticipate their future demands in terms of accuracy and performance. The new code system will be implemented world-wide and will replace several code systems currently used in various global regions. An extensive phase of verification and validation of the new code system is currently in progress. One of the principal components of this new system is the core simulator, ARTEMIS. Besides the stand-alone tests on the individual computational modules, integrated tests on the overall code are being performed in order to check for non-regression as well as for verification of the code. Several benchmark problems have been successfully calculated. Full-core depletion cycles of different plant types from AREVA's French, American and German regions (e.g. N4 and KONVOI types) have been performed with ARTEMIS (using APOLLO2-A cross sections) and compared directly with current production codes, e.g. with SCIENCE and CASCADE-3D, and additionally with measurements. (authors)
Probabilistic accident consequence uncertainty analysis: Food chain uncertainty assessment. Volume 1: Main report

Energy Technology Data Exchange (ETDEWEB)

Brown, J. [National Radiological Protection Board (United Kingdom); Goossens, L.H.J.; Kraan, B.C.P. [Delft Univ. of Technology (Netherlands)] [and others

1997-06-01

This volume is the first of a two-volume document that summarizes a joint project conducted by the US Nuclear Regulatory Commission and the European Commission to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These codes were developed primarily for estimating the risks presented by nuclear reactors based on postulated frequencies and magnitudes of potential accidents. This document reports on an ongoing project to assess uncertainty in the MACCS and COSYMA calculations for the offsite consequences of radionuclide releases by hypothetical nuclear power plant accidents. A panel of sixteen experts was formed to compile credible and traceable uncertainty distributions for food chain variables that affect calculations of offsite consequences. The expert judgment elicitation procedure and its outcomes are described in these volumes. Other panels were formed to consider uncertainty in other aspects of the codes. Their results are described in companion reports. Volume 1 contains background information and a complete description of the joint consequence uncertainty study. Volume 2 contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures for both panels, (3) the rationales and results for the panels on soil and plant transfer and animal transfer, (4) short biographies of the experts, and (5) the aggregated results of their responses.
Probabilistic accident consequence uncertainty analysis: Food chain uncertainty assessment. Volume 1: Main report

International Nuclear Information System (INIS)

Brown, J.; Goossens, L.H.J.; Kraan, B.C.P.

1997-06-01

This volume is the first of a two-volume document that summarizes a joint project conducted by the US Nuclear Regulatory Commission and the European Commission to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These codes were developed primarily for estimating the risks presented by nuclear reactors based on postulated frequencies and magnitudes of potential accidents. This document reports on an ongoing project to assess uncertainty in the MACCS and COSYMA calculations for the offsite consequences of radionuclide releases by hypothetical nuclear power plant accidents. A panel of sixteen experts was formed to compile credible and traceable uncertainty distributions for food chain variables that affect calculations of offsite consequences. The expert judgment elicitation procedure and its outcomes are described in these volumes. Other panels were formed to consider uncertainty in other aspects of the codes. Their results are described in companion reports. Volume 1 contains background information and a complete description of the joint consequence uncertainty study. Volume 2 contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures for both panels, (3) the rationales and results for the panels on soil and plant transfer and animal transfer, (4) short biographies of the experts, and (5) the aggregated results of their responses
Novel BCH Code Design for Mitigation of Phase Noise Induced Cycle Slips in DQPSK Systems

DEFF Research Database (Denmark)

Leong, M. Y.; Larsen, Knud J.; Jacobsen, G.

2014-01-01

We show that by proper code design, phase noise induced cycle slips causing an error floor can be mitigated for 28 Gbau d DQPSK systems. Performance of BCH codes are investigated in terms of required overhead......We show that by proper code design, phase noise induced cycle slips causing an error floor can be mitigated for 28 Gbau d DQPSK systems. Performance of BCH codes are investigated in terms of required overhead...
Conceptual-driven classification for coding advise in health insurance reimbursement.

Science.gov (United States)

Li, Sheng-Tun; Chen, Chih-Chuan; Huang, Fernando

2011-01-01

With the non-stop increases in medical treatment fees, the economic survival of a hospital in Taiwan relies on the reimbursements received from the Bureau of National Health Insurance, which in turn depend on the accuracy and completeness of the content of the discharge summaries as well as the correctness of their International Classification of Diseases (ICD) codes. The purpose of this research is to enforce the entire disease classification framework by supporting disease classification specialists in the coding process. This study developed an ICD code advisory system (ICD-AS) that performed knowledge discovery from discharge summaries and suggested ICD codes. Natural language processing and information retrieval techniques based on Zipf's Law were applied to process the content of discharge summaries, and fuzzy formal concept analysis was used to analyze and represent the relationships between the medical terms identified by MeSH. In addition, a certainty factor used as reference during the coding process was calculated to account for uncertainty and strengthen the credibility of the outcome. Two sets of 360 and 2579 textual discharge summaries of patients suffering from cerebrovascular disease was processed to build up ICD-AS and to evaluate the prediction performance. A number of experiments were conducted to investigate the impact of system parameters on accuracy and compare the proposed model to traditional classification techniques including linear-kernel support vector machines. The comparison results showed that the proposed system achieves the better overall performance in terms of several measures. In addition, some useful implication rules were obtained, which improve comprehension of the field of cerebrovascular disease and give insights to the relationships between relevant medical terms. Our system contributes valuable guidance to disease classification specialists in the process of coding discharge summaries, which consequently brings benefits in
A UML profile for code generation of component based distributed systems

International Nuclear Information System (INIS)

Chiozzi, G.; Karban, R.; Andolfato, L.; Tejeda, A.

2012-01-01

A consistent and unambiguous implementation of code generation (model to text transformation) from UML (must rely on a well defined UML (Unified Modelling Language) profile, customizing UML for a particular application domain. Such a profile must have a solid foundation in a formally correct ontology, formalizing the concepts and their relations in the specific domain, in order to avoid a maze or set of wildly created stereotypes. The paper describes a generic profile for the code generation of component based distributed systems for control applications, the process to distill the ontology and define the profile, and the strategy followed to implement the code generator. The main steps that take place iteratively include: defining the terms and relations with an ontology, mapping the ontology to the appropriate UML meta-classes, testing the profile by creating modelling examples, and generating the code. This has allowed us to work on the modelling of E-ELT (European Extremely Large Telescope) control system and instrumentation without knowing what infrastructure will be finally used
Method of laser beam coding for control systems

Science.gov (United States)

Pałys, Tomasz; Arciuch, Artur; Walczak, Andrzej; Murawski, Krzysztof

2017-08-01

The article presents the method of encoding a laser beam for control systems. The experiments were performed using a red laser emitting source with a wavelength of λ = 650 nm and a power of P ≍ 3 mW. The aim of the study was to develop methods of modulation and demodulation of the laser beam. Results of research, in which we determined the effect of selected camera parameters, such as image resolution, number of frames per second on the result of demodulation of optical signal, is also shown in the paper. The experiments showed that the adopted coding method provides sufficient information encoded in a single laser beam (36 codes with the effectiveness of decoding at 99.9%).
Simulation and verification of the EBR-II automatic control rod drive system with continuous system modeling codes

International Nuclear Information System (INIS)

Larson, H.A.; Dean, E.M.

1985-01-01

The two computer programs are successful in modeling the EBR-II ACRDS. In fact, this is very convenient for a presampling of the consequences of a desired power movement. The ACRDS is to be modified so that the error signal is a comparison between demand position and measured position. Purpose of this change is to permit pseudo-random binary types of reactivity transfer function experiments at EBR-II. Questions asked about the computer software and hardware to accommodate this change can be quickly answered with either of the verified codes discussed here
Development of computing code system for level 3 PSA

International Nuclear Information System (INIS)

Jeong, Jong Tae; Yu, Dong Han; Kim, Seung Hwan.

1997-07-01

Among the various research areas of the level 3 PSA, the effect of terrain on the transport of radioactive material was investigated through wind tunnel experiment. These results will give a physical insight in the development of a new dispersion model. Because there are some discrepancies between the results from Gaussian plume model and those from field test, the effect of terrain on the atmospheric dispersion was investigated by using CTDMPLUS code. Through this study we find that the model which can treat terrain effect is essential in the atmospheric dispersion of radioactive materials and the CTDMPLUS model can be used as a useful tool. And it is suggested that modification of a model and experimental study should be made through the continuous effort. The health effect assessment near the Yonggwang site by using IPE (Individual plant examination) results and its site data was performed. The health effect assessment is an important part of consequence analysis of a nuclear power plant site. The MACCS was used in the assessment. Based on the calculation of CCDF for each risk measure, it is shown that CCDF has a slow slope and thus wide probability distribution in cases of early fatality, early injury, total early fatality risk, and total weighted early fatality risk. And in cases of cancer fatality and population dose within 48km and 80km, the CCDF curve have a steep slope and thus narrow probability distribution. The establishment of methodologies for necessary models for consequence analysis resulting form a server accident in the nuclear power plant was made and a program for consequence analysis was developed. The models include atmospheric transport and diffusion, calculation of exposure doses for various pathways, and assessment of health effects and associated risks. Finally, the economic impact resulting form an accident in a nuclear power plant was investigated. In this study, estimation models for each cost terms that considered in economic
Development of computing code system for level 3 PSA

Energy Technology Data Exchange (ETDEWEB)

Jeong, Jong Tae; Yu, Dong Han; Kim, Seung Hwan

1997-07-01

Among the various research areas of the level 3 PSA, the effect of terrain on the transport of radioactive material was investigated through wind tunnel experiment. These results will give a physical insight in the development of a new dispersion model. Because there are some discrepancies between the results from Gaussian plume model and those from field test, the effect of terrain on the atmospheric dispersion was investigated by using CTDMPLUS code. Through this study we find that the model which can treat terrain effect is essential in the atmospheric dispersion of radioactive materials and the CTDMPLUS model can be used as a useful tool. And it is suggested that modification of a model and experimental study should be made through the continuous effort. The health effect assessment near the Yonggwang site by using IPE (Individual plant examination) results and its site data was performed. The health effect assessment is an important part of consequence analysis of a nuclear power plant site. The MACCS was used in the assessment. Based on the calculation of CCDF for each risk measure, it is shown that CCDF has a slow slope and thus wide probability distribution in cases of early fatality, early injury, total early fatality risk, and total weighted early fatality risk. And in cases of cancer fatality and population dose within 48km and 80km, the CCDF curve have a steep slope and thus narrow probability distribution. The establishment of methodologies for necessary models for consequence analysis resulting form a server accident in the nuclear power plant was made and a program for consequence analysis was developed. The models include atmospheric transport and diffusion, calculation of exposure doses for various pathways, and assessment of health effects and associated risks. Finally, the economic impact resulting form an accident in a nuclear power plant was investigated. In this study, estimation models for each cost terms that considered in economic
Preliminary study for unified management of CANDU safety codes and construction of database system

International Nuclear Information System (INIS)

Min, Byung Joo; Kim, Hyoung Tae

2003-03-01

It is needed to develop the Graphical User Interface(GUI) for the unified management of CANDU safety codes and to construct database system for the validation of safety codes, for which the preliminary study is done in the first stage of the present work. The input and output structures and data flow of CATHENA and PRESCON2 are investigated and the interaction of the variables between CATHENA and PRESCON2 are identified. Furthermore, PC versions of CATHENA and PRESCON2 codes are developed for the interaction of these codes and GUI(Graphic User Interface). The PC versions are assessed by comparing the calculation results with those by HP workstation or from FSAR(Final Safety Analysis Report). Preliminary study on the GUI for the safety codes in the unified management system are done. The sample of GUI programming is demonstrated preliminarily. Visual C++ is selected as the programming language for the development of GUI system. The data for Wolsong plants, reactor core, and thermal-hydraulic experiments executed in the inside and outside of the country, are collected and classified following the structure of the database system, of which two types are considered for the final web-based database system. The preliminary GUI programming for database system is demonstrated, which is updated in the future work
New serial time codes for seismic short period and long period data acquisition systems

International Nuclear Information System (INIS)

Kolvankar, V.G.; Rao, D.S.

1988-01-01

This paper discusses a new time code for time indexing multichannel short period (1 to 25 hz) seismic event data recorded on a single track of magnetic tape in digital format and discusses its usefulness in contrast to Vela time code used in continuous analog multichannel data recording system on multitrack instrumentation tape deck. This paper also discusses another time code, used for time indexing of seismic long period (DC to 2.5 seconds) multichannel data recorded on a single track of magnetic tape in digital format. The time code decoding and display system developed to provide quick access to any desired portion of the tape in both data recording and repro duce system is also discussed. (author). 7 figs
Severe Accident Progression and Consequence Assessment Methodology Upgrades in ISAAC for Wolsong CANDU6

International Nuclear Information System (INIS)

Song, Y.M.; Kim, D.H.; Nijhawan, Sunil

2015-01-01

Amongst the applications of integrated severe accident analysis codes like ISAAC, the principal are to a) help develop an understanding of the severe accident progression and its consequences; b) support the design of mitigation measures by providing for them the state of the reactor following an accident; and c) to provide a training platform for accident management actions. After Fukushima accident there is an increased awareness of the need to implement effective and appropriate mitigation measures and empower the operators with training and understanding about severe accident progression and control opportunities. An updated code with reduced uncertainties can better serve these needs of the utility making decisions about mitigation measures and corrective actions. Optimal deployment of systems such as PARS and filtered containment venting require information on reactor transients for a number of critical parameters. Thus there is a greater consensus now for a demonstrated ability to perform accident progression and consequence assessment analyses with reduced uncertainties. Analyses must now provide source term transients that represent the best in available understanding and so meaningfully support mitigation measures. This requires removal of known simplifications and inclusion of all quantifiable and risk significant phenomena. Advances in understanding of CANDU6 severe accident progression reflected in the severe accident integrated code ROSHNI are being incorporated into ISAAC using CANDU specific component and system models developed and verified for Wolsong CANDU 6 reactors. A significant and comprehensive upgrade of core behavior models is being implemented in ISAAC to properly reflect the large variability amongst fuel channels in feeder geometry, fuel thermal powers and burnup. The paper summarizes the models that have been added and provides some results to illustrate code capabilities. ISAAC is being updated to meet the current requirements and
Severe Accident Progression and Consequence Assessment Methodology Upgrades in ISAAC for Wolsong CANDU6

Energy Technology Data Exchange (ETDEWEB)

Song, Y.M.; Kim, D.H. [KAERI, Daejeon (Korea, Republic of); Nijhawan, Sunil [Prolet Inc. 98 Burbank Drive, Toronto (Canada)

2015-05-15

Amongst the applications of integrated severe accident analysis codes like ISAAC, the principal are to a) help develop an understanding of the severe accident progression and its consequences; b) support the design of mitigation measures by providing for them the state of the reactor following an accident; and c) to provide a training platform for accident management actions. After Fukushima accident there is an increased awareness of the need to implement effective and appropriate mitigation measures and empower the operators with training and understanding about severe accident progression and control opportunities. An updated code with reduced uncertainties can better serve these needs of the utility making decisions about mitigation measures and corrective actions. Optimal deployment of systems such as PARS and filtered containment venting require information on reactor transients for a number of critical parameters. Thus there is a greater consensus now for a demonstrated ability to perform accident progression and consequence assessment analyses with reduced uncertainties. Analyses must now provide source term transients that represent the best in available understanding and so meaningfully support mitigation measures. This requires removal of known simplifications and inclusion of all quantifiable and risk significant phenomena. Advances in understanding of CANDU6 severe accident progression reflected in the severe accident integrated code ROSHNI are being incorporated into ISAAC using CANDU specific component and system models developed and verified for Wolsong CANDU 6 reactors. A significant and comprehensive upgrade of core behavior models is being implemented in ISAAC to properly reflect the large variability amongst fuel channels in feeder geometry, fuel thermal powers and burnup. The paper summarizes the models that have been added and provides some results to illustrate code capabilities. ISAAC is being updated to meet the current requirements and
Analysis of quantum error-correcting codes: Symplectic lattice codes and toric codes

Science.gov (United States)

Harrington, James William

Quantum information theory is concerned with identifying how quantum mechanical resources (such as entangled quantum states) can be utilized for a number of information processing tasks, including data storage, computation, communication, and cryptography. Efficient quantum algorithms and protocols have been developed for performing some tasks (e.g. , factoring large numbers, securely communicating over a public channel, and simulating quantum mechanical systems) that appear to be very difficult with just classical resources. In addition to identifying the separation between classical and quantum computational power, much of the theoretical focus in this field over the last decade has been concerned with finding novel ways of encoding quantum information that are robust against errors, which is an important step toward building practical quantum information processing devices. In this thesis I present some results on the quantum error-correcting properties of oscillator codes (also described as symplectic lattice codes) and toric codes. Any harmonic oscillator system (such as a mode of light) can be encoded with quantum information via symplectic lattice codes that are robust against shifts in the system's continuous quantum variables. I show the existence of lattice codes whose achievable rates match the one-shot coherent information over the Gaussian quantum channel. Also, I construct a family of symplectic self-dual lattices and search for optimal encodings of quantum information distributed between several oscillators. Toric codes provide encodings of quantum information into two-dimensional spin lattices that are robust against local clusters of errors and which require only local quantum operations for error correction. Numerical simulations of this system under various error models provide a calculation of the accuracy threshold for quantum memory using toric codes, which can be related to phase transitions in certain condensed matter models. I also present

DABIE: a data banking system of integral experiments for reactor core characteristics computer codes

International Nuclear Information System (INIS)

Matsumoto, Kiyoshi; Naito, Yoshitaka; Ohkubo, Shuji; Aoyanagi, Hideo.

1987-05-01

A data banking system of integral experiments for reactor core characteristics computer codes, DABIE, has been developed to lighten the burden on searching so many documents to obtain experiment data required for verification of reactor core characteristics computer code. This data banking system, DABIE, has capabilities of systematic classification, registration and easy retrieval of experiment data. DABIE consists of data bank and supporting programs. Supporting programs are data registration program, data reference program and maintenance program. The system is designed so that user can easily register information of experiment systems including figures as well as geometry data and measured data or obtain those data through TSS terminal interactively. This manual describes the system structure, how-to-use and sample uses of this code system. (author)
Auto Code Generation for Simulink-Based Attitude Determination Control System

Science.gov (United States)

MolinaFraticelli, Jose Carlos

2012-01-01

This paper details the work done to auto generate C code from a Simulink-Based Attitude Determination Control System (ADCS) to be used in target platforms. NASA Marshall Engineers have developed an ADCS Simulink simulation to be used as a component for the flight software of a satellite. This generated code can be used for carrying out Hardware in the loop testing of components for a satellite in a convenient manner with easily tunable parameters. Due to the nature of the embedded hardware components such as microcontrollers, this simulation code cannot be used directly, as it is, on the target platform and must first be converted into C code; this process is known as auto code generation. In order to generate C code from this simulation; it must be modified to follow specific standards set in place by the auto code generation process. Some of these modifications include changing certain simulation models into their atomic representations which can bring new complications into the simulation. The execution order of these models can change based on these modifications. Great care must be taken in order to maintain a working simulation that can also be used for auto code generation. After modifying the ADCS simulation for the auto code generation process, it is shown that the difference between the output data of the former and that of the latter is between acceptable bounds. Thus, it can be said that the process is a success since all the output requirements are met. Based on these results, it can be argued that this generated C code can be effectively used by any desired platform as long as it follows the specific memory requirements established in the Simulink Model.
An examination of the consequences in high consequence operations

Energy Technology Data Exchange (ETDEWEB)

Spray, S.D.; Cooper, J.A.

1996-06-01

Traditional definitions of risk partition concern into the probability of occurrence and the consequence of the event. Most safety analyses focus on probabilistic assessment of an occurrence and the amount of some measurable result of the event, but the real meaning of the ``consequence`` partition is usually afforded less attention. In particular, acceptable social consequence (consequence accepted by the public) frequently differs significantly from the metrics commonly proposed by risk analysts. This paper addresses some of the important system development issues associated with consequences, focusing on ``high consequence operations safety.``
Sequence Coding and Search System Backfit Quality Assurance Program Plan

International Nuclear Information System (INIS)

Lovell, C.J.; Stepina, P.L.

1985-03-01

The Sequence Coding and Search System is a computer-based encoding system for events described in Licensee Event Reports. This data system contains LERs from 1981 to present. Backfit of the data system to include LERs prior to 1981 is required. This report documents the Quality Assurance Program Plan that EG and G Idaho, Inc. will follow while encoding 1980 LERs
Comparison of PSF maxima and minima of multiple annuli coded aperture (MACA) and complementary multiple annuli coded aperture (CMACA) systems

Science.gov (United States)

Ratnam, Challa; Lakshmana Rao, Vadlamudi; Lachaa Goud, Sivagouni

2006-10-01

In the present paper, and a series of papers to follow, the Fourier analytical properties of multiple annuli coded aperture (MACA) and complementary multiple annuli coded aperture (CMACA) systems are investigated. First, the transmission function for MACA and CMACA is derived using Fourier methods and, based on the Fresnel-Kirchoff diffraction theory, the formulae for the point spread function are formulated. The PSF maxima and minima are calculated for both the MACA and CMACA systems. The dependence of these properties on the number of zones is studied and reported in this paper.
Comparison of PSF maxima and minima of multiple annuli coded aperture (MACA) and complementary multiple annuli coded aperture (CMACA) systems

International Nuclear Information System (INIS)

Ratnam, Challa; Rao, Vadlamudi Lakshmana; Goud, Sivagouni Lachaa

2006-01-01

In the present paper, and a series of papers to follow, the Fourier analytical properties of multiple annuli coded aperture (MACA) and complementary multiple annuli coded aperture (CMACA) systems are investigated. First, the transmission function for MACA and CMACA is derived using Fourier methods and, based on the Fresnel-Kirchoff diffraction theory, the formulae for the point spread function are formulated. The PSF maxima and minima are calculated for both the MACA and CMACA systems. The dependence of these properties on the number of zones is studied and reported in this paper
Development of 3-D Flow Analysis Code for Fuel Assembly using Unstructured Grid System

Energy Technology Data Exchange (ETDEWEB)

Myong, Hyon Kook; Kim, Jong Eun; Ahn, Jong Ki; Yang, Seung Yong [Kookmin Univ., Seoul (Korea, Republic of)

2007-03-15

The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Final objective of this study is to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system. In order to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system, the following researches are made: - Development of numerical algorithm for CFD code's solver - Grid and geometric connectivity data - Development of software(PowerCFD code) for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system - Modulation of software(PowerCFD code) - Development of turbulence model - Development of analysis module of RANS/LES hybrid models - Analysis of turbulent flow and heat transfer - Basic study on LES analysis - Development of main frame on pre/post processors based on GUI - Algorithm for fully-developed flow.
Development of 3-D Flow Analysis Code for Fuel Assembly using Unstructured Grid System

International Nuclear Information System (INIS)

Myong, Hyon Kook; Kim, Jong Eun; Ahn, Jong Ki; Yang, Seung Yong

2007-03-01

The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Final objective of this study is to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system. In order to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system, the following researches are made: - Development of numerical algorithm for CFD code's solver - Grid and geometric connectivity data - Development of software(PowerCFD code) for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system - Modulation of software(PowerCFD code) - Development of turbulence model - Development of analysis module of RANS/LES hybrid models - Analysis of turbulent flow and heat transfer - Basic study on LES analysis - Development of main frame on pre/post processors based on GUI - Algorithm for fully-developed flow
Processing Expert Judgements in Accident Consequence Modelling (invited paper)

International Nuclear Information System (INIS)

Kraan, B.C.P.; Cooke, R.M.

2000-01-01

In performing uncertainty analysis a distribution on the code input parameters is required. The construction of the distribution on the code input parameters for the joint CEC/USNRC Accident Consequence Code Uncertainty Analysis using Expert Judgement is discussed. An example from the food chain module is used to illustrate the construction. Different mathematical techniques have been developed to transform the expert judgements into the required format. Finally, the effect of taking account of correlations in performing uncertainty analysis is investigated. (author)
About the coding system of rivers, catchment basing and their characteristics of the republic of Armenia

International Nuclear Information System (INIS)

Avagyan, A.A.; Arakelyan, A.A.

2011-01-01

The coding of rivers, catchements, lakes and seas is one of the most important requirements of Water Framework Directive of the European Union. This coding provides solutions to actual problems of planning and management of water resources of the Republic of Armenia. The coding system provides the hierarchy of water bodies and watersheds with their typology as well as their geographic and natural conditions, anthropogenic pressures and ecological status. This approach is a fundamentally new complex solution to the coding of water resources. The coding technique allows you to automate the assessment and mapping of environmental risks and areas of water bodies which are subjected to significant pressure and also helps to solve other problems concerning the planning and the management of water resources. A complex code of each water body consists of the following groups of codes: Hydrographic code - an identifier of a water body in the hydrographic system of the country; Codes of static attributes in the system requirements of the Water Framework Directive of the European Union; Codes of static attributes of the qualifiers of the RA National Water Program; Codes of dynamic attributes that define the quality of water and characteristics of water use; Codes of dynamic attributes describing the human impact and determining the ecological status of water body
Perceptual consequences of disrupted auditory nerve activity.

Science.gov (United States)

Zeng, Fan-Gang; Kong, Ying-Yee; Michalewski, Henry J; Starr, Arnold

2005-06-01

Perceptual consequences of disrupted auditory nerve activity were systematically studied in 21 subjects who had been clinically diagnosed with auditory neuropathy (AN), a recently defined disorder characterized by normal outer hair cell function but disrupted auditory nerve function. Neurological and electrophysical evidence suggests that disrupted auditory nerve activity is due to desynchronized or reduced neural activity or both. Psychophysical measures showed that the disrupted neural activity has minimal effects on intensity-related perception, such as loudness discrimination, pitch discrimination at high frequencies, and sound localization using interaural level differences. In contrast, the disrupted neural activity significantly impairs timing related perception, such as pitch discrimination at low frequencies, temporal integration, gap detection, temporal modulation detection, backward and forward masking, signal detection in noise, binaural beats, and sound localization using interaural time differences. These perceptual consequences are the opposite of what is typically observed in cochlear-impaired subjects who have impaired intensity perception but relatively normal temporal processing after taking their impaired intensity perception into account. These differences in perceptual consequences between auditory neuropathy and cochlear damage suggest the use of different neural codes in auditory perception: a suboptimal spike count code for intensity processing, a synchronized spike code for temporal processing, and a duplex code for frequency processing. We also proposed two underlying physiological models based on desynchronized and reduced discharge in the auditory nerve to successfully account for the observed neurological and behavioral data. These methods and measures cannot differentiate between these two AN models, but future studies using electric stimulation of the auditory nerve via a cochlear implant might. These results not only show the unique
“PROCESS”: A systems code for fusion power plants—Part 1: Physics

Energy Technology Data Exchange (ETDEWEB)

Kovari, M., E-mail: michael.kovari@ccfe.ac.uk; Kemp, R.; Lux, H.; Knight, P.; Morris, J.; Ward, D.J.

2014-12-15

Highlights: • PROCESS is a fusion reactor systems code. • It optimises a figure of merit subject to constraints chosen by the user. • CCFE are working to make the assumptions and equations explicit and public. • The PROCESS homepage is (www.ccfe.ac.uk/powerplants.aspx). - Abstract: PROCESS is a reactor systems code – it assesses the engineering and economic viability of a hypothetical fusion power station using simple models of all parts of a reactor system, from the basic plasma physics to the generation of electricity. It has been used for many years, but details of its operation have not been previously published. This paper describes some of its capabilities. PROCESS is usually used in optimisation mode, in which it finds a set of parameters that maximise (or minimise) a figure of merit chosen by the user, while being consistent with the inputs and the specified constraints. Because the user can apply all the physically relevant constraints, while allowing a large number of parameters to vary, it is in principle only necessary to run the code once to produce a self-consistent, physically plausible reactor model. The scope of PROCESS is very wide and goes well beyond reactor physics, including conversion of heat to electricity, buildings, and costs, but this paper describes only the plasma physics and magnetic field calculations. The capabilities of PROCESS in plasma physics are limited, as its main aim is to combine engineering, physics and economics. A model is described which shows the main plasma features of an inductive ITER scenario. Significant differences between the PROCESS results and the published scenario include the bootstrap current and loop voltage. The PROCESS models for these are being revised. Two new models for DEMO have been obtained. The first, DEMO A, is intended to be “conservative” in that it might be possible to build it using the technology of the near future. For example, since current drive technologies are not yet
Noise suppression system of OCDMA with spectral/spatial 2D hybrid code

Science.gov (United States)

Matem, Rima; Aljunid, S. A.; Junita, M. N.; Rashidi, C. B. M.; Shihab Aqrab, Israa

2017-11-01

In this paper, we propose a novel 2D spectral/spatial hybrid code based on 1D ZCC and 1D MD where the both present a zero cross correlation property analyzed and the influence of the noise of optical as Phase Induced Intensity Noise (PIIN), shot and thermal noise. This new code is shown effectively to mitigate the PIIN and suppresses MAI. Using 2D ZCC/MD code the performance of the system can be improved in term of as well as to support more simultaneous users compared of the 2D FCC/MDW and 2D DPDC codes.
Noise suppression system of OCDMA with spectral/spatial 2D hybrid code

Directory of Open Access Journals (Sweden)

Matem Rima

2017-01-01

Full Text Available In this paper, we propose a novel 2D spectral/spatial hybrid code based on 1D ZCC and 1D MD where the both present a zero cross correlation property analyzed and the influence of the noise of optical as Phase Induced Intensity Noise (PIIN, shot and thermal noise. This new code is shown effectively to mitigate the PIIN and suppresses MAI. Using 2D ZCC/MD code the performance of the system can be improved in term of as well as to support more simultaneous users compared of the 2D FCC/MDW and 2D DPDC codes.
Gene-Auto: Automatic Software Code Generation for Real-Time Embedded Systems

Science.gov (United States)

Rugina, A.-E.; Thomas, D.; Olive, X.; Veran, G.

2008-08-01

This paper gives an overview of the Gene-Auto ITEA European project, which aims at building a qualified C code generator from mathematical models under Matlab-Simulink and Scilab-Scicos. The project is driven by major European industry partners, active in the real-time embedded systems domains. The Gene- Auto code generator will significantly improve the current development processes in such domains by shortening the time to market and by guaranteeing the quality of the generated code through the use of formal methods. The first version of the Gene-Auto code generator has already been released and has gone thought a validation phase on real-life case studies defined by each project partner. The validation results are taken into account in the implementation of the second version of the code generator. The partners aim at introducing the Gene-Auto results into industrial development by 2010.
Extending CANTUP code analysis to probabilistic evaluations

International Nuclear Information System (INIS)

Florea, S.

2001-01-01

The structural analysis with numerical methods based on final element method plays at present a central role in evaluations and predictions of structural systems which require safety and reliable operation in aggressive environmental conditions. This is the case too for the CANDU - 600 fuel channel, where besides the corrosive and thermal aggression upon the Zr97.5Nb2.5 pressure tubes, a lasting irradiation adds which has marked consequences upon the materials properties evolution. This results in an unavoidable spreading in the materials properties in time, affected by high uncertainties. Consequently, the deterministic evaluation with computation codes based on finite element method are supplemented by statistic and probabilistic methods of evaluation of the response of structural components. This paper reports the works on extending the thermo-mechanical evaluation of the fuel channel components in the frame of probabilistic structure mechanics based on statistical methods and developed upon deterministic CANTUP code analyses. CANTUP code was adapted from LAHEY 77 platform onto Microsoft Developer Studio - Fortran Power Station 4.0 platform. To test the statistical evaluation of the creeping behaviour of pressure tube, the value of longitudinal elasticity modulus (Young) was used, as random variable, with a normal distribution around value, as used in deterministic analyses. The influence of the random quantity upon the hog and effective stress developed in the pressure tube for to time values, specific to primary and secondary creep was studied. The results obtained after a five year creep, corresponding to the secondary creep are presented
Turbo-Gallager Codes: The Emergence of an Intelligent Coding ...

African Journals Online (AJOL)

Today, both turbo codes and low-density parity-check codes are largely superior to other code families and are being used in an increasing number of modern communication systems including 3G standards, satellite and deep space communications. However, the two codes have certain distinctive characteristics that ...
The design of the CMOS wireless bar code scanner applying optical system based on ZigBee

Science.gov (United States)

Chen, Yuelin; Peng, Jian

2008-03-01

The traditional bar code scanner is influenced by the length of data line, but the farthest distance of the wireless bar code scanner of wireless communication is generally between 30m and 100m on the market. By rebuilding the traditional CCD optical bar code scanner, a CMOS code scanner is designed based on the ZigBee to meet the demands of market. The scan system consists of the CMOS image sensor and embedded chip S3C2401X, when the two dimensional bar code is read, the results show the inaccurate and wrong code bar, resulted from image defile, disturber, reads image condition badness, signal interference, unstable system voltage. So we put forward the method which uses the matrix evaluation and Read-Solomon arithmetic to solve them. In order to construct the whole wireless optics of bar code system and to ensure its ability of transmitting bar code image signals digitally with long distances, ZigBee is used to transmit data to the base station, and this module is designed based on image acquisition system, and at last the wireless transmitting/receiving CC2430 module circuit linking chart is established. And by transplanting the embedded RTOS system LINUX to the MCU, an applying wireless CMOS optics bar code scanner and multi-task system is constructed. Finally, performance of communication is tested by evaluation software Smart RF. In broad space, every ZIGBEE node can realize 50m transmission with high reliability. When adding more ZigBee nodes, the transmission distance can be several thousands of meters long.
ERRORJ. Covariance processing code system for JENDL. Version 2

International Nuclear Information System (INIS)

Chiba, Gou

2003-09-01

ERRORJ is the covariance processing code system for Japanese Evaluated Nuclear Data Library (JENDL) that can produce group-averaged covariance data to apply it to the uncertainty analysis of nuclear characteristics. ERRORJ can treat the covariance data for cross sections including resonance parameters as well as angular distributions and energy distributions of secondary neutrons which could not be dealt with by former covariance processing codes. In addition, ERRORJ can treat various forms of multi-group cross section and produce multi-group covariance file with various formats. This document describes an outline of ERRORJ and how to use it. (author)
A Source Term Calculation for the APR1400 NSSS Auxiliary System Components Using the Modified SHIELD Code

International Nuclear Information System (INIS)

Park, Hong Sik; Kim, Min; Park, Seong Chan; Seo, Jong Tae; Kim, Eun Kee

2005-01-01

The SHIELD code has been used to calculate the source terms of NSSS Auxiliary System (comprising CVCS, SIS, and SCS) components of the OPR1000. Because the code had been developed based upon the SYSTEM80 design and the APR1400 NSSS Auxiliary System design is considerably changed from that of SYSTEM80 or OPR1000, the SHIELD code cannot be used directly for APR1400 radiation design. Thus the hand-calculation is needed for the portion of design changes using the results of the SHIELD code calculation. In this study, the SHIELD code is modified to incorporate the APR1400 design changes and the source term calculation is performed for the APR1400 NSSS Auxiliary System components

Coupling of RELAP5-3D and GAMMA codes for Nuclear Hydrogen System Analysis

International Nuclear Information System (INIS)

Jin, Hyung Gon

2007-02-01

RELAP5-3D is one of the most important system analysis codes in nuclear field, which has been developed for best-estimate transient simulation of light water reactor coolant systems during postulated accidents. The GAMMA code is a multi-dimensional multi-component mixture analysis code with the complete set of chemical reaction models which is developed for safety analysis of HTGR (High Temperature Gas Cooled Reactor) air-ingress. The two codes, RELAP5-3D and GAMMA, are coupled to be used for nuclear-hydrogen system analysis, which requires the capability of the analysis of multi-component gas mixture and two-phase flow. In order to couple the two codes, 4 steps are needed. Before coupling, the GAMMA code was transformed into DLL (dynamic link liberally) from executive type and RELAP5-3D was recompiled into Compaq Visual Fortran environments for our debugging purpose. As the second step, two programs - RELAP5-3D and GAMMA codes - must be synchronized in terms of time and time step. Based on that time coupling, the coupled code can calculate simultaneously. Time-step coupling had been accomplished successfully and it is tested by using a simple test input. As a next step, source-term coupling was done and it was also tested in two different test inputs. The fist case is a simple test condition, which has no chemical reaction. And the other test set is the chemical reaction model, including four non-condensable gas species, which are He, O2, CO, CO2. Finally, in order to analyze combined cycle system, heat-flux coupling has been made and a simple heat exchanger model was demonstrated
Development of a system of computer codes for severe accident analyses and its applications

Energy Technology Data Exchange (ETDEWEB)

Chang, Soon Hong; Cheon, Moon Heon; Cho, Nam jin; No, Hui Cheon; Chang, Hyeon Seop; Moon, Sang Kee; Park, Seok Jeong; Chung, Jee Hwan [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1991-12-15

The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in Nuclear Power Plants. This system of codes is necessary to conduct individual plant examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident resistance. The scope and contents of this study are as follows : development of a system of computer codes for severe accident analyses, development of severe accident management strategy.
Development of a system of computer codes for severe accident analyses and its applications

International Nuclear Information System (INIS)

Chang, Soon Hong; Cheon, Moon Heon; Cho, Nam jin; No, Hui Cheon; Chang, Hyeon Seop; Moon, Sang Kee; Park, Seok Jeong; Chung, Jee Hwan

1991-12-01

The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in Nuclear Power Plants. This system of codes is necessary to conduct individual plant examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident resistance. The scope and contents of this study are as follows : development of a system of computer codes for severe accident analyses, development of severe accident management strategy
Distributed magnetic field positioning system using code division multiple access

Science.gov (United States)

Prigge, Eric A. (Inventor)

2003-01-01

An apparatus and methods for a magnetic field positioning system use a fundamentally different, and advantageous, signal structure and multiple access method, known as Code Division Multiple Access (CDMA). This signal architecture, when combined with processing methods, leads to advantages over the existing technologies, especially when applied to a system with a large number of magnetic field generators (beacons). Beacons at known positions generate coded magnetic fields, and a magnetic sensor measures a sum field and decomposes it into component fields to determine the sensor position and orientation. The apparatus and methods can have a large `building-sized` coverage area. The system allows for numerous beacons to be distributed throughout an area at a number of different locations. A method to estimate position and attitude, with no prior knowledge, uses dipole fields produced by these beacons in different locations.
DARWIN. An evolution code system for a large range of applications

International Nuclear Information System (INIS)

Tsilanizara, A.; Diop, C.M.; Nimal, B.

2000-01-01

The aim of this article is to present the main capabilities of an evolution code system, DARWIN, developed at CEA (France). It is devoted to radioactivity studies in various application fields such as nuclear fuel cycle, dismantling, thermonuclear fusion, accelerator driven system, medecine etc. All types of nuclides are dealt with: actinides, fission products, activation products, spallation products. Physical quantities calculated by the code are isotope concentration, isotope mass, activity, radiotoxicity, gamma spectra, beta spectra, alpha spectra, neutron production by spontaneous fission and (α, n) reaction, residual heating, for any cooling times until geological times. Both analytical and numerical schemes are developed in the PEPIN2 depletion module of DARWIN to solve the generalized coupled differential depletion equations. The depletion module PEPIN2 is automatically linked to international evaluations (JEF2, ENDF/B6, EAF97...) both for decay data and cross-sections, and to some transport codes such as TRIPOLI, APOLLO2 and ERANOS. These transport codes provide neutronic data as self-shielded cross-sections and neutron fluxes. DARWIN includes a generator of radioisotope chain built automatically from decay modes and nuclear reaction types specified in the evaluation libraries. A 'search engine' allows to determine all formation ways of a considered isotope. Several examples are given for illustrating capabilities of DARWIN in different field applications. Some comparisons with other codes such as ORIGEN, FISPIN and FISPACT are also presented. (author)
Symmetries in Genetic Systems and the Concept of Geno-Logical Coding

Directory of Open Access Journals (Sweden)

Sergey V. Petoukhov

2016-12-01

Full Text Available The genetic code of amino acid sequences in proteins does not allow understanding and modeling of inherited processes such as inborn coordinated motions of living bodies, innate principles of sensory information processing, quasi-holographic properties, etc. To be able to model these phenomena, the concept of geno-logical coding, which is connected with logical functions and Boolean algebra, is put forward. The article describes basic pieces of evidence in favor of the existence of the geno-logical code, which exists in parallel with the known genetic code of amino acid sequences but which serves for transferring inherited processes along chains of generations. These pieces of evidence have been received due to the analysis of symmetries in structures of molecular-genetic systems. The analysis has revealed a close connection of the genetic system with dyadic groups of binary numbers and with other mathematical objects, which are related with dyadic groups: Walsh functions (which are algebraic characters of dyadic groups, bit-reversal permutations, logical holography, etc. These results provide a new approach for mathematical modeling of genetic structures, which uses known mathematical formalisms from technological fields of noise-immunity coding of information, binary analysis, logical holography, and digital devices of artificial intellect. Some opportunities for a development of algebraic-logical biology are opened.
Comparative calculations on selected two-phase flow phenomena using major PWR system codes

International Nuclear Information System (INIS)

1990-01-01

In 1988 a comparative study on important features and models in six major best estimate thermal hydraulic codes for PWR systems was implemented (Comparison of thermal hydraulic safety codes for PWR Graham, Trotman, London, EUR 11522). It was a limitation of that study that the source codes themselves were not available but the comparison had to be based on the available documentation. In the present study, the source codes were available and the capability of four system codes to predict complex two-phase flow phenomena has been assessed. Two areas of investigation were selected: (a) pressurized spray phenomena; (b) boil-up phenomena in rod bundles. As regards the first area, experimental data obtained in 1972 on the Neptunus Facility (Delft University of Technology) were compared with the results of the calculations using Athlet, Cathare, Relap 5 and TRAC-PT1 and, concerning the second area, the results of two experimental facilities obtained in 1980 and 1985 on Thetis (UKEA) and Pericles (CEA-Grenoble) were considered
CASKETSS-2: a computer code system for thermal and structural analysis of nuclear fuel shipping casks (version 2)

International Nuclear Information System (INIS)

Ikushima, Takeshi

1991-08-01

A computer program CASKETSS-2 has been developed for the purpose of thermal and structural analysis of nuclear fuel shipping casks. CASKETSS-2 means a modular code system for CASK Evaluation code system Thermal and Structural Safety (Version 2). Main features of CASKETSS-2 are as follow; (1) Thermal and structural analysis computer programs for one-, two-, three-dimensional geometries are contained in the code system. (2) There are simplified computer programs and a detailed one in the structural analysis part in the code system. (3) Input data generator is provided in the code system. (4) 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)
Quality Improvement of MARS Code and Establishment of Code Coupling

International Nuclear Information System (INIS)

Chung, Bub Dong; Jeong, Jae Jun; Kim, Kyung Doo

2010-04-01

The improvement of MARS code quality and coupling with regulatory auditing code have been accomplished for the establishment of self-reliable technology based regulatory auditing system. The unified auditing system code was realized also by implementing the CANDU specific models and correlations. As a part of the quality assurance activities, the various QA reports were published through the code assessments. The code manuals were updated and published a new manual which describe the new models and correlations. The code coupling methods were verified though the exercise of plant application. The education-training seminar and technology transfer were performed for the code users. The developed MARS-KS is utilized as reliable auditing tool for the resolving the safety issue and other regulatory calculations. The code can be utilized as a base technology for GEN IV reactor applications
Integrated transport code system for a multicomponent plasma in a gas dynamic trap

International Nuclear Information System (INIS)

Anikeev, A.V.; Karpushov, A.N.; Noak, K.; Strogalova, S.L.

2000-01-01

This report is focused on the development of the theoretical and numerical models of multicomponent high-β plasma confinement and transport in the gas-dynamic trap (GDT). In order to simulate the plasma behavior in the GDT as well as that in the GDT-based neutron source the Integrated Transport Code System is developed from existing stand-alone codes calculating the target plasma, the fast ions and the neutral gas in the GDT. The code system considers the full dependence of the transport phenomena on space, time, energy and angle variables as well as the interactions between the particle fields [ru
Investigation and proposal of the system to affect nuclear fuel type authorization and analysis code certification

International Nuclear Information System (INIS)

2006-03-01

In order to develop the system to affect more advanced and rational regulations of nuclear fuels and earlier introduction of new technologies in nuclear power plants, domestic and overseas safety regulation systems and state of their implementation for water cooled reactor fuel and safety analysis code had been investigated and new regulation system to affect nuclear fuel type authorization and analysis code certification was proposed. Topical report system for common parts related with nuclear fuel type authorization and analysis code certification was firstly proposed for knowledge base. Maintaining consistent safety examination supported by experts, introduction of domestic efficient system for lead irradiation test fuel, and analysis code certification and quality assurance were also proposed. (T. Tanaka)
A Spanish version for the new ERA-EDTA coding system for primary renal disease

Directory of Open Access Journals (Sweden)

Óscar Zurriaga

2015-07-01

Conclusions: Translation and adaptation into Spanish represent an improvement that will help to introduce and use the new coding system for PKD, as it can help reducing the time devoted to coding and also the period of adaptation of health workers to the new codes.
LHC beam dump system Consequences of abnormal operation

CERN Document Server

Kramer, T; Uythoven, J

2010-01-01

The LHC beam dump system is one of the most critical systems concerning machine protection and safe operation. It is used to dispose of high intensity beams between 450 GeV and 7 TeV. Studies into the consequences of abnormal beam dump actions have been performed. Different error scenarios have been evaluated using particle tracking in MAD-X, including an asynchronous dump action, and the impact of different orbit and collimator settings. Losses at locations in the ring and the beam dump transfer lines have been quantified as a function of different settings of the dump system protection elements. The implications for the setting up and operation of these protection elements are discussed.
Calculation of mixed HEU-LEU cores for the HOR research reactor with the scale code system

International Nuclear Information System (INIS)

Leege, P.F.A. de; Gibcus, H.P.M.; Hoogenboom, J.E.; Vries, J.W. de

1997-01-01

The HOR reactor of Interfaculty Reactor Institute (IRI), Delft, The Netherlands, will be converted to use low enriched fuel (LEU) assemblies. As there are still many usable high enriched (HEU) fuel assemblies present, there will be a considerable reactor operation time with mixed cores with both HEU and LEU fuel assemblies. At IRI a comprehensive reactor physics code system and evaluated nuclear data is implemented for detailed core calculations. One of the backbones of the IRI code system is the well-known SCALE code system package. Full core calculations are performed with the diffusion theory code BOLD VENTURE, the nodal code SILWER, and the Monte Carlo code KENO Va. Results are displayed of a strategy from a HEU core to a mixed HEU-LEU core and eventually a LEU core. (author)
PROMETHEUS - a code system for dynamic 3-D analysis of nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Khotylev, V.A.; Hoogenboom, J.E.; Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.

1996-09-01

The paper presents a multidimensional, general-purpose neutronics code system. It solves a number of steady-state and/or transient problems with coupled thermal hydraulics in one-, two-, or three-dimensional geometry. Due to a number of specialized features such as cavity treatment, automated convergence control, burnup treatment using the full isotopic transition matrix, the code system can be applied for the analysis of fast and slow transients in small, large, and innovative reactor cores. (author)
Development of simplified decommissioning cost estimation code for nuclear facilities

International Nuclear Information System (INIS)

Tachibana, Mitsuo; Shiraishi, Kunio; Ishigami, Tsutomu

2010-01-01

The simplified decommissioning cost estimation code for nuclear facilities (DECOST code) was developed in consideration of features and structures of nuclear facilities and similarity of dismantling methods. The DECOST code could calculate 8 evaluation items of decommissioning cost. Actual dismantling in the Japan Atomic Energy Agency (JAEA) was evaluated; unit conversion factors used to calculate the manpower of dismantling activities were evaluated. Consequently, unit conversion factors of general components could be classified into three kinds. Weights of components and structures of the facility were necessary for calculation of manpower. Methods for evaluating weights of components and structures of the facility were studied. Consequently, the weight of components in the facility was proportional to the weight of structures of the facility. The weight of structures of the facility was proportional to the total area of floors in the facility. Decommissioning costs of 7 nuclear facilities in the JAEA were calculated by using the DECOST code. To verify the calculated results, the calculated manpower was compared with the manpower gained from actual dismantling. Consequently, the calculated manpower and actual manpower were almost equal. The outline of the DECOST code, evaluation results of unit conversion factors, the evaluation method of the weights of components and structures of the facility are described in this report. (author)
DNBR calculation in digital core protection system by a subchannel analysis code

International Nuclear Information System (INIS)

In, W. K.; Yoo, Y. J.; Hwang, T. H.; Ji, S. K.

2001-01-01

The DNBR calculation uncertainty and DNBR margin were evaluated in digital core protection system by a thermal-hydrualic subchannel analysis code MATRA. A simplified thermal-hydraulic code CETOP is used to calculate on-line DNBR in core protection system at a digital PWR. The DNBR tuning process against a best-estimate subchannel analysis code is required for CETOP to ensure accurate and conservative DNBR calculation but not necessary for MATRA. The DNBR calculations by MATRA and CETOP were performed for a large number of operating condition in Yonggwang nulcear units 3-4 where the digitial core protection system is initially implemented in Korea. MATRA resulted in a less negative mean value (i.e., reduce the overconservatism) and a somewhat larger standard deviation of the DNBR error. The uncertainty corrected minimum DNBR by MATRA was shown to be higher by 1.8% -9.9% that the CETOP DNBR
Erosion corrosion in power plant piping systems - Calculation code for predicting wall thinning

International Nuclear Information System (INIS)

Kastner, W.; Erve, M.; Henzel, N.; Stellwag, B.

1990-01-01

Extensive experimental and theoretical investigations have been performed to develop a calculation code for wall thinning due to erosion corrosion in power plant piping systems. The so-called WATHEC code can be applied to single-phase water flow as well as to two-phase water/steam flow. Only input data which are available to the operator of the plant are taken into consideration. Together with a continuously updated erosion corrosion data base the calculation code forms one element of a weak point analysis for power plant piping systems which can be applied to minimize material loss due to erosion corrosion, reduce non-destructive testing and curtail monitoring programs for piping systems, recommend life-extending measures. (author). 12 refs, 17 figs
Probabilistic accident consequence uncertainty analysis: Dispersion and deposition uncertainty assessment, appendices A and B

International Nuclear Information System (INIS)

Harper, F.T.; Young, M.L.; Miller, L.A.; Hora, S.C.; Lui, C.H.; Goossens, L.H.J.; Cooke, R.M.; Paesler-Sauer, J.; Helton, J.C.

1995-01-01

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, completed in 1990, estimate the risks presented by nuclear installations based on postulated frequencies and magnitudes of potential accidents. In 1991, the US Nuclear Regulatory Commission (NRC) and the Commission of the European Communities (CEC) began a joint uncertainty analysis of the two codes. The objective was to develop credible and traceable uncertainty distributions for the input variables of the codes. Expert elicitation, developed independently, was identified as the best technology available for developing a library of uncertainty distributions for the selected consequence parameters. The study was formulated jointly and was limited to the current code models and to physical quantities that could be measured in experiments. To validate the distributions generated for the wet deposition input variables, samples were taken from these distributions and propagated through the wet deposition code model along with the Gaussian plume model (GPM) implemented in the MACCS and COSYMA codes. Resulting distributions closely replicated the aggregated elicited wet deposition distributions. Project teams from the NRC and CEC cooperated successfully to develop and implement a unified process for the elaboration of uncertainty distributions on consequence code input parameters. Formal expert judgment elicitation proved valuable for synthesizing the best available information. Distributions on measurable atmospheric dispersion and deposition parameters were successfully elicited from experts involved in the many phenomenological areas of consequence analysis. This volume is the second of a three-volume document describing the project and contains two appendices describing the rationales for the dispersion and deposition data along with short biographies of the 16 experts who participated in the project
Probabilistic accident consequence uncertainty analysis: Dispersion and deposition uncertainty assessment, appendices A and B

Energy Technology Data Exchange (ETDEWEB)

Harper, F.T.; Young, M.L.; Miller, L.A. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States); Lui, C.H. [Nuclear Regulatory Commission, Washington, DC (United States); Goossens, L.H.J.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Paesler-Sauer, J. [Research Center, Karlsruhe (Germany); Helton, J.C. [and others

1995-01-01

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, completed in 1990, estimate the risks presented by nuclear installations based on postulated frequencies and magnitudes of potential accidents. In 1991, the US Nuclear Regulatory Commission (NRC) and the Commission of the European Communities (CEC) began a joint uncertainty analysis of the two codes. The objective was to develop credible and traceable uncertainty distributions for the input variables of the codes. Expert elicitation, developed independently, was identified as the best technology available for developing a library of uncertainty distributions for the selected consequence parameters. The study was formulated jointly and was limited to the current code models and to physical quantities that could be measured in experiments. To validate the distributions generated for the wet deposition input variables, samples were taken from these distributions and propagated through the wet deposition code model along with the Gaussian plume model (GPM) implemented in the MACCS and COSYMA codes. Resulting distributions closely replicated the aggregated elicited wet deposition distributions. Project teams from the NRC and CEC cooperated successfully to develop and implement a unified process for the elaboration of uncertainty distributions on consequence code input parameters. Formal expert judgment elicitation proved valuable for synthesizing the best available information. Distributions on measurable atmospheric dispersion and deposition parameters were successfully elicited from experts involved in the many phenomenological areas of consequence analysis. This volume is the second of a three-volume document describing the project and contains two appendices describing the rationales for the dispersion and deposition data along with short biographies of the 16 experts who participated in the project.

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