Plutonium explosive dispersal modeling using the MACCS2 computer code

International Nuclear Information System (INIS)

Steele, C.M.; Wald, T.L.; Chanin, D.I.

1998-01-01

The purpose of this paper is to derive the necessary parameters to be used to establish a defensible methodology to perform explosive dispersal modeling of respirable plutonium using Gaussian methods. A particular code, MACCS2, has been chosen for this modeling effort due to its application of sophisticated meteorological statistical sampling in accordance with the philosophy of Nuclear Regulatory Commission (NRC) Regulatory Guide 1.145, ''Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants''. A second advantage supporting the selection of the MACCS2 code for modeling purposes is that meteorological data sets are readily available at most Department of Energy (DOE) and NRC sites. This particular MACCS2 modeling effort focuses on the calculation of respirable doses and not ground deposition. Once the necessary parameters for the MACCS2 modeling are developed and presented, the model is benchmarked against empirical test data from the Double Tracks shot of project Roller Coaster (Shreve 1965) and applied to a hypothetical plutonium explosive dispersal scenario. Further modeling with the MACCS2 code is performed to determine a defensible method of treating the effects of building structure interaction on the respirable fraction distribution as a function of height. These results are related to the Clean Slate 2 and Clean Slate 3 bunkered shots of Project Roller Coaster. Lastly a method is presented to determine the peak 99.5% sector doses on an irregular site boundary in the manner specified in NRC Regulatory Guide 1.145 (1983). Parametric analyses are performed on the major analytic assumptions in the MACCS2 model to define the potential errors that are possible in using this methodology

Plutonium explosive dispersal modeling using the MACCS2 computer code

Energy Technology Data Exchange (ETDEWEB)

Steele, C.M.; Wald, T.L.; Chanin, D.I.

1998-11-01

The purpose of this paper is to derive the necessary parameters to be used to establish a defensible methodology to perform explosive dispersal modeling of respirable plutonium using Gaussian methods. A particular code, MACCS2, has been chosen for this modeling effort due to its application of sophisticated meteorological statistical sampling in accordance with the philosophy of Nuclear Regulatory Commission (NRC) Regulatory Guide 1.145, ``Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants``. A second advantage supporting the selection of the MACCS2 code for modeling purposes is that meteorological data sets are readily available at most Department of Energy (DOE) and NRC sites. This particular MACCS2 modeling effort focuses on the calculation of respirable doses and not ground deposition. Once the necessary parameters for the MACCS2 modeling are developed and presented, the model is benchmarked against empirical test data from the Double Tracks shot of project Roller Coaster (Shreve 1965) and applied to a hypothetical plutonium explosive dispersal scenario. Further modeling with the MACCS2 code is performed to determine a defensible method of treating the effects of building structure interaction on the respirable fraction distribution as a function of height. These results are related to the Clean Slate 2 and Clean Slate 3 bunkered shots of Project Roller Coaster. Lastly a method is presented to determine the peak 99.5% sector doses on an irregular site boundary in the manner specified in NRC Regulatory Guide 1.145 (1983). Parametric analyses are performed on the major analytic assumptions in the MACCS2 model to define the potential errors that are possible in using this methodology.

Regulatory Endorsement Activities for ASME Nuclear Codes and Standards

International Nuclear Information System (INIS)

West, Raymond A.

2006-01-01

The ASME Board on Nuclear Codes and Standards (BNCS) has formed a Task Group on Regulatory Endorsement (TG-RE) that is currently in discussions with the United States Nuclear Regulatory Commission (NRC) to look at suggestions and recommendations that can be used to help with the endorsement of new and revised ASME Nuclear Codes and Standards (NC and S). With the coming of new reactors in the USA in the very near future we need to look at both the regulations and all the ASME NC and S to determine where we need to make changes to support these new plants. At the same time it is important that we maintain our operating plants while addressing ageing management needs of our existing reactors. This is going to take new thinking, time, resources, and money. For all this to take place the regulations and requirements that we use must be clear concise and necessary for safety and to that end both the NRC and ASME are working together to make this happen. Because of the influence that the USA has in the world in dealing with these issues, this paper is written to inform the international nuclear engineering community about the issues and what actions are being addressed under this effort. (author)

The Non-Coding Regulatory RNA Revolution in Archaea

Directory of Open Access Journals (Sweden)

Diego Rivera Gelsinger

2018-03-01

Full Text Available Small non-coding RNAs (sRNAs are ubiquitously found in the three domains of life playing large-scale roles in gene regulation, transposable element silencing and defense against foreign elements. While a substantial body of experimental work has been done to uncover function of sRNAs in Bacteria and Eukarya, the functional roles of sRNAs in Archaea are still poorly understood. Recently, high throughput studies using RNA-sequencing revealed that sRNAs are broadly expressed in the Archaea, comprising thousands of transcripts within the transcriptome during non-challenged and stressed conditions. Antisense sRNAs, which overlap a portion of a gene on the opposite strand (cis-acting, are the most abundantly expressed non-coding RNAs and they can be classified based on their binding patterns to mRNAs (3′ untranslated region (UTR, 5′ UTR, CDS-binding. These antisense sRNAs target many genes and pathways, suggesting extensive roles in gene regulation. Intergenic sRNAs are less abundantly expressed and their targets are difficult to find because of a lack of complete overlap between sRNAs and target mRNAs (trans-acting. While many sRNAs have been validated experimentally, a regulatory role has only been reported for very few of them. Further work is needed to elucidate sRNA-RNA binding mechanisms, the molecular determinants of sRNA-mediated regulation, whether protein components are involved and how sRNAs integrate with complex regulatory networks.

Regulatory context and evolutions - Public Health Code

International Nuclear Information System (INIS)

Rodde, S.

2009-01-01

After having recalled that numerous laws, decrees and orders have been published between 2001 and 2007 due to the transposition of EURATOM directives defining standards for population and worker health protection against dangers resulting from ionizing radiations, the author reviews the regulatory evolutions which occurred in 2008 and 2009, and those currently in progress. They concern the protection of people exposed to radon, the field of radiotherapy, authorizations issued in application of the French public health code. Some decisions are about to be finalized. They concern the activities submitted to a declaration, the modalities of prolongation of the lifetime of sealed sources, a list of apparatus categories the handling of which requires an ability certificate

GIS-assisted spatial analysis for urban regulatory detailed planning: designer's dimension in the Chinese code system

Science.gov (United States)

Yu, Yang; Zeng, Zheng

2009-10-01

By discussing the causes behind the high amendments ratio in the implementation of urban regulatory detailed plans in China despite its law-ensured status, the study aims to reconcile conflict between the legal authority of regulatory detailed planning and the insufficient scientific support in its decision-making and compilation by introducing into the process spatial analysis based on GIS technology and 3D modeling thus present a more scientific and flexible approach to regulatory detailed planning in China. The study first points out that the current compilation process of urban regulatory detailed plan in China employs mainly an empirical approach which renders it constantly subjected to amendments; the study then discusses the need and current utilization of GIS in the Chinese system and proposes the framework of a GIS-assisted 3D spatial analysis process from the designer's perspective which can be regarded as an alternating processes between the descriptive codes and physical design in the compilation of regulatory detailed planning. With a case study of the processes and results from the application of the framework, the paper concludes that the proposed framework can be an effective instrument which provides more rationality, flexibility and thus more efficiency to the compilation and decision-making process of urban regulatory detailed plan in China.

How effective is the revised regulatory code for alcohol advertising in Australia?

Science.gov (United States)

Jones, Sandra C; Hall, Danika; Munro, Geoffrey

2008-01-01

Australia, like several other countries, has a self-regulatory approach to advertising. However, in recent years the effectiveness of the regulatory system has been questioned, and there have been increasing public calls for an overhaul of the system. Following a formal review in 2003, the Ministerial Council on Drug Strategy proposed a revised Alcoholic Beverages Advertising Code (ABAC), which came into operation in 2004. The purpose of the present study was to examine the effectiveness of this revised system. From May 2004 until March 2005 television and magazine advertising campaigns were monitored for alcohol products. Over this period 14 complaints against alcohol advertisements were lodged with the self-regulatory board, and the authors recruited an independent expert panel to assess the advertisements and complaints. In eight of the 14 cases a majority of the judges perceived the advertisement to be in breach of the code, and in no cases did a majority perceive no breach. Conversely, however, none of the complaints were upheld by the Advertising Standards Board (ASB) and only one by the ABAC Panel. The results of this study suggest that the decisions made by the ASB in relation to complaints against alcohol advertisements are not in harmony with the judgement of independent experts, and that the ASB may not be performing an adequate job of representing community standards or protecting the community from offensive or inappropriate advertisements. Further, it appears that the revisions to the ABAC code, and associated processes, have not reduced the problems associated with alcohol advertising in Australia.

Fusion safety codes International modeling with MELCOR and ATHENA- INTRA

CERN Document Server

Marshall, T; Topilski, L; Merrill, B

2002-01-01

For a number of years, the world fusion safety community has been involved in benchmarking their safety analyses codes against experiment data to support regulatory approval of a next step fusion device. This paper discusses the benchmarking of two prominent fusion safety thermal-hydraulic computer codes. The MELCOR code was developed in the US for fission severe accident safety analyses and has been modified for fusion safety analyses. The ATHENA code is a multifluid version of the US-developed RELAP5 code that is also widely used for fusion safety analyses. The ENEA Fusion Division uses ATHENA in conjunction with the INTRA code for its safety analyses. The INTRA code was developed in Germany and predicts containment building pressures, temperatures and fluid flow. ENEA employs the French-developed ISAS system to couple ATHENA and INTRA. This paper provides a brief introduction of the MELCOR and ATHENA-INTRA codes and presents their modeling results for the following breaches of a water cooling line into the...

The MESORAD dose assessment model: Computer code

International Nuclear Information System (INIS)

Ramsdell, J.V.; Athey, G.F.; Bander, T.J.; Scherpelz, R.I.

1988-10-01

MESORAD is a dose equivalent model for emergency response applications that is designed to be run on minicomputers. It has been developed by the Pacific Northwest Laboratory for use as part of the Intermediate Dose Assessment System in the US Nuclear Regulatory Commission Operations Center in Washington, DC, and the Emergency Management System in the US Department of Energy Unified Dose Assessment Center in Richland, Washington. This volume describes the MESORAD computer code and contains a listing of the code. The technical basis for MESORAD is described in the first volume of this report (Scherpelz et al. 1986). A third volume of the documentation planned. That volume will contain utility programs and input and output files that can be used to check the implementation of MESORAD. 18 figs., 4 tabs

SCDAP/RELAP5/MOD3 code development

International Nuclear Information System (INIS)

Allison, C.M.; Siefken, J.L.; Coryell, E.W.

1992-01-01

The SCOAP/RELAP5/MOD3 computer code is designed to describe the overall reactor coolant system (RCS) thermal-hydraulic response, core damage progression, and fission product release and transport during severe accidents. The code is being developed at the Idaho National Engineering Laboratory (INEL) under the primary sponsorship of the Office of Nuclear Regulatory Research of the US Nuclear Regulatory Commission (NRC). Code development activities are currently focused on three main areas - (a) code usability, (b) early phase melt progression model improvements, and (c) advanced reactor thermal-hydraulic model extensions. This paper describes the first two activities. A companion paper describes the advanced reactor model improvements being performed under RELAP5/MOD3 funding

Auto-Regulatory RNA Editing Fine-Tunes mRNA Re-Coding and Complex Behaviour in Drosophila

Science.gov (United States)

Savva, Yiannis A.; Jepson, James E.C; Sahin, Asli; Sugden, Arthur U.; Dorsky, Jacquelyn S.; Alpert, Lauren; Lawrence, Charles; Reenan, Robert A.

2014-01-01

Auto-regulatory feedback loops are a common molecular strategy used to optimize protein function. In Drosophila many mRNAs involved in neuro-transmission are re-coded at the RNA level by the RNA editing enzyme dADAR, leading to the incorporation of amino acids that are not directly encoded by the genome. dADAR also re-codes its own transcript, but the consequences of this auto-regulation in vivo are unclear. Here we show that hard-wiring or abolishing endogenous dADAR auto-regulation dramatically remodels the landscape of re-coding events in a site-specific manner. These molecular phenotypes correlate with altered localization of dADAR within the nuclear compartment. Furthermore, auto-editing exhibits sexually dimorphic patterns of spatial regulation and can be modified by abiotic environmental factors. Finally, we demonstrate that modifying dAdar auto-editing affects adaptive complex behaviors. Our results reveal the in vivo relevance of auto-regulatory control over post-transcriptional mRNA re-coding events in fine-tuning brain function and organismal behavior. PMID:22531175

HLA-E regulatory and coding region variability and haplotypes in a Brazilian population sample.

Science.gov (United States)

Ramalho, Jaqueline; Veiga-Castelli, Luciana C; Donadi, Eduardo A; Mendes-Junior, Celso T; Castelli, Erick C

2017-11-01

The HLA-E gene is characterized by low but wide expression on different tissues. HLA-E is considered a conserved gene, being one of the least polymorphic class I HLA genes. The HLA-E molecule interacts with Natural Killer cell receptors and T lymphocytes receptors, and might activate or inhibit immune responses depending on the peptide associated with HLA-E and with which receptors HLA-E interacts to. Variable sites within the HLA-E regulatory and coding segments may influence the gene function by modifying its expression pattern or encoded molecule, thus, influencing its interaction with receptors and the peptide. Here we propose an approach to evaluate the gene structure, haplotype pattern and the complete HLA-E variability, including regulatory (promoter and 3'UTR) and coding segments (with introns), by using massively parallel sequencing. We investigated the variability of 420 samples from a very admixed population such as Brazilians by using this approach. Considering a segment of about 7kb, 63 variable sites were detected, arranged into 75 extended haplotypes. We detected 37 different promoter sequences (but few frequent ones), 27 different coding sequences (15 representing new HLA-E alleles) and 12 haplotypes at the 3'UTR segment, two of them presenting a summed frequency of 90%. Despite the number of coding alleles, they encode mainly two different full-length molecules, known as E*01:01 and E*01:03, which corresponds to about 90% of all. In addition, differently from what has been previously observed for other non classical HLA genes, the relationship among the HLA-E promoter, coding and 3'UTR haplotypes is not straightforward because the same promoter and 3'UTR haplotypes were many times associated with different HLA-E coding haplotypes. This data reinforces the presence of only two main full-length HLA-E molecules encoded by the many HLA-E alleles detected in our population sample. In addition, this data does indicate that the distal HLA-E promoter is by

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

Neural model of gene regulatory network: a survey on supportive meta-heuristics.

Science.gov (United States)

Biswas, Surama; Acharyya, Sriyankar

2016-06-01

Gene regulatory network (GRN) is produced as a result of regulatory interactions between different genes through their coded proteins in cellular context. Having immense importance in disease detection and drug finding, GRN has been modelled through various mathematical and computational schemes and reported in survey articles. Neural and neuro-fuzzy models have been the focus of attraction in bioinformatics. Predominant use of meta-heuristic algorithms in training neural models has proved its excellence. Considering these facts, this paper is organized to survey neural modelling schemes of GRN and the efficacy of meta-heuristic algorithms towards parameter learning (i.e. weighting connections) within the model. This survey paper renders two different structure-related approaches to infer GRN which are global structure approach and substructure approach. It also describes two neural modelling schemes, such as artificial neural network/recurrent neural network based modelling and neuro-fuzzy modelling. The meta-heuristic algorithms applied so far to learn the structure and parameters of neutrally modelled GRN have been reviewed here.

Description of codes and models to be used in risk assessment

International Nuclear Information System (INIS)

1991-09-01

Human health and environmental risk assessments will be performed as part of the Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA) remedial investigation/feasibility study (RI/FS) activities at the Hanford Site. Analytical and computer encoded numerical models are commonly used during both the remedial investigation (RI) and feasibility study (FS) to predict or estimate the concentration of contaminants at the point of exposure to humans and/or the environment. This document has been prepared to identify the computer codes that will be used in support of RI/FS human health and environmental risk assessments at the Hanford Site. In addition to the CERCLA RI/FS process, it is recommended that these computer codes be used when fate and transport analyses is required for other activities. Additional computer codes may be used for other purposes (e.g., design of tracer tests, location of observation wells, etc.). This document provides guidance for unit managers in charge of RI/FS activities. Use of the same computer codes for all analytical activities at the Hanford Site will promote consistency, reduce the effort required to develop, validate, and implement models to simulate Hanford Site conditions, and expedite regulatory review. The discussion provides a description of how models will likely be developed and utilized at the Hanford Site. It is intended to summarize previous environmental-related modeling at the Hanford Site and provide background for future model development. The modeling capabilities that are desirable for the Hanford Site and the codes that were evaluated. The recommendations include the codes proposed to support future risk assessment modeling at the Hanford Site, and provides the rational for the codes selected. 27 refs., 3 figs., 1 tab

SCDAP/RELAP5 code development and assessment

International Nuclear Information System (INIS)

Allison, C.M.; Hohorst, J.K.

1996-01-01

The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The current version of the code is SCDAP/RELAP5/MOD3.1e. Although MOD3.1e contains a number of significant improvements since the initial version of MOD3.1 was released, new models to treat the behavior of the fuel and cladding during reflood have had the most dramatic impact on the code's calculations. This paper provides a brief description of the new reflood models, presents highlights of the assessment of the current version of MOD3.1, and discusses future SCDAP/RELAP5/MOD3.2 model development activities

Model checking optimal finite-horizon control for probabilistic gene regulatory networks.

Science.gov (United States)

Wei, Ou; Guo, Zonghao; Niu, Yun; Liao, Wenyuan

2017-12-14

Probabilistic Boolean networks (PBNs) have been proposed for analyzing external control in gene regulatory networks with incorporation of uncertainty. A context-sensitive PBN with perturbation (CS-PBNp), extending a PBN with context-sensitivity to reflect the inherent biological stability and random perturbations to express the impact of external stimuli, is considered to be more suitable for modeling small biological systems intervened by conditions from the outside. In this paper, we apply probabilistic model checking, a formal verification technique, to optimal control for a CS-PBNp that minimizes the expected cost over a finite control horizon. We first describe a procedure of modeling a CS-PBNp using the language provided by a widely used probabilistic model checker PRISM. We then analyze the reward-based temporal properties and the computation in probabilistic model checking; based on the analysis, we provide a method to formulate the optimal control problem as minimum reachability reward properties. Furthermore, we incorporate control and state cost information into the PRISM code of a CS-PBNp such that automated model checking a minimum reachability reward property on the code gives the solution to the optimal control problem. We conduct experiments on two examples, an apoptosis network and a WNT5A network. Preliminary experiment results show the feasibility and effectiveness of our approach. The approach based on probabilistic model checking for optimal control avoids explicit computation of large-size state transition relations associated with PBNs. It enables a natural depiction of the dynamics of gene regulatory networks, and provides a canonical form to formulate optimal control problems using temporal properties that can be automated solved by leveraging the analysis power of underlying model checking engines. This work will be helpful for further utilization of the advances in formal verification techniques in system biology.

Direct containment heating models in the CONTAIN code

International Nuclear Information System (INIS)

Washington, K.E.; Williams, D.C.

1995-08-01

The potential exists in a nuclear reactor core melt severe accident for molten core debris to be dispersed under high pressure into the containment building. If this occurs, the set of phenomena that result in the transfer of energy to the containment atmosphere and its surroundings is referred to as direct containment heating (DCH). Because of the potential for DCH to lead to early containment failure, the U.S. Nuclear Regulatory Commission (USNRC) has sponsored an extensive research program consisting of experimental, analytical, and risk integration components. An important element of the analytical research has been the development and assessment of direct containment heating models in the CONTAIN code. This report documents the DCH models in the CONTAIN code. DCH models in CONTAIN for representing debris transport, trapping, chemical reactions, and heat transfer from debris to the containment atmosphere and surroundings are described. The descriptions include the governing equations and input instructions in CONTAIN unique to performing DCH calculations. Modifications made to the combustion models in CONTAIN for representing the combustion of DCH-produced and pre-existing hydrogen under DCH conditions are also described. Input table options for representing the discharge of debris from the RPV and the entrainment phase of the DCH process are also described. A sample calculation is presented to demonstrate the functionality of the models. The results show that reasonable behavior is obtained when the models are used to predict the sixth Zion geometry integral effects test at 1/10th scale

Direct containment heating models in the CONTAIN code

Energy Technology Data Exchange (ETDEWEB)

Washington, K.E.; Williams, D.C.

1995-08-01

The potential exists in a nuclear reactor core melt severe accident for molten core debris to be dispersed under high pressure into the containment building. If this occurs, the set of phenomena that result in the transfer of energy to the containment atmosphere and its surroundings is referred to as direct containment heating (DCH). Because of the potential for DCH to lead to early containment failure, the U.S. Nuclear Regulatory Commission (USNRC) has sponsored an extensive research program consisting of experimental, analytical, and risk integration components. An important element of the analytical research has been the development and assessment of direct containment heating models in the CONTAIN code. This report documents the DCH models in the CONTAIN code. DCH models in CONTAIN for representing debris transport, trapping, chemical reactions, and heat transfer from debris to the containment atmosphere and surroundings are described. The descriptions include the governing equations and input instructions in CONTAIN unique to performing DCH calculations. Modifications made to the combustion models in CONTAIN for representing the combustion of DCH-produced and pre-existing hydrogen under DCH conditions are also described. Input table options for representing the discharge of debris from the RPV and the entrainment phase of the DCH process are also described. A sample calculation is presented to demonstrate the functionality of the models. The results show that reasonable behavior is obtained when the models are used to predict the sixth Zion geometry integral effects test at 1/10th scale.

Development of multipurpose regulatory PSA model

International Nuclear Information System (INIS)

Lee, Chang Ju; Sung, Key Yong; Kim, Hho Jung; Yang, Joon Eon; Ha, Jae Joo

2004-01-01

Generally, risk information for nuclear facilities comes from the results of Probabilistic safety assessment (PSA). PSA is a systematic tool to ensure the safety of nuclear facilities, since it is based on thorough and consistent application of probability models. In particular, the PSA has been widely utilized for risk-informed regulation (RIR), including various licensee-initiated risk-informed applications (RIA). In any regulatory decision, the main goal is to make a sound safety decision based on technically defensible information. Also, due to the increased public requests for giving a safety guarantee, the regulator should provide the visible means of safety. The use of PSA by the regulator can give the answer on this problem. Therefore, in order to study the applicability of risk information for regulatory safety management, it is a demanding task to prepare a well-established regulatory PSA model and tool. In 2002, KINS and KAERI together made a research cooperation to form a working group to develop the regulatory PSA model - so-called MPAS model. The MPAS stands for multipurpose probabilistic analysis of safety. For instance, a role of the MPAS model is to give some risk insights in the preparation of various regulatory programs. Another role of this model is to provide an independent risk information to the regulator during regulatory decision-making, not depending on the licensee's information

A regulatory code for neuron-specific odor receptor expression.

Directory of Open Access Journals (Sweden)

Anandasankar Ray

2008-05-01

Full Text Available Olfactory receptor neurons (ORNs must select-from a large repertoire-which odor receptors to express. In Drosophila, most ORNs express one of 60 Or genes, and most Or genes are expressed in a single ORN class in a process that produces a stereotyped receptor-to-neuron map. The construction of this map poses a problem of receptor gene regulation that is remarkable in its dimension and about which little is known. By using a phylogenetic approach and the genome sequences of 12 Drosophila species, we systematically identified regulatory elements that are evolutionarily conserved and specific for individual Or genes of the maxillary palp. Genetic analysis of these elements supports a model in which each receptor gene contains a zip code, consisting of elements that act positively to promote expression in a subset of ORN classes, and elements that restrict expression to a single ORN class. We identified a transcription factor, Scalloped, that mediates repression. Some elements are used in other chemosensory organs, and some are conserved upstream of axon-guidance genes. Surprisingly, the odor response spectra and organization of maxillary palp ORNs have been extremely well-conserved for tens of millions of years, even though the amino acid sequences of the receptors are not highly conserved. These results, taken together, define the logic by which individual ORNs in the maxillary palp select which odor receptors to express.

Evolutionary analysis reveals regulatory and functional landscape of coding and non-coding RNA editing.

Science.gov (United States)

Zhang, Rui; Deng, Patricia; Jacobson, Dionna; Li, Jin Billy

2017-02-01

Adenosine-to-inosine RNA editing diversifies the transcriptome and promotes functional diversity, particularly in the brain. A plethora of editing sites has been recently identified; however, how they are selected and regulated and which are functionally important are largely unknown. Here we show the cis-regulation and stepwise selection of RNA editing during Drosophila evolution and pinpoint a large number of functional editing sites. We found that the establishment of editing and variation in editing levels across Drosophila species are largely explained and predicted by cis-regulatory elements. Furthermore, editing events that arose early in the species tree tend to be more highly edited in clusters and enriched in slowly-evolved neuronal genes, thus suggesting that the main role of RNA editing is for fine-tuning neurological functions. While nonsynonymous editing events have been long recognized as playing a functional role, in addition to nonsynonymous editing sites, a large fraction of 3'UTR editing sites is evolutionarily constrained, highly edited, and thus likely functional. We find that these 3'UTR editing events can alter mRNA stability and affect miRNA binding and thus highlight the functional roles of noncoding RNA editing. Our work, through evolutionary analyses of RNA editing in Drosophila, uncovers novel insights of RNA editing regulation as well as its functions in both coding and non-coding regions.

TASS code topical report. V.1 TASS code technical manual

International Nuclear Information System (INIS)

Sim, Suk K.; Chang, W. P.; Kim, K. D.; Kim, H. C.; Yoon, H. Y.

1997-02-01

TASS 1.0 code has been developed at KAERI for the initial and reload non-LOCA safety analysis for the operating PWRs as well as the PWRs under construction in Korea. TASS code will replace various vendor's non-LOCA safety analysis codes currently used for the Westinghouse and ABB-CE type PWRs in Korea. This can be achieved through TASS code input modifications specific to each reactor type. The TASS code can be run interactively through the keyboard operation. A simimodular configuration used in developing the TASS code enables the user easily implement new models. TASS code has been programmed using FORTRAN77 which makes it easy to install and port for different computer environments. The TASS code can be utilized for the steady state simulation as well as the non-LOCA transient simulations such as power excursions, reactor coolant pump trips, load rejections, loss of feedwater, steam line breaks, steam generator tube ruptures, rod withdrawal and drop, and anticipated transients without scram (ATWS). The malfunctions of the control systems, components, operator actions and the transients caused by the malfunctions can be easily simulated using the TASS code. This technical report describes the TASS 1.0 code models including reactor thermal hydraulic, reactor core and control models. This TASS code models including reactor thermal hydraulic, reactor core and control models. This TASS code technical manual has been prepared as a part of the TASS code manual which includes TASS code user's manual and TASS code validation report, and will be submitted to the regulatory body as a TASS code topical report for a licensing non-LOCA safety analysis for the Westinghouse and ABB-CE type PWRs operating and under construction in Korea. (author). 42 refs., 29 tabs., 32 figs

Impacts of Model Building Energy Codes

Energy Technology Data Exchange (ETDEWEB)

Athalye, Rahul A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sivaraman, Deepak [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Elliott, Douglas B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Liu, Bing [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bartlett, Rosemarie [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2016-10-31

The U.S. Department of Energy (DOE) Building Energy Codes Program (BECP) periodically evaluates national and state-level impacts associated with energy codes in residential and commercial buildings. Pacific Northwest National Laboratory (PNNL), funded by DOE, conducted an assessment of the prospective impacts of national model building energy codes from 2010 through 2040. A previous PNNL study evaluated the impact of the Building Energy Codes Program; this study looked more broadly at overall code impacts. This report describes the methodology used for the assessment and presents the impacts in terms of energy savings, consumer cost savings, and reduced CO₂ emissions at the state level and at aggregated levels. This analysis does not represent all potential savings from energy codes in the U.S. because it excludes several states which have codes which are fundamentally different from the national model energy codes or which do not have state-wide codes. Energy codes follow a three-phase cycle that starts with the development of a new model code, proceeds with the adoption of the new code by states and local jurisdictions, and finishes when buildings comply with the code. The development of new model code editions creates the potential for increased energy savings. After a new model code is adopted, potential savings are realized in the field when new buildings (or additions and alterations) are constructed to comply with the new code. Delayed adoption of a model code and incomplete compliance with the code’s requirements erode potential savings. The contributions of all three phases are crucial to the overall impact of codes, and are considered in this assessment.

PEAR code review

International Nuclear Information System (INIS)

De Wit, R.; Jamieson, T.; Lord, M.; Lafortune, J.F.

1997-07-01

As a necessary component in the continuous improvement and refinement of methodologies employed in the nuclear industry, regulatory agencies need to periodically evaluate these processes to improve confidence in results and ensure appropriate levels of safety are being achieved. The independent and objective review of industry-standard computer codes forms an essential part of this program. To this end, this work undertakes an in-depth review of the computer code PEAR (Public Exposures from Accidental Releases), developed by Atomic Energy of Canada Limited (AECL) to assess accidental releases from CANDU reactors. PEAR is based largely on the models contained in the Canadian Standards Association (CSA) N288.2-M91. This report presents the results of a detailed technical review of the PEAR code to identify any variations from the CSA standard and other supporting documentation, verify the source code, assess the quality of numerical models and results, and identify general strengths and weaknesses of the code. The version of the code employed in this review is the one which AECL intends to use for CANDU 9 safety analyses. (author)

Fatigue modelling according to the JCSS Probabilistic model code

NARCIS (Netherlands)

Vrouwenvelder, A.C.W.M.

2007-01-01

The Joint Committee on Structural Safety is working on a Model Code for full probabilistic design. The code consists out of three major parts: Basis of design, Load Models and Models for Material and Structural Properties. The code is intended as the operational counter part of codes like ISO,

Pump Component Model in SPACE Code

International Nuclear Information System (INIS)

Kim, Byoung Jae; Kim, Kyoung Doo

2010-08-01

This technical report describes the pump component model in SPACE code. A literature survey was made on pump models in existing system codes. The models embedded in SPACE code were examined to check the confliction with intellectual proprietary rights. Design specifications, computer coding implementation, and test results are included in this report

Studies on DANESS Code Modeling

International Nuclear Information System (INIS)

Jeong, Chang Joon

2009-09-01

The DANESS code modeling study has been performed. DANESS code is widely used in a dynamic fuel cycle analysis. Korea Atomic Energy Research Institute (KAERI) has used the DANESS code for the Korean national nuclear fuel cycle scenario analysis. In this report, the important models such as Energy-demand scenario model, New Reactor Capacity Decision Model, Reactor and Fuel Cycle Facility History Model, and Fuel Cycle Model are investigated. And, some models in the interface module are refined and inserted for Korean nuclear fuel cycle model. Some application studies have also been performed for GNEP cases and for US fast reactor scenarios with various conversion ratios

Regulatory Models and the Environment: Practice, Pitfalls, and Prospects

Energy Technology Data Exchange (ETDEWEB)

Holmes, K. John; Graham, Judith A.; McKone, Thomas; Whipple, Chris

2008-06-01

Computational models support environmental regulatory activities by providing the regulator an ability to evaluate available knowledge, assess alternative regulations, and provide a framework to assess compliance. But all models face inherent uncertainties, because human and natural systems are always more complex and heterogeneous than can be captured in a model. Here we provide a summary discussion of the activities, findings, and recommendations of the National Research Council's Committee on Regulatory Environmental Models, a committee funded by the US Environmental Protection Agency to provide guidance on the use of computational models in the regulatory process. Modeling is a difficult enterprise even outside of the potentially adversarial regulatory environment. The demands grow when the regulatory requirements for accountability, transparency, public accessibility, and technical rigor are added to the challenges. Moreover, models cannot be validated (declared true) but instead should be evaluated with regard to their suitability as tools to address a specific question. The committee concluded that these characteristics make evaluation of a regulatory model more complex than simply comparing measurement data with model results. Evaluation also must balance the need for a model to be accurate with the need for a model to be reproducible, transparent, and useful for the regulatory decision at hand. Meeting these needs requires model evaluation to be applied over the"life cycle" of a regulatory model with an approach that includes different forms of peer review, uncertainty analysis, and extrapolation methods than for non-regulatory models.

A deeper look into transcription regulatory code by preferred pair distance templates for transcription factor binding sites

KAUST Repository

Kulakovskiy, Ivan V.

2011-08-18

Motivation: Modern experimental methods provide substantial information on protein-DNA recognition. Studying arrangements of transcription factor binding sites (TFBSs) of interacting transcription factors (TFs) advances understanding of the transcription regulatory code. Results: We constructed binding motifs for TFs forming a complex with HIF-1α at the erythropoietin 3\\'-enhancer. Corresponding TFBSs were predicted in the segments around transcription start sites (TSSs) of all human genes. Using the genome-wide set of regulatory regions, we observed several strongly preferred distances between hypoxia-responsive element (HRE) and binding sites of a particular cofactor protein. The set of preferred distances was called as a preferred pair distance template (PPDT). PPDT dramatically depended on the TF and orientation of its binding sites relative to HRE. PPDT evaluated from the genome-wide set of regulatory sequences was used to detect significant PPDT-consistent binding site pairs in regulatory regions of hypoxia-responsive genes. We believe PPDT can help to reveal the layout of eukaryotic regulatory segments. © The Author 2011. Published by Oxford University Press. All rights reserved.

Current approaches to gene regulatory network modelling

Directory of Open Access Journals (Sweden)

Brazma Alvis

2007-09-01

Full Text Available Abstract Many different approaches have been developed to model and simulate gene regulatory networks. We proposed the following categories for gene regulatory network models: network parts lists, network topology models, network control logic models, and dynamic models. Here we will describe some examples for each of these categories. We will study the topology of gene regulatory networks in yeast in more detail, comparing a direct network derived from transcription factor binding data and an indirect network derived from genome-wide expression data in mutants. Regarding the network dynamics we briefly describe discrete and continuous approaches to network modelling, then describe a hybrid model called Finite State Linear Model and demonstrate that some simple network dynamics can be simulated in this model.

Detecting non-coding selective pressure in coding regions

Directory of Open Access Journals (Sweden)

Blanchette Mathieu

2007-02-01

Full Text Available Abstract Background Comparative genomics approaches, where orthologous DNA regions are compared and inter-species conserved regions are identified, have proven extremely powerful for identifying non-coding regulatory regions located in intergenic or intronic regions. However, non-coding functional elements can also be located within coding region, as is common for exonic splicing enhancers, some transcription factor binding sites, and RNA secondary structure elements affecting mRNA stability, localization, or translation. Since these functional elements are located in regions that are themselves highly conserved because they are coding for a protein, they generally escaped detection by comparative genomics approaches. Results We introduce a comparative genomics approach for detecting non-coding functional elements located within coding regions. Codon evolution is modeled as a mixture of codon substitution models, where each component of the mixture describes the evolution of codons under a specific type of coding selective pressure. We show how to compute the posterior distribution of the entropy and parsimony scores under this null model of codon evolution. The method is applied to a set of growth hormone 1 orthologous mRNA sequences and a known exonic splicing elements is detected. The analysis of a set of CORTBP2 orthologous genes reveals a region of several hundred base pairs under strong non-coding selective pressure whose function remains unknown. Conclusion Non-coding functional elements, in particular those involved in post-transcriptional regulation, are likely to be much more prevalent than is currently known. With the numerous genome sequencing projects underway, comparative genomics approaches like that proposed here are likely to become increasingly powerful at detecting such elements.

Improvement of MARS code reflood model

International Nuclear Information System (INIS)

Hwang, Moonkyu; Chung, Bub-Dong

2011-01-01

A specifically designed heat transfer model for the reflood process which normally occurs at low flow and low pressure was originally incorporated in the MARS code. The model is essentially identical to that of the RELAP5/MOD3.3 code. The model, however, is known to have under-estimated the peak cladding temperature (PCT) with earlier turn-over. In this study, the original MARS code reflood model is improved. Based on the extensive sensitivity studies for both hydraulic and wall heat transfer models, it is found that the dispersed flow film boiling (DFFB) wall heat transfer is the most influential process determining the PCT, whereas the interfacial drag model most affects the quenching time through the liquid carryover phenomenon. The model proposed by Bajorek and Young is incorporated for the DFFB wall heat transfer. Both space grid and droplet enhancement models are incorporated. Inverted annular film boiling (IAFB) is modeled by using the original PSI model of the code. The flow transition between the DFFB and IABF, is modeled using the TRACE code interpolation. A gas velocity threshold is also added to limit the top-down quenching effect. Assessment calculations are performed for the original and modified MARS codes for the Flecht-Seaset test and RBHT test. Improvements are observed in terms of the PCT and quenching time predictions in the Flecht-Seaset assessment. In case of the RBHT assessment, the improvement over the original MARS code is found marginal. A space grid effect, however, is clearly seen from the modified version of the MARS code. (author)

Implications of duplicated cis-regulatory elements in the evolution of metazoans: the DDI model or how simplicity begets novelty.

Science.gov (United States)

Jiménez-Delgado, Senda; Pascual-Anaya, Juan; Garcia-Fernàndez, Jordi

2009-07-01

The discovery that most regulatory genes were conserved among animals from distant phyla challenged the ideas that gene duplication and divergence of homologous coding sequences were the basis for major morphological changes in metazoan evolution. In recent years, however, the interest for the roles, conservation and changes of non-coding sequences grew-up in parallel with genome sequencing projects. Presently, many independent studies are highlighting the importance that subtle changes in cis-regulatory regions had in the evolution of morphology trough the Animal Kingdom. Here we will show and discuss some of these studies, and underscore the future of cis-Evo-Devo research. Nevertheless, we would also explore how gene duplication, which includes duplication of regulatory regions, may have been critical for spatial or temporal co-option of new regulatory networks, causing the deployment of new transcriptome scenarios, and how these induced morphological changes were critical for the evolution of new forms. Forty years after Susumu Ohno famous sentence 'natural selection merely modifies, while redundancy creates', we suggest the alternative: 'natural selection modifies, while redundancy of cis-regulatory elements innovates', and propose the Duplication-Degeneration-Innovation model to explain the increased evolvability of duplicated cis-regulatory regions. Paradoxically, making regulation simpler by subfunctionalization paved the path for future complexity or, in other words, 'to make it simple to make it complex'.

Comprehensive Reconstruction and Visualization of Non-Coding Regulatory Networks in Human

Science.gov (United States)

Bonnici, Vincenzo; Russo, Francesco; Bombieri, Nicola; Pulvirenti, Alfredo; Giugno, Rosalba

2014-01-01

Research attention has been powered to understand the functional roles of non-coding RNAs (ncRNAs). Many studies have demonstrated their deregulation in cancer and other human disorders. ncRNAs are also present in extracellular human body fluids such as serum and plasma, giving them a great potential as non-invasive biomarkers. However, non-coding RNAs have been relatively recently discovered and a comprehensive database including all of them is still missing. Reconstructing and visualizing the network of ncRNAs interactions are important steps to understand their regulatory mechanism in complex systems. This work presents ncRNA-DB, a NoSQL database that integrates ncRNAs data interactions from a large number of well established on-line repositories. The interactions involve RNA, DNA, proteins, and diseases. ncRNA-DB is available at http://ncrnadb.scienze.univr.it/ncrnadb/. It is equipped with three interfaces: web based, command-line, and a Cytoscape app called ncINetView. By accessing only one resource, users can search for ncRNAs and their interactions, build a network annotated with all known ncRNAs and associated diseases, and use all visual and mining features available in Cytoscape. PMID:25540777

Comprehensive reconstruction and visualization of non-coding regulatory networks in human.

Science.gov (United States)

Bonnici, Vincenzo; Russo, Francesco; Bombieri, Nicola; Pulvirenti, Alfredo; Giugno, Rosalba

2014-01-01

Research attention has been powered to understand the functional roles of non-coding RNAs (ncRNAs). Many studies have demonstrated their deregulation in cancer and other human disorders. ncRNAs are also present in extracellular human body fluids such as serum and plasma, giving them a great potential as non-invasive biomarkers. However, non-coding RNAs have been relatively recently discovered and a comprehensive database including all of them is still missing. Reconstructing and visualizing the network of ncRNAs interactions are important steps to understand their regulatory mechanism in complex systems. This work presents ncRNA-DB, a NoSQL database that integrates ncRNAs data interactions from a large number of well established on-line repositories. The interactions involve RNA, DNA, proteins, and diseases. ncRNA-DB is available at http://ncrnadb.scienze.univr.it/ncrnadb/. It is equipped with three interfaces: web based, command-line, and a Cytoscape app called ncINetView. By accessing only one resource, users can search for ncRNAs and their interactions, build a network annotated with all known ncRNAs and associated diseases, and use all visual and mining features available in Cytoscape.

76 FR 21932 - Self-Regulatory Organizations; Financial Industry Regulatory Authority, Inc.; Order Granting...

Science.gov (United States)

2011-04-19

... statement therein, as follows: I. Introduction On February 4, 2011, the Financial Industry Regulatory...-Regulatory Organizations; Financial Industry Regulatory Authority, Inc.; Order Granting Approval of a... Financial Industry Regulatory Authority, Inc. (``FINRA'') to amend Rule 13806 of the Code of Arbitration...

A Positive Regulatory Loop between a Wnt-Regulated Non-coding RNA and ASCL2 Controls Intestinal Stem Cell Fate.

Science.gov (United States)

Giakountis, Antonis; Moulos, Panagiotis; Zarkou, Vasiliki; Oikonomou, Christina; Harokopos, Vaggelis; Hatzigeorgiou, Artemis G; Reczko, Martin; Hatzis, Pantelis

2016-06-21

The canonical Wnt pathway plays a central role in stem cell maintenance, differentiation, and proliferation in the intestinal epithelium. Constitutive, aberrant activity of the TCF4/β-catenin transcriptional complex is the primary transforming factor in colorectal cancer. We identify a nuclear long non-coding RNA, termed WiNTRLINC1, as a direct target of TCF4/β-catenin in colorectal cancer cells. WiNTRLINC1 positively regulates the expression of its genomic neighbor ASCL2, a transcription factor that controls intestinal stem cell fate. WiNTRLINC1 interacts with TCF4/β-catenin to mediate the juxtaposition of its promoter with the regulatory regions of ASCL2. ASCL2, in turn, regulates WiNTRLINC1 transcriptionally, closing a feedforward regulatory loop that controls stem cell-related gene expression. This regulatory circuitry is highly amplified in colorectal cancer and correlates with increased metastatic potential and decreased patient survival. Our results uncover the interplay between non-coding RNA-mediated regulation and Wnt signaling and point to the diagnostic and therapeutic potential of WiNTRLINC1. Copyright © 2016 The Author(s). Published by Elsevier Inc. All rights reserved.

Coding with partially hidden Markov models

DEFF Research Database (Denmark)

Forchhammer, Søren; Rissanen, J.

1995-01-01

Partially hidden Markov models (PHMM) are introduced. They are a variation of the hidden Markov models (HMM) combining the power of explicit conditioning on past observations and the power of using hidden states. (P)HMM may be combined with arithmetic coding for lossless data compression. A general...... 2-part coding scheme for given model order but unknown parameters based on PHMM is presented. A forward-backward reestimation of parameters with a redefined backward variable is given for these models and used for estimating the unknown parameters. Proof of convergence of this reestimation is given....... The PHMM structure and the conditions of the convergence proof allows for application of the PHMM to image coding. Relations between the PHMM and hidden Markov models (HMM) are treated. Results of coding bi-level images with the PHMM coding scheme is given. The results indicate that the PHMM can adapt...

XSOR codes users manual

International Nuclear Information System (INIS)

Jow, Hong-Nian; Murfin, W.B.; Johnson, J.D.

1993-11-01

This report describes the source term estimation codes, XSORs. The codes are written for three pressurized water reactors (Surry, Sequoyah, and Zion) and two boiling water reactors (Peach Bottom and Grand Gulf). The ensemble of codes has been named ''XSOR''. The purpose of XSOR codes is to estimate the source terms which would be released to the atmosphere in severe accidents. A source term includes the release fractions of several radionuclide groups, the timing and duration of releases, the rates of energy release, and the elevation of releases. The codes have been developed by Sandia National Laboratories for the US Nuclear Regulatory Commission (NRC) in support of the NUREG-1150 program. The XSOR codes are fast running parametric codes and are used as surrogates for detailed mechanistic codes. The XSOR codes also provide the capability to explore the phenomena and their uncertainty which are not currently modeled by the mechanistic codes. The uncertainty distributions of input parameters may be used by an. XSOR code to estimate the uncertainty of source terms

Computer simulations of a generic truck cask in a regulatory fire using the Container Analysis Fire Environment (CAFE) code

International Nuclear Information System (INIS)

Ju, H.; Greiner, M.; Suo-Anttila, A.

2002-01-01

The Container Analysis Fire Environment (CAFE) computer code is designed to predict accurately convection and radiation heat transfer to a thermally massive object engulfed in a large pool fire. It is well suited for design and risk analyses of spent nuclear fuel transport systems. CAFE employs computational fluid dynamics and several fire and radiation models. These models must be benchmarked using experimental results. In this paper, a set of wind velocity conditions are determined which allow CAFE accurately to reproduce recent heat transfer measurements for a thick walled calorimeter in a ST-1 regulatory pool fire. CAFE is then used to predict the response of an intack (thin walled) generic legal weight truck cask. The maximum temperatures reached by internal components are within safe limits. A simple 800 deg. C, grey-radiation fire model gives maximum component temperatures that are somewhat below those predicted by CAFE. (author)

Gap Conductance model Validation in the TASS/SMR-S code using MARS code

International Nuclear Information System (INIS)

Ahn, Sang Jun; Yang, Soo Hyung; Chung, Young Jong; Lee, Won Jae

2010-01-01

Korea Atomic Energy Research Institute (KAERI) has been developing the TASS/SMR-S (Transient and Setpoint Simulation/Small and Medium Reactor) code, which is a thermal hydraulic code for the safety analysis of the advanced integral reactor. An appropriate work to validate the applicability of the thermal hydraulic models within the code should be demanded. Among the models, the gap conductance model which is describes the thermal gap conductivity between fuel and cladding was validated through the comparison with MARS code. The validation of the gap conductance model was performed by evaluating the variation of the gap temperature and gap width as the changed with the power fraction. In this paper, a brief description of the gap conductance model in the TASS/SMR-S code is presented. In addition, calculated results to validate the gap conductance model are demonstrated by comparing with the results of the MARS code with the test case

Sequence-based model of gap gene regulatory network.

Science.gov (United States)

Kozlov, Konstantin; Gursky, Vitaly; Kulakovskiy, Ivan; Samsonova, Maria

2014-01-01

The detailed analysis of transcriptional regulation is crucially important for understanding biological processes. The gap gene network in Drosophila attracts large interest among researches studying mechanisms of transcriptional regulation. It implements the most upstream regulatory layer of the segmentation gene network. The knowledge of molecular mechanisms involved in gap gene regulation is far less complete than that of genetics of the system. Mathematical modeling goes beyond insights gained by genetics and molecular approaches. It allows us to reconstruct wild-type gene expression patterns in silico, infer underlying regulatory mechanism and prove its sufficiency. We developed a new model that provides a dynamical description of gap gene regulatory systems, using detailed DNA-based information, as well as spatial transcription factor concentration data at varying time points. We showed that this model correctly reproduces gap gene expression patterns in wild type embryos and is able to predict gap expression patterns in Kr mutants and four reporter constructs. We used four-fold cross validation test and fitting to random dataset to validate the model and proof its sufficiency in data description. The identifiability analysis showed that most model parameters are well identifiable. We reconstructed the gap gene network topology and studied the impact of individual transcription factor binding sites on the model output. We measured this impact by calculating the site regulatory weight as a normalized difference between the residual sum of squares error for the set of all annotated sites and for the set with the site of interest excluded. The reconstructed topology of the gap gene network is in agreement with previous modeling results and data from literature. We showed that 1) the regulatory weights of transcription factor binding sites show very weak correlation with their PWM score; 2) sites with low regulatory weight are important for the model output; 3

User's manual for a process model code

International Nuclear Information System (INIS)

Kern, E.A.; Martinez, D.P.

1981-03-01

The MODEL code has been developed for computer modeling of materials processing facilities associated with the nuclear fuel cycle. However, it can also be used in other modeling applications. This report provides sufficient information for a potential user to apply the code to specific process modeling problems. Several examples that demonstrate most of the capabilities of the code are provided

The GNASH preequilibrium-statistical nuclear model code

International Nuclear Information System (INIS)

Arthur, E. D.

1988-01-01

The following report is based on materials presented in a series of lectures at the International Center for Theoretical Physics, Trieste, which were designed to describe the GNASH preequilibrium statistical model code and its use. An overview is provided of the code with emphasis upon code's calculational capabilities and the theoretical models that have been implemented in it. Two sample problems are discussed, the first dealing with neutron reactions on 58 Ni. the second illustrates the fission model capabilities implemented in the code and involves n + 235 U reactions. Finally a description is provided of current theoretical model and code development underway. Examples of calculated results using these new capabilities are also given. 19 refs., 17 figs., 3 tabs

Rapid installation of numerical models in multiple parent codes

Energy Technology Data Exchange (ETDEWEB)

Brannon, R.M.; Wong, M.K.

1996-10-01

A set of``model interface guidelines``, called MIG, is offered as a means to more rapidly install numerical models (such as stress-strain laws) into any parent code (hydrocode, finite element code, etc.) without having to modify the model subroutines. The model developer (who creates the model package in compliance with the guidelines) specifies the model`s input and storage requirements in a standardized way. For portability, database management (such as saving user inputs and field variables) is handled by the parent code. To date, NUG has proved viable in beta installations of several diverse models in vectorized and parallel codes written in different computer languages. A NUG-compliant model can be installed in different codes without modifying the model`s subroutines. By maintaining one model for many codes, MIG facilitates code-to-code comparisons and reduces duplication of effort potentially reducing the cost of installing and sharing models.

ChIPBase v2.0: decoding transcriptional regulatory networks of non-coding RNAs and protein-coding genes from ChIP-seq data.

Science.gov (United States)

Zhou, Ke-Ren; Liu, Shun; Sun, Wen-Ju; Zheng, Ling-Ling; Zhou, Hui; Yang, Jian-Hua; Qu, Liang-Hu

2017-01-04

The abnormal transcriptional regulation of non-coding RNAs (ncRNAs) and protein-coding genes (PCGs) is contributed to various biological processes and linked with human diseases, but the underlying mechanisms remain elusive. In this study, we developed ChIPBase v2.0 (http://rna.sysu.edu.cn/chipbase/) to explore the transcriptional regulatory networks of ncRNAs and PCGs. ChIPBase v2.0 has been expanded with ∼10 200 curated ChIP-seq datasets, which represent about 20 times expansion when comparing to the previous released version. We identified thousands of binding motif matrices and their binding sites from ChIP-seq data of DNA-binding proteins and predicted millions of transcriptional regulatory relationships between transcription factors (TFs) and genes. We constructed 'Regulator' module to predict hundreds of TFs and histone modifications that were involved in or affected transcription of ncRNAs and PCGs. Moreover, we built a web-based tool, Co-Expression, to explore the co-expression patterns between DNA-binding proteins and various types of genes by integrating the gene expression profiles of ∼10 000 tumor samples and ∼9100 normal tissues and cell lines. ChIPBase also provides a ChIP-Function tool and a genome browser to predict functions of diverse genes and visualize various ChIP-seq data. This study will greatly expand our understanding of the transcriptional regulations of ncRNAs and PCGs. © The Author(s) 2016. Published by Oxford University Press on behalf of Nucleic Acids Research.

Network perturbation by recurrent regulatory variants in cancer.

Directory of Open Access Journals (Sweden)

Kiwon Jang

2017-03-01

Full Text Available Cancer driving genes have been identified as recurrently affected by variants that alter protein-coding sequences. However, a majority of cancer variants arise in noncoding regions, and some of them are thought to play a critical role through transcriptional perturbation. Here we identified putative transcriptional driver genes based on combinatorial variant recurrence in cis-regulatory regions. The identified genes showed high connectivity in the cancer type-specific transcription regulatory network, with high outdegree and many downstream genes, highlighting their causative role during tumorigenesis. In the protein interactome, the identified transcriptional drivers were not as highly connected as coding driver genes but appeared to form a network module centered on the coding drivers. The coding and regulatory variants associated via these interactions between the coding and transcriptional drivers showed exclusive and complementary occurrence patterns across tumor samples. Transcriptional cancer drivers may act through an extensive perturbation of the regulatory network and by altering protein network modules through interactions with coding driver genes.

Steam condensation modelling in aerosol codes

International Nuclear Information System (INIS)

Dunbar, I.H.

1986-01-01

The principal subject of this study is the modelling of the condensation of steam into and evaporation of water from aerosol particles. These processes introduce a new type of term into the equation for the development of the aerosol particle size distribution. This new term faces the code developer with three major problems: the physical modelling of the condensation/evaporation process, the discretisation of the new term and the separate accounting for the masses of the water and of the other components. This study has considered four codes which model the condensation of steam into and its evaporation from aerosol particles: AEROSYM-M (UK), AEROSOLS/B1 (France), NAUA (Federal Republic of Germany) and CONTAIN (USA). The modelling in the codes has been addressed under three headings. These are the physical modelling of condensation, the mathematics of the discretisation of the equations, and the methods for modelling the separate behaviour of different chemical components of the aerosol. The codes are least advanced in area of solute effect modelling. At present only AEROSOLS/B1 includes the effect. The effect is greater for more concentrated solutions. Codes without the effect will be more in error (underestimating the total airborne mass) the less condensation they predict. Data are needed on the water vapour pressure above concentrated solutions of the substances of interest (especially CsOH and CsI) if the extent to which aerosols retain water under superheated conditions is to be modelled. 15 refs

Regulatory Roles for Long ncRNA and mRNA

International Nuclear Information System (INIS)

Karapetyan, Armen R.; Buiting, Coen; Kuiper, Renske A.; Coolen, Marcel W.

2013-01-01

Recent advances in high-throughput sequencing technology have identified the transcription of a much larger portion of the genome than previously anticipated. Especially in the context of cancer it has become clear that aberrant transcription of both protein-coding and long non-coding RNAs (lncRNAs) are frequent events. The current dogma of RNA function describes mRNA to be responsible for the synthesis of proteins, whereas non-coding RNA can have regulatory or epigenetic functions. However, this distinction between protein coding and regulatory ability of transcripts may not be that strict. Here, we review the increasing body of evidence for the existence of multifunctional RNAs that have both protein-coding and trans-regulatory roles. Moreover, we demonstrate that coding transcripts bind to components of the Polycomb Repressor Complex 2 (PRC2) with similar affinities as non-coding transcripts, revealing potential epigenetic regulation by mRNAs. We hypothesize that studies on the regulatory ability of disease-associated mRNAs will form an important new field of research

Regulatory Roles for Long ncRNA and mRNA

Energy Technology Data Exchange (ETDEWEB)

Karapetyan, Armen R.; Buiting, Coen; Kuiper, Renske A.; Coolen, Marcel W., E-mail: M.Coolen@gen.umcn.nl [Department of Human Genetics, Nijmegen Centre for Molecular Life Sciences (NCMLS), Radboud University Nijmegen Medical Centre, P.O. Box 9101, Nijmegen 6500 HB (Netherlands)

2013-04-26

Recent advances in high-throughput sequencing technology have identified the transcription of a much larger portion of the genome than previously anticipated. Especially in the context of cancer it has become clear that aberrant transcription of both protein-coding and long non-coding RNAs (lncRNAs) are frequent events. The current dogma of RNA function describes mRNA to be responsible for the synthesis of proteins, whereas non-coding RNA can have regulatory or epigenetic functions. However, this distinction between protein coding and regulatory ability of transcripts may not be that strict. Here, we review the increasing body of evidence for the existence of multifunctional RNAs that have both protein-coding and trans-regulatory roles. Moreover, we demonstrate that coding transcripts bind to components of the Polycomb Repressor Complex 2 (PRC2) with similar affinities as non-coding transcripts, revealing potential epigenetic regulation by mRNAs. We hypothesize that studies on the regulatory ability of disease-associated mRNAs will form an important new field of research.

SCDAP/RELAP5/MOD3 code development and assessment

International Nuclear Information System (INIS)

Allison, C.M.; Heath, C.H.; Siefken, L.J.; Hohorst, J.K.

1991-01-01

The SCDAP/RELAP5/MOD3 computer code is designed to describe the overall reactor coolant system (RCS) thermal-hydraulic response, core damage progression, and fission product release and transport during severe accidents. The code is being developed at the Idaho National Engineering Laboratory (INEL) under the primary sponsorship of the Office of Nuclear Regulatory Research of the Nuclear Regulatory Commission (NRC). SCDAP/RELAP5/MOD3, created in January, 1991, is the result of merging RELAP5/MOD3 with SCDAP and TRAP-MELT models from SCDAP/RELAP5/MOD2.5. The RELAP5 models calculate the overall RCS thermal-hydraulics, control system interactions, reactor kinetics, and the transport of noncondensible gases, fission products, and aerosols. The SCDAP models calculate the damage progression in the core structures, the formation, heatup, and melting of debris, and the creep rupture failure of the lower head and other RCS structures. The TRAP-MELT models calculate the deposition of fission products upon aerosols or structural surfaces; the formation, growth, or deposition of aerosols; and the evaporation of species from surfaces. The systematic assessment of modeling uncertainties in SCDAP/RELAP5 code is currently underway. This assessment includes (a) the evaluation of code-to-data comparisons using stand-alone SCDAP and SCDAP/RELAP5/MOD3, (b) the estimation of modeling and experimental uncertainties, and (c) the determination of the influence of those uncertainties on predicted severe accident behavior

Modeling Dynamic Regulatory Processes in Stroke

Science.gov (United States)

McDermott, Jason E.; Jarman, Kenneth; Taylor, Ronald; Lancaster, Mary; Shankaran, Harish; Vartanian, Keri B.; Stevens, Susan L.; Stenzel-Poore, Mary P.; Sanfilippo, Antonio

2012-01-01

The ability to examine the behavior of biological systems in silico has the potential to greatly accelerate the pace of discovery in diseases, such as stroke, where in vivo analysis is time intensive and costly. In this paper we describe an approach for in silico examination of responses of the blood transcriptome to neuroprotective agents and subsequent stroke through the development of dynamic models of the regulatory processes observed in the experimental gene expression data. First, we identified functional gene clusters from these data. Next, we derived ordinary differential equations (ODEs) from the data relating these functional clusters to each other in terms of their regulatory influence on one another. Dynamic models were developed by coupling these ODEs into a model that simulates the expression of regulated functional clusters. By changing the magnitude of gene expression in the initial input state it was possible to assess the behavior of the networks through time under varying conditions since the dynamic model only requires an initial starting state, and does not require measurement of regulatory influences at each time point in order to make accurate predictions. We discuss the implications of our models on neuroprotection in stroke, explore the limitations of the approach, and report that an optimized dynamic model can provide accurate predictions of overall system behavior under several different neuroprotective paradigms. PMID:23071432

Use of Lump Parameter Codes at SNSA

International Nuclear Information System (INIS)

Muehleisen, A.

2006-01-01

The lump parameter codes are due to the specifics of Slovenian regulation used only in a very limited scope by the SNSA itself. The law requires that most of the analysis needed for regulatory decision making have to be performed by technical support organisations (TSOs). The use of lump parameter codes is therefore limited to the amount needed to maintain necessary technical competence and to support, to a degree, the reasoning for raising new issues and methodologies. SNSA has available its own NPP MELCOR model and uses for its own purposes NPP Krsko RELAP model. RELAP model is also part of the SNSA NPA analyser. Here presented recent uses at SNSA include use of NPA in support of a project, aimed at estimating maturity and uses of CFD codes for regulatory purposes, transition from MELCOR 1.8.3 to 1.8.5 model and its validation, developing MELCOR PAR model and use of NPA for training purposes. NPA use in support of investigation of CFD usability has been in performing lump parameter code calculation against which the CFD results could be compared. The case of SI injection and the following boron distribution in the reactor vessel has been used for this purpose. The comparison showed that for the particular case there is no urgent need for CFD code calculations, nevertheless the project clearly demonstrated wealth of additional information that can be gained by the use of CFD code. As far as MELCOR model is concerned, only transition of the model to the newer code version has been performed and PAR input prepared and tested. Even though there is a feeling at SNSA that some preliminary analysis with it (such as analysis of typical accidents with PARs present and analysis in support of wet cavity modification) would be useful as a support for decision making as well as for simple training purposes we have not been able to perform them due to other priorities and lack of human resources. SNSA is additionally tasked with support to TSOs in their efforts to maintain and

PetriCode: A Tool for Template-Based Code Generation from CPN Models

DEFF Research Database (Denmark)

Simonsen, Kent Inge

2014-01-01

Code generation is an important part of model driven methodologies. In this paper, we present PetriCode, a software tool for generating protocol software from a subclass of Coloured Petri Nets (CPNs). The CPN subclass is comprised of hierarchical CPN models describing a protocol system at different...

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

Towards Product Lining Model-Driven Development Code Generators

OpenAIRE

Roth, Alexander; Rumpe, Bernhard

2015-01-01

A code generator systematically transforms compact models to detailed code. Today, code generation is regarded as an integral part of model-driven development (MDD). Despite its relevance, the development of code generators is an inherently complex task and common methodologies and architectures are lacking. Additionally, reuse and extension of existing code generators only exist on individual parts. A systematic development and reuse based on a code generator product line is still in its inf...

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

Heat Transfer treatment in computer codes for safety analysis

International Nuclear Information System (INIS)

Jerele, A.; Gregoric, M.

1984-01-01

Increased number of operating nuclear power plants has stressed importance of nuclear safety evaluation. For this reason, accordingly to regulatory commission request, safety analyses with computer codes are preformed. In this paper part of this thermohydraulic models dealing with wall-to-fluid heat transfer correlations in computer codes TRAC=PF1, RELAP4/MOD5, RELAP5/MOD1 and COBRA-IV is discussed. (author)

Review of the SCDAP/RELAP5/MOD3.1 code structure and core T/H model before core damage

International Nuclear Information System (INIS)

Kim, See Darl; Kim, Dong Ha

1998-04-01

The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code is being developed at the INEL under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. NRC. As The current time, the SCDAP/RELAP5/MOD3.1 code is the result of merging the RELAP5/MOD3 and SCDAP models. The code models the coupled behavior of the reactor coolant system, core, fission product released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. Major purpose of the report is to provide information about the characteristics of SCDAP/RELAP5/MOD3.1 core T/H models for an integrated severe accident computer code being developed under the mid/long-term project. This report analyzes the overall code structure which consists of the input processor, transient controller, and plot file handler. The basic governing equations to simulate the thermohydraulics of the primary system are also described. As the focus is currently concentrated in the core, core nodalization parameters of the intact geometry and the phenomenological subroutines for the damaged core are summarized for the future usage. In addition, the numerical approach for the heat conduction model is investigated along with heat convection model. These studies could provide a foundation for input preparation and model improvement. (author). 6 refs., 3 tabs., 4 figs

Regulatory Roles for Long ncRNA and mRNA

Directory of Open Access Journals (Sweden)

Marcel W. Coolen

2013-04-01

Full Text Available Recent advances in high-throughput sequencing technology have identified the transcription of a much larger portion of the genome than previously anticipated. Especially in the context of cancer it has become clear that aberrant transcription of both protein-coding and long non-coding RNAs (lncRNAs are frequent events. The current dogma of RNA function describes mRNA to be responsible for the synthesis of proteins, whereas non-coding RNA can have regulatory or epigenetic functions. However, this distinction between protein coding and regulatory ability of transcripts may not be that strict. Here, we review the increasing body of evidence for the existence of multifunctional RNAs that have both protein-coding and trans-regulatory roles. Moreover, we demonstrate that coding transcripts bind to components of the Polycomb Repressor Complex 2 (PRC2 with similar affinities as non-coding transcripts, revealing potential epigenetic regulation by mRNAs. We hypothesize that studies on the regulatory ability of disease-associated mRNAs will form an important new field of research.

Modeling post-transcriptional regulation activity of small non-coding RNAs in Escherichia coli.

Science.gov (United States)

Wang, Rui-Sheng; Jin, Guangxu; Zhang, Xiang-Sun; Chen, Luonan

2009-04-29

Transcriptional regulation is a fundamental process in biological systems, where transcription factors (TFs) have been revealed to play crucial roles. In recent years, in addition to TFs, an increasing number of non-coding RNAs (ncRNAs) have been shown to mediate post-transcriptional processes and regulate many critical pathways in both prokaryotes and eukaryotes. On the other hand, with more and more high-throughput biological data becoming available, it is possible and imperative to quantitatively study gene regulation in a systematic and detailed manner. Most existing studies for inferring transcriptional regulatory interactions and the activity of TFs ignore the possible post-transcriptional effects of ncRNAs. In this work, we propose a novel framework to infer the activity of regulators including both TFs and ncRNAs by exploring the expression profiles of target genes and (post)transcriptional regulatory relationships. We model the integrated regulatory system by a set of biochemical reactions which lead to a log-bilinear problem. The inference process is achieved by an iterative algorithm, in which two linear programming models are efficiently solved. In contrast to available related studies, the effects of ncRNAs on transcription process are considered in this work, and thus more reasonable and accurate reconstruction can be expected. In addition, the approach is suitable for large-scale problems from the viewpoint of computation. Experiments on two synthesized data sets and a model system of Escherichia coli (E. coli) carbon source transition from glucose to acetate illustrate the effectiveness of our model and algorithm. Our results show that incorporating the post-transcriptional regulation of ncRNAs into system model can mine the hidden effects from the regulation activity of TFs in transcription processes and thus can uncover the biological mechanisms in gene regulation in a more accurate manner. The software for the algorithm in this paper is available

Regulatory framework and business models for charging plug-in electric vehicles: Infrastructure, agents, and commercial relationships

Energy Technology Data Exchange (ETDEWEB)

Gomez San Roman, Tomas [Instituto de Investigacion Tecnologica, Universidad Pontificia Comillas, Madrid (Spain); Momber, Ilan, E-mail: ilan.momber@iit.upcomillas.es [Instituto de Investigacion Tecnologica, Universidad Pontificia Comillas, Madrid (Spain); Rivier Abbad, Michel; Sanchez Miralles, Alvaro [Instituto de Investigacion Tecnologica, Universidad Pontificia Comillas, Madrid (Spain)

2011-10-15

Electric vehicles (EVs) present efficiency and environmental advantages over conventional transportation. It is expected that in the next decade this technology will progressively penetrate the market. The integration of plug-in electric vehicles in electric power systems poses new challenges in terms of regulation and business models. This paper proposes a conceptual regulatory framework for charging EVs. Two new electricity market agents, the EV charging manager and the EV aggregator, in charge of developing charging infrastructure and providing charging services are introduced. According to that, several charging modes such as EV home charging, public charging on streets, and dedicated charging stations are formulated. Involved market agents and their commercial relationships are analysed in detail. The paper elaborates the opportunities to formulate more sophisticated business models for vehicle-to-grid applications under which the storage capability of EV batteries is used for providing peak power or frequency regulation to support the power system operation. Finally penetration phase dependent policy and regulatory recommendations are given concerning time-of-use pricing, smart meter deployment, stable and simple regulation for reselling energy on private property, roll-out of public charging infrastructure as well as reviewing of grid codes and operational system procedures for interactions between network operators and vehicle aggregators. - Highlights: > A conceptual regulatory framework for charging EVs is proposed. > 2 new agents, EV charging point manager, EV aggregator and their functions are introduced. > Depending on private or public access of charging points, contractual relations change. > A classification of charging scenarios alludes implications on regulatory topics. > EV penetration phase dependent policy and regulatory recommendations are given.

The nuclear reaction model code MEDICUS

International Nuclear Information System (INIS)

Ibishia, A.I.

2008-01-01

The new computer code MEDICUS has been used to calculate cross sections of nuclear reactions. The code, implemented in MATLAB 6.5, Mathematica 5, and Fortran 95 programming languages, can be run in graphical and command line mode. Graphical User Interface (GUI) has been built that allows the user to perform calculations and to plot results just by mouse clicking. The MS Windows XP and Red Hat Linux platforms are supported. MEDICUS is a modern nuclear reaction code that can compute charged particle-, photon-, and neutron-induced reactions in the energy range from thresholds to about 200 MeV. The calculation of the cross sections of nuclear reactions are done in the framework of the Exact Many-Body Nuclear Cluster Model (EMBNCM), Direct Nuclear Reactions, Pre-equilibrium Reactions, Optical Model, DWBA, and Exciton Model with Cluster Emission. The code can be used also for the calculation of nuclear cluster structure of nuclei. We have calculated nuclear cluster models for some nuclei such as 177 Lu, 90 Y, and 27 Al. It has been found that nucleus 27 Al can be represented through the two different nuclear cluster models: 25 Mg + d and 24 Na + 3 He. Cross sections in function of energy for the reaction 27 Al( 3 He,x) 22 Na, established as a production method of 22 Na, are calculated by the code MEDICUS. Theoretical calculations of cross sections are in good agreement with experimental results. Reaction mechanisms are taken into account. (author)

CMCpy: Genetic Code-Message Coevolution Models in Python

Science.gov (United States)

Becich, Peter J.; Stark, Brian P.; Bhat, Harish S.; Ardell, David H.

2013-01-01

Code-message coevolution (CMC) models represent coevolution of a genetic code and a population of protein-coding genes (“messages”). Formally, CMC models are sets of quasispecies coupled together for fitness through a shared genetic code. Although CMC models display plausible explanations for the origin of multiple genetic code traits by natural selection, useful modern implementations of CMC models are not currently available. To meet this need we present CMCpy, an object-oriented Python API and command-line executable front-end that can reproduce all published results of CMC models. CMCpy implements multiple solvers for leading eigenpairs of quasispecies models. We also present novel analytical results that extend and generalize applications of perturbation theory to quasispecies models and pioneer the application of a homotopy method for quasispecies with non-unique maximally fit genotypes. Our results therefore facilitate the computational and analytical study of a variety of evolutionary systems. CMCpy is free open-source software available from http://pypi.python.org/pypi/CMCpy/. PMID:23532367

24 CFR 200.925c - Model codes.

Science.gov (United States)

2010-04-01

... below. (1) Model Building Codes—(i) The BOCA National Building Code, 1993 Edition, The BOCA National..., Administration, for the Building, Plumbing and Mechanical Codes and the references to fire retardant treated wood... number 2 (Chapter 7) of the Building Code, but including the Appendices of the Code. Available from...

Pleiotropy constrains the evolution of protein but not regulatory sequences in a transcription regulatory network influencing complex social behaviours

Directory of Open Access Journals (Sweden)

Daria eMolodtsova

2014-12-01

Full Text Available It is increasingly apparent that genes and networks that influence complex behaviour are evolutionary conserved, which is paradoxical considering that behaviour is labile over evolutionary timescales. How does adaptive change in behaviour arise if behaviour is controlled by conserved, pleiotropic, and likely evolutionary constrained genes? Pleiotropy and connectedness are known to constrain the general rate of protein evolution, prompting some to suggest that the evolution of complex traits, including behaviour, is fuelled by regulatory sequence evolution. However, we seldom have data on the strength of selection on mutations in coding and regulatory sequences, and this hinders our ability to study how pleiotropy influences coding and regulatory sequence evolution. Here we use population genomics to estimate the strength of selection on coding and regulatory mutations for a transcriptional regulatory network that influences complex behaviour of honey bees. We found that replacement mutations in highly connected transcription factors and target genes experience significantly stronger negative selection relative to weakly connected transcription factors and targets. Adaptively evolving proteins were significantly more likely to reside at the periphery of the regulatory network, while proteins with signs of negative selection were near the core of the network. Interestingly, connectedness and network structure had minimal influence on the strength of selection on putative regulatory sequences for both transcription factors and their targets. Our study indicates that adaptive evolution of complex behaviour can arise because of positive selection on protein-coding mutations in peripheral genes, and on regulatory sequence mutations in both transcription factors and their targets throughout the network.

Regulatory agencies and regulatory risk

OpenAIRE

Knieps, Günter; Weiß, Hans-Jörg

2008-01-01

The aim of this paper is to show that regulatory risk is due to the discretionary behaviour of regulatory agencies, caused by a too extensive regulatory mandate provided by the legislator. The normative point of reference and a behavioural model of regulatory agencies based on the positive theory of regulation are presented. Regulatory risk with regard to the future behaviour of regulatory agencies is modelled as the consequence of the ex ante uncertainty about the relative influence of inter...

Interfacial and Wall Transport Models for SPACE-CAP Code

International Nuclear Information System (INIS)

Hong, Soon Joon; Choo, Yeon Joon; Han, Tae Young; Hwang, Su Hyun; Lee, Byung Chul; Choi, Hoon; Ha, Sang Jun

2009-01-01

The development project for the domestic design code was launched to be used for the safety and performance analysis of pressurized light water reactors. And CAP (Containment Analysis Package) code has been also developed for the containment safety and performance analysis side by side with SPACE. The CAP code treats three fields (gas, continuous liquid, and dispersed drop) for the assessment of containment specific phenomena, and is featured by its multidimensional assessment capabilities. Thermal hydraulics solver was already developed and now under testing of its stability and soundness. As a next step, interfacial and wall transport models was setup. In order to develop the best model and correlation package for the CAP code, various models currently used in major containment analysis codes, which are GOTHIC, CONTAIN2.0, and CONTEMPT-LT, have been reviewed. The origins of the selected models used in these codes have also been examined to find out if the models have not conflict with a proprietary right. In addition, a literature survey of the recent studies has been performed in order to incorporate the better models for the CAP code. The models and correlations of SPACE were also reviewed. CAP models and correlations are composed of interfacial heat/mass, and momentum transport models, and wall heat/mass, and momentum transport models. This paper discusses on those transport models in the CAP code

Interfacial and Wall Transport Models for SPACE-CAP Code

Energy Technology Data Exchange (ETDEWEB)

Hong, Soon Joon; Choo, Yeon Joon; Han, Tae Young; Hwang, Su Hyun; Lee, Byung Chul [FNC Tech., Seoul (Korea, Republic of); Choi, Hoon; Ha, Sang Jun [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

2009-10-15

The development project for the domestic design code was launched to be used for the safety and performance analysis of pressurized light water reactors. And CAP (Containment Analysis Package) code has been also developed for the containment safety and performance analysis side by side with SPACE. The CAP code treats three fields (gas, continuous liquid, and dispersed drop) for the assessment of containment specific phenomena, and is featured by its multidimensional assessment capabilities. Thermal hydraulics solver was already developed and now under testing of its stability and soundness. As a next step, interfacial and wall transport models was setup. In order to develop the best model and correlation package for the CAP code, various models currently used in major containment analysis codes, which are GOTHIC, CONTAIN2.0, and CONTEMPT-LT, have been reviewed. The origins of the selected models used in these codes have also been examined to find out if the models have not conflict with a proprietary right. In addition, a literature survey of the recent studies has been performed in order to incorporate the better models for the CAP code. The models and correlations of SPACE were also reviewed. CAP models and correlations are composed of interfacial heat/mass, and momentum transport models, and wall heat/mass, and momentum transport models. This paper discusses on those transport models in the CAP code.

Provisional safety analyses for SGT stage 2 -- Models, codes and general modelling approach

International Nuclear Information System (INIS)

2014-12-01

In the framework of the provisional safety analyses for Stage 2 of the Sectoral Plan for Deep Geological Repositories (SGT), deterministic modelling of radionuclide release from the barrier system along the groundwater pathway during the post-closure period of a deep geological repository is carried out. The calculated radionuclide release rates are interpreted as annual effective dose for an individual and assessed against the regulatory protection criterion 1 of 0.1 mSv per year. These steps are referred to as dose calculations. Furthermore, from the results of the dose calculations so-called characteristic dose intervals are determined, which provide input to the safety-related comparison of the geological siting regions in SGT Stage 2. Finally, the results of the dose calculations are also used to illustrate and to evaluate the post-closure performance of the barrier systems under consideration. The principal objective of this report is to describe comprehensively the technical aspects of the dose calculations. These aspects comprise: · the generic conceptual models of radionuclide release from the solid waste forms, of radionuclide transport through the system of engineered and geological barriers, of radionuclide transfer in the biosphere, as well as of the potential radiation exposure of the population, · the mathematical models for the explicitly considered release and transport processes, as well as for the radiation exposure pathways that are included, · the implementation of the mathematical models in numerical codes, including an overview of these codes and the most relevant verification steps, · the general modelling approach when using the codes, in particular the generic assumptions needed to model the near field and the geosphere, along with some numerical details, · a description of the work flow related to the execution of the calculations and of the software tools that are used to facilitate the modelling process, and · an overview of the

Provisional safety analyses for SGT stage 2 -- Models, codes and general modelling approach

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-12-15

In the framework of the provisional safety analyses for Stage 2 of the Sectoral Plan for Deep Geological Repositories (SGT), deterministic modelling of radionuclide release from the barrier system along the groundwater pathway during the post-closure period of a deep geological repository is carried out. The calculated radionuclide release rates are interpreted as annual effective dose for an individual and assessed against the regulatory protection criterion 1 of 0.1 mSv per year. These steps are referred to as dose calculations. Furthermore, from the results of the dose calculations so-called characteristic dose intervals are determined, which provide input to the safety-related comparison of the geological siting regions in SGT Stage 2. Finally, the results of the dose calculations are also used to illustrate and to evaluate the post-closure performance of the barrier systems under consideration. The principal objective of this report is to describe comprehensively the technical aspects of the dose calculations. These aspects comprise: · the generic conceptual models of radionuclide release from the solid waste forms, of radionuclide transport through the system of engineered and geological barriers, of radionuclide transfer in the biosphere, as well as of the potential radiation exposure of the population, · the mathematical models for the explicitly considered release and transport processes, as well as for the radiation exposure pathways that are included, · the implementation of the mathematical models in numerical codes, including an overview of these codes and the most relevant verification steps, · the general modelling approach when using the codes, in particular the generic assumptions needed to model the near field and the geosphere, along with some numerical details, · a description of the work flow related to the execution of the calculations and of the software tools that are used to facilitate the modelling process, and · an overview of the

High burnup models in computer code fair

Energy Technology Data Exchange (ETDEWEB)

Dutta, B K; Swami Prasad, P; Kushwaha, H S; Mahajan, S C; Kakodar, A [Bhabha Atomic Research Centre, Bombay (India)

1997-08-01

An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ``Light water reactor fuel rod modelling code evaluation`` and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs.

High burnup models in computer code fair

International Nuclear Information System (INIS)

Dutta, B.K.; Swami Prasad, P.; Kushwaha, H.S.; Mahajan, S.C.; Kakodar, A.

1997-01-01

An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ''Light water reactor fuel rod modelling code evaluation'' and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs

One dimensional neutron kinetics in the TRAC-BF1 code

International Nuclear Information System (INIS)

Weaver, W.L. III; Wagner, K.C.

1987-01-01

The TRAC-BWR code development program at the Idaho National Engineering Laboratory is developing a version of the TRAC code for the U.S. Nuclear Regulatory Commission (USNRC) to provide a best-estimate analysis capability for the simulation of postulated accidents in boiling water reactor (BWR) power systems and related experimental facilities. Recent development efforts in the TRAC-BWR program have focused on improving the computational efficiency through the incorporation of a hybrid Courant- limit-violating numerical solution scheme in the one-dimensional component models and on improving code accuracy through the development of a one-dimensional neutron kinetics model. Many other improvements have been incorporated into TRAC-BWR to improve code portability, accuracy, efficiency, and maintainability. This paper will describe the one- dimensional neutron kinetics model, the generation of the required input data for this model, and present results of the first calculations using the model

ITER Dynamic Tritium Inventory Modeling Code

International Nuclear Information System (INIS)

Cristescu, Ioana-R.; Doerr, L.; Busigin, A.; Murdoch, D.

2005-01-01

A tool for tritium inventory evaluation within each sub-system of the Fuel Cycle of ITER is vital, with respect to both the process of licensing ITER and also for operation. It is very likely that measurements of total tritium inventories may not be possible for all sub-systems, however tritium accounting may be achieved by modeling its hold-up within each sub-system and by validating these models in real-time against the monitored flows and tritium streams between the systems. To get reliable results, an accurate dynamic modeling of the tritium content in each sub-system is necessary. In order to optimize the configuration and operation of the ITER fuel cycle, a dynamic fuel cycle model was developed progressively in the decade up to 2000-2001. As the design for some sub-systems from the fuel cycle (i.e. Vacuum pumping, Neutral Beam Injectors (NBI)) have substantially progressed meanwhile, a new code developed under a different platform to incorporate these modifications has been developed. The new code is taking over the models and algorithms for some subsystems, such as Isotope Separation System (ISS); where simplified models have been previously considered, more detailed have been introduced, as for the Water Detritiation System (WDS). To reflect all these changes, the new code developed inside EU participating team was nominated TRIMO (Tritium Inventory Modeling), to emphasize the use of the code on assessing the tritium inventory within ITER

Modeling stochasticity and robustness in gene regulatory networks.

Science.gov (United States)

Garg, Abhishek; Mohanram, Kartik; Di Cara, Alessandro; De Micheli, Giovanni; Xenarios, Ioannis

2009-06-15

Understanding gene regulation in biological processes and modeling the robustness of underlying regulatory networks is an important problem that is currently being addressed by computational systems biologists. Lately, there has been a renewed interest in Boolean modeling techniques for gene regulatory networks (GRNs). However, due to their deterministic nature, it is often difficult to identify whether these modeling approaches are robust to the addition of stochastic noise that is widespread in gene regulatory processes. Stochasticity in Boolean models of GRNs has been addressed relatively sparingly in the past, mainly by flipping the expression of genes between different expression levels with a predefined probability. This stochasticity in nodes (SIN) model leads to over representation of noise in GRNs and hence non-correspondence with biological observations. In this article, we introduce the stochasticity in functions (SIF) model for simulating stochasticity in Boolean models of GRNs. By providing biological motivation behind the use of the SIF model and applying it to the T-helper and T-cell activation networks, we show that the SIF model provides more biologically robust results than the existing SIN model of stochasticity in GRNs. Algorithms are made available under our Boolean modeling toolbox, GenYsis. The software binaries can be downloaded from http://si2.epfl.ch/ approximately garg/genysis.html.

Assessment of the Effects on PCT Evaluation of Enhanced Fuel Model Facilitated by Coupling the MARS Code with the FRAPTRAN Code

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyong Chol; Lee, Young Jin; Han, Sam Hee [NSE Technology Inc., Daejeon (Korea, Republic of)

2016-10-15

The principal objectives of the two safety criteria, peak cladding temperature (PCT) and total oxidation limits, are to ensure that the fuel rod claddings remain sufficiently ductile so that they do not crack and fragment during a LOCA. Another important purpose of the PCT limit is to ensure that the fuel cladding does not enter the regime of runaway oxidation and uncontrollable heat-up. However, even when the PCT limit is satisfied, it is known that cladding failures may still occur in a certain percentage of the fuel rods during a LOCA. This is largely because a 100% fuel failure is assumed for the radiological consequence analysis in the US regulatory practices. In this study, we analyze the effects of cladding failure and other fuel model features on PCT during a LOCA using the MARS-FRAPTRAN coupled code. MARS code has been coupled with FRAPTRAN code to extend fuel modeling capability. The coupling allows feedback of FRAPTRAN results in real time. Because of the significant impact of fuel models on key safety parameters such as PCT, detailed and accurate fuel models should be employed when evaluating PCT in LOCA analysis. It is noteworthy that the ECCS evaluation models laid out in the Appendix K to 10CFR50 require a provision for predicting cladding swelling and rupture and require to assume that the inside of the cladding react with steam after the rupture. The metal-water reaction energy can have significantly large effect on the reflood PCT, especially when fuel failure occurs. Effects of applying an advanced fuel model on the PCT evaluation can be clearly seen when comparing the MARS and the FRAPTRAN results in both the one-way calculation and the feedback calculation. As long as MARS and FRAPTRAN are used respectively in the ranges where they have been validated, the coupled calculation results are expected to be valid and to reveal various aspects of phenomena which have not been discovered in previous uncoupled calculations by MARS or FRAPTRAN.

Advanced thermal-hydraulic and neutronic codes: current and future applications. Summary and conclusions

International Nuclear Information System (INIS)

2001-05-01

An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for

Regulatory framework and business models for charging plug-in electric vehicles: Infrastructure, agents, and commercial relationships

International Nuclear Information System (INIS)

Gomez San Roman, Tomas; Momber, Ilan; Rivier Abbad, Michel; Sanchez Miralles, Alvaro

2011-01-01

Electric vehicles (EVs) present efficiency and environmental advantages over conventional transportation. It is expected that in the next decade this technology will progressively penetrate the market. The integration of plug-in electric vehicles in electric power systems poses new challenges in terms of regulation and business models. This paper proposes a conceptual regulatory framework for charging EVs. Two new electricity market agents, the EV charging manager and the EV aggregator, in charge of developing charging infrastructure and providing charging services are introduced. According to that, several charging modes such as EV home charging, public charging on streets, and dedicated charging stations are formulated. Involved market agents and their commercial relationships are analysed in detail. The paper elaborates the opportunities to formulate more sophisticated business models for vehicle-to-grid applications under which the storage capability of EV batteries is used for providing peak power or frequency regulation to support the power system operation. Finally penetration phase dependent policy and regulatory recommendations are given concerning time-of-use pricing, smart meter deployment, stable and simple regulation for reselling energy on private property, roll-out of public charging infrastructure as well as reviewing of grid codes and operational system procedures for interactions between network operators and vehicle aggregators. - Highlights: → A conceptual regulatory framework for charging EVs is proposed. → 2 new agents, EV charging point manager, EV aggregator and their functions are introduced. → Depending on private or public access of charging points, contractual relations change. → A classification of charging scenarios alludes implications on regulatory topics. → EV penetration phase dependent policy and regulatory recommendations are given.

Modeling report of DYMOND code (DUPIC version)

International Nuclear Information System (INIS)

Park, Joo Hwan; Yacout, Abdellatif M.

2003-04-01

The DYMOND code employs the ITHINK dynamic modeling platform to assess the 100-year dynamic evolution scenarios for postulated global nuclear energy parks. Firstly, DYMOND code has been developed by ANL(Argonne National Laboratory) to perform the fuel cycle analysis of LWR once-through and LWR-FBR mixed plant. Since the extensive application of DYMOND code has been requested, the first version of DYMOND has been modified to adapt the DUPIC, MSR and RTF fuel cycle. DYMOND code is composed of three parts; the source language platform, input supply and output. But those platforms are not clearly distinguished. This report described all the equations which were modeled in the modified DYMOND code (which is called as DYMOND-DUPIC version). It divided into five parts;Part A deals model in reactor history which is included amount of the requested fuels and spent fuels. Part B aims to describe model of fuel cycle about fuel flow from the beginning to the end of fuel cycle. Part C is for model in re-processing which is included recovery of burned uranium, plutonium, minor actinide and fission product as well as the amount of spent fuels in storage and disposal. Part D is for model in other fuel cycle which is considered the thorium fuel cycle for MSR and RTF reactor. Part E is for model in economics. This part gives all the information of cost such as uranium mining cost, reactor operating cost, fuel cost etc

Modeling report of DYMOND code (DUPIC version)

Energy Technology Data Exchange (ETDEWEB)

Park, Joo Hwan [KAERI, Taejon (Korea, Republic of); Yacout, Abdellatif M [Argonne National Laboratory, Ilinois (United States)

2003-04-01

The DYMOND code employs the ITHINK dynamic modeling platform to assess the 100-year dynamic evolution scenarios for postulated global nuclear energy parks. Firstly, DYMOND code has been developed by ANL(Argonne National Laboratory) to perform the fuel cycle analysis of LWR once-through and LWR-FBR mixed plant. Since the extensive application of DYMOND code has been requested, the first version of DYMOND has been modified to adapt the DUPIC, MSR and RTF fuel cycle. DYMOND code is composed of three parts; the source language platform, input supply and output. But those platforms are not clearly distinguished. This report described all the equations which were modeled in the modified DYMOND code (which is called as DYMOND-DUPIC version). It divided into five parts;Part A deals model in reactor history which is included amount of the requested fuels and spent fuels. Part B aims to describe model of fuel cycle about fuel flow from the beginning to the end of fuel cycle. Part C is for model in re-processing which is included recovery of burned uranium, plutonium, minor actinide and fission product as well as the amount of spent fuels in storage and disposal. Part D is for model in other fuel cycle which is considered the thorium fuel cycle for MSR and RTF reactor. Part E is for model in economics. This part gives all the information of cost such as uranium mining cost, reactor operating cost, fuel cost etc.

Content Coding of Psychotherapy Transcripts Using Labeled Topic Models.

Science.gov (United States)

Gaut, Garren; Steyvers, Mark; Imel, Zac E; Atkins, David C; Smyth, Padhraic

2017-03-01

Psychotherapy represents a broad class of medical interventions received by millions of patients each year. Unlike most medical treatments, its primary mechanisms are linguistic; i.e., the treatment relies directly on a conversation between a patient and provider. However, the evaluation of patient-provider conversation suffers from critical shortcomings, including intensive labor requirements, coder error, nonstandardized coding systems, and inability to scale up to larger data sets. To overcome these shortcomings, psychotherapy analysis needs a reliable and scalable method for summarizing the content of treatment encounters. We used a publicly available psychotherapy corpus from Alexander Street press comprising a large collection of transcripts of patient-provider conversations to compare coding performance for two machine learning methods. We used the labeled latent Dirichlet allocation (L-LDA) model to learn associations between text and codes, to predict codes in psychotherapy sessions, and to localize specific passages of within-session text representative of a session code. We compared the L-LDA model to a baseline lasso regression model using predictive accuracy and model generalizability (measured by calculating the area under the curve (AUC) from the receiver operating characteristic curve). The L-LDA model outperforms the lasso logistic regression model at predicting session-level codes with average AUC scores of 0.79, and 0.70, respectively. For fine-grained level coding, L-LDA and logistic regression are able to identify specific talk-turns representative of symptom codes. However, model performance for talk-turn identification is not yet as reliable as human coders. We conclude that the L-LDA model has the potential to be an objective, scalable method for accurate automated coding of psychotherapy sessions that perform better than comparable discriminative methods at session-level coding and can also predict fine-grained codes.

MDEP Technical Report TR-CSWG-01. Technical Report: Regulatory Frameworks for the Use of Nuclear Pressure Boundary Codes and Standards in MDEP Countries

International Nuclear Information System (INIS)

2013-01-01

The Codes and Standards Working Group (CSWG) is one of the issue-specific working groups that the MDEP members are undertaking; its long term goal is harmonisation of regulatory and code requirements for design and construction of pressure-retaining components in order to improve the effectiveness and efficiency of the regulatory design reviews, increase quality of safety assessments, and to enable each regulator to become stronger in its ability to make safety decisions. The CSWG has interacted closely with the Standards Development Organisations (SDOs) and CORDEL in code comparison and code convergence. The Code Comparison Report STP-NU-051 has been issued by SDO members to identify the extent of similarities and differences amongst the pressure-boundary codes and standards used in various countries. Besides the differences in codes and standards, the way how the codes and standards are applied to systems, structures and components also affects the design and construction of nuclear power plant. Therefore, to accomplish the goal of potential harmonisation, it is also vital that the regulators learn about each other's procedures, processes, and regulations. To facilitate the learning process, the CSWG meets regularly to discuss issues relevant to licensing new reactors and using codes and standards in licensing safety reviews. The CSWG communicates very frequently with the SDOs to discuss similarities and differences among the various codes and how to proceed with potential harmonisation. It should be noted that the IAEA is invited to all of the issue-specific working groups within MDEP to ensure consistency with IAEA standards. The primary focus of this technical report is to consolidate information shared and accomplishments achieved by the member countries. This report seeks to document how each MDEP regulator utilises national or regional mechanical codes and standards in its safety reviews and licensing of new reactors. The preparation of this report

Hydrogen recycle modeling in transport codes

International Nuclear Information System (INIS)

Howe, H.C.

1979-01-01

The hydrogen recycling models now used in Tokamak transport codes are reviewed and the method by which realistic recycling models are being added is discussed. Present models use arbitrary recycle coefficients and therefore do not model the actual recycling processes at the wall. A model for the hydrogen concentration in the wall serves two purposes: (1) it allows a better understanding of the density behavior in present gas puff, pellet, and neutral beam heating experiments; and (2) it allows one to extrapolate to long pulse devices such as EBT, ISX-C and reactors where the walls are observed or expected to saturate. Several wall models are presently being studied for inclusion in transport codes

Model comparisons of the reactive burn model SURF in three ASC codes

Energy Technology Data Exchange (ETDEWEB)

Whitley, Von Howard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stalsberg, Krista Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reichelt, Benjamin Lee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Shipley, Sarah Jayne [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2018-01-12

A study of the SURF reactive burn model was performed in FLAG, PAGOSA and XRAGE. In this study, three different shock-to-detonation transition experiments were modeled in each code. All three codes produced similar model results for all the experiments modeled and at all resolutions. Buildup-to-detonation time, particle velocities and resolution dependence of the models was notably similar between the codes. Given the current PBX 9502 equations of state and SURF calibrations, each code is equally capable of predicting the correct detonation time and distance when impacted by a 1D impactor at pressures ranging from 10-16 GPa, as long as the resolution of the mesh is not too coarse.

40 CFR 194.23 - Models and computer codes.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 24 2010-07-01 2010-07-01 false Models and computer codes. 194.23... General Requirements § 194.23 Models and computer codes. (a) Any compliance application shall include: (1... obtain stable solutions; (iv) Computer models accurately implement the numerical models; i.e., computer...

Cavitation Modeling in Euler and Navier-Stokes Codes

Science.gov (United States)

Deshpande, Manish; Feng, Jinzhang; Merkle, Charles L.

1993-01-01

Many previous researchers have modeled sheet cavitation by means of a constant pressure solution in the cavity region coupled with a velocity potential formulation for the outer flow. The present paper discusses the issues involved in extending these cavitation models to Euler or Navier-Stokes codes. The approach taken is to start from a velocity potential model to ensure our results are compatible with those of previous researchers and available experimental data, and then to implement this model in both Euler and Navier-Stokes codes. The model is then augmented in the Navier-Stokes code by the inclusion of the energy equation which allows the effect of subcooling in the vicinity of the cavity interface to be modeled to take into account the experimentally observed reduction in cavity pressures that occurs in cryogenic fluids such as liquid hydrogen. Although our goal is to assess the practicality of implementing these cavitation models in existing three-dimensional, turbomachinery codes, the emphasis in the present paper will center on two-dimensional computations, most specifically isolated airfoils and cascades. Comparisons between velocity potential, Euler and Navier-Stokes implementations indicate they all produce consistent predictions. Comparisons with experimental results also indicate that the predictions are qualitatively correct and give a reasonable first estimate of sheet cavitation effects in both cryogenic and non-cryogenic fluids. The impact on CPU time and the code modifications required suggests that these models are appropriate for incorporation in current generation turbomachinery codes.

Measuring and Modeling the U.S. Regulatory Ecosystem

Science.gov (United States)

Bommarito, Michael J., II; Katz, Daniel Martin

2017-09-01

Over the last 23 years, the U.S. Securities and Exchange Commission has required over 34,000 companies to file over 165,000 annual reports. These reports, the so-called "Form 10-Ks," contain a characterization of a company's financial performance and its risks, including the regulatory environment in which a company operates. In this paper, we analyze over 4.5 million references to U.S. Federal Acts and Agencies contained within these reports to measure the regulatory ecosystem, in which companies are organisms inhabiting a regulatory environment. While individuals across the political, economic, and academic world frequently refer to trends in this regulatory ecosystem, far less attention has been paid to supporting such claims with large-scale, longitudinal data. In this paper, in addition to positing a model of regulatory ecosystems, we document an increase in the regulatory energy per filing, i.e., a warming "temperature." We also find that the diversity of the regulatory ecosystem has been increasing over the past two decades. These findings support the claim that regulatory activity and complexity are increasing, and this framework contributes an important step towards improving academic and policy discussions around legal complexity and regulation.

CONTAIN code analyses of direct containment heating experiments

International Nuclear Information System (INIS)

Williams, D.C.; Griffith, R.O.; Tadios, E.L.; Washington, K.E.

1995-01-01

In some nuclear reactor core-melt accidents, a potential exists for molten core-debris to be dispersed into the containment under high pressure. Resulting energy transfer to the containment atmosphere can pressurize the containment. This process, known as direct containment heating (DCH), has been the subject of extensive experimental and analytical programs sponsored by the U.S. Nuclear Regulatory Commission (NRC). The DCH modeling has been an important focus for the development of the CONTAIN code. Results of a detailed independent peer review of the CONTAIN code were published recently. This paper summarizes work performed in support of the peer review in which the CONTAIN code was applied to analyze DCH experiments. Goals of this work were comparison of calculated and experimental results, CONTAIN DCH model assessment, and development of guidance for code users, including development of a standardized input prescription for DCH analysis

Mainstreaming solar : Stretching the regulatory regime through business model innovation

NARCIS (Netherlands)

Huijben, J.C.C.M.; Verbong, G.P.J.; Podoynitsyna, K.S.

This paper explores how the regulatory regime for Solar PV, defined as a combination of niche shielding and mainstream regulations, affects niche business models, using the Dutch and Flemish regulatory regimes as examples. The regulatory regime does not influence all components of the business

Systematic review regulatory principles of non-coding RNAs in cardiovascular diseases.

Science.gov (United States)

Li, Yongsheng; Huo, Caiqin; Pan, Tao; Li, Lili; Jin, Xiyun; Lin, Xiaoyu; Chen, Juan; Zhang, Jinwen; Guo, Zheng; Xu, Juan; Li, Xia

2017-08-16

Cardiovascular diseases (CVDs) continue to be a major cause of morbidity and mortality, and non-coding RNAs (ncRNAs) play critical roles in CVDs. With the recent emergence of high-throughput technologies, including small RNA sequencing, investigations of CVDs have been transformed from candidate-based studies into genome-wide undertakings, and a number of ncRNAs in CVDs were discovered in various studies. A comprehensive review of these ncRNAs would be highly valuable for researchers to get a complete picture of the ncRNAs in CVD. To address these knowledge gaps and clinical needs, in this review, we first discussed dysregulated ncRNAs and their critical roles in cardiovascular development and related diseases. Moreover, we reviewed >28 561 published papers and documented the ncRNA-CVD association benchmarking data sets to summarize the principles of ncRNA regulation in CVDs. This data set included 13 249 curated relationships between 9503 ncRNAs and 139 CVDs in 12 species. Based on this comprehensive resource, we summarized the regulatory principles of dysregulated ncRNAs in CVDs, including the complex associations between ncRNA and CVDs, tissue specificity and ncRNA synergistic regulation. The highlighted principles are that CVD microRNAs (miRNAs) are highly expressed in heart tissue and that they play central roles in miRNA-miRNA functional synergistic network. In addition, CVD-related miRNAs are close to one another in the functional network, indicating the modular characteristic features of CVD miRNAs. We believe that the regulatory principles summarized here will further contribute to our understanding of ncRNA function and dysregulation mechanisms in CVDs. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Collection of regulatory texts relative to radiation protection. Part 1: laws and decrees (Extracts of the Public Health Code and of the Labour Code dealing with the protection of population, patients and workers against the hazards of ionizing radiations

International Nuclear Information System (INIS)

Rivas, Robert; Feries, Jean; Marzorati, Frank; Chevalier, Celine; Lachaume, Jean-Luc

2013-01-01

This first part contains legal and regulatory texts extracted from the Public Health Code and related to health general protection and to health products (medical devices), from the Social Security Code, and from the Labour Code related to individual work relationships, to health and safety at work, to work places, to work equipment and means of protection, to the prevention of some exposure risks and of risks related to some activities. This document is an update of the previous version from January 25, 2011

Tardos fingerprinting codes in the combined digit model

NARCIS (Netherlands)

Skoric, B.; Katzenbeisser, S.; Schaathun, H.G.; Celik, M.U.

2009-01-01

We introduce a new attack model for collusion-secure codes, called the combined digit model, which represents signal processing attacks against the underlying watermarking level better than existing models. In this paper, we analyze the performance of two variants of the Tardos code and show that

Models and applications of the UEDGE code

International Nuclear Information System (INIS)

Rensink, M.E.; Knoll, D.A.; Porter, G.D.; Rognlien, T.D.; Smith, G.R.; Wising, F.

1996-09-01

The transport of particles and energy from the core of a tokamak to nearby material surfaces is an important problem for understanding present experiments and for designing reactor-grade devices. A number of fluid transport codes have been developed to model the plasma in the edge and scrape-off layer (SOL) regions. This report will focus on recent model improvements and illustrative results from the UEDGE code. Some geometric and mesh considerations are introduced, followed by a general description of the plasma and neutral fluid models. A few comments on computational issues are given and then two important applications are illustrated concerning benchmarking and the ITER radiative divertor. Finally, we report on some recent work to improve the models in UEDGE by coupling to a Monte Carlo neutrals code and by utilizing an adaptive grid

COMPBRN III: a computer code for modeling compartment fires

International Nuclear Information System (INIS)

Ho, V.; Siu, N.; Apostolakis, G.; Flanagan, G.F.

1986-07-01

The computer code COMPBRN III deterministically models the behavior of compartment fires. This code is an improvement of the original COMPBRN codes. It employs a different air entrainment model and numerical scheme to estimate properties of the ceiling hot gas layer model. Moreover, COMPBRN III incorporates a number of improvements in shape factor calculations and error checking, which distinguish it from the COMPBRN II code. This report presents the ceiling hot gas layer model employed by COMPBRN III as well as several other modifications. Information necessary to run COMPBRN III, including descriptions of required input and resulting output, are also presented. Simulation of experiments and a sample problem are included to demonstrate the usage of the code. 37 figs., 46 refs

Chemistry models in the Victoria code

International Nuclear Information System (INIS)

Grimley, A.J. III

1988-01-01

The VICTORIA Computer code consists of the fission product release and chemistry models for the MELPROG severe accident analysis code. The chemistry models in VICTORIA are used to treat multi-phase interactions in four separate physical regions: fuel grains, gap/open porosity/clad, coolant/aerosols, and structure surfaces. The physical and chemical environment of each region is very different from the others and different models are required for each. The common thread in the modelling is the use of a chemical equilibrium assumption. The validity of this assumption along with a description of the various physical constraints applicable to each region will be discussed. The models that result from the assumptions and constraints will be presented along with samples of calculations in each region

Software requirements specification document for the AREST code development

International Nuclear Information System (INIS)

Engel, D.W.; McGrail, B.P.; Whitney, P.D.; Gray, W.J.; Williford, R.E.; White, M.D.; Eslinger, P.W.; Altenhofen, M.K.

1993-11-01

The Analysis of the Repository Source Term (AREST) computer code was selected in 1992 by the U.S. Department of Energy. The AREST code will be used to analyze the performance of an underground high level nuclear waste repository. The AREST code is being modified by the Pacific Northwest Laboratory (PNL) in order to evaluate the engineered barrier and waste package designs, model regulatory compliance, analyze sensitivities, and support total systems performance assessment modeling. The current version of the AREST code was developed to be a very useful tool for analyzing model uncertainties and sensitivities to input parameters. The code has also been used successfully in supplying source-terms that were used in a total systems performance assessment. The current version, however, has been found to be inadequate for the comparison and selection of a design for the waste package. This is due to the assumptions and simplifications made in the selection of the process and system models. Thus, the new version of the AREST code will be designed to focus on the details of the individual processes and implementation of more realistic models. This document describes the requirements of the new models that will be implemented. Included in this document is a section describing the near-field environmental conditions for this waste package modeling, description of the new process models that will be implemented, and a description of the computer requirements for the new version of the AREST code

Sodium/water pool-deposit bed model of the CONACS code

International Nuclear Information System (INIS)

Peak, R.D.

1983-01-01

A new Pool-Bed model of the CONACS (Containment Analysis Code System) code represents a major advance over the pool models of other containment analysis code (NABE code of France, CEDAN code of Japan and CACECO and CONTAIN codes of the United States). This new model advances pool-bed modeling because of the number of significant materials and processes which are included with appropriate rigor. This CONACS pool-bed model maintains material balances for eight chemical species (C, H 2 O, Na, NaH, Na 2 O, Na 2 O 2 , Na 2 CO 3 and NaOH) that collect in the stationary liquid pool on the floor and in the desposit bed on the elevated shelf of the standard CONACS analysis cell

Sparsity in Model Gene Regulatory Networks

International Nuclear Information System (INIS)

Zagorski, M.

2011-01-01

We propose a gene regulatory network model which incorporates the microscopic interactions between genes and transcription factors. In particular the gene's expression level is determined by deterministic synchronous dynamics with contribution from excitatory interactions. We study the structure of networks that have a particular '' function '' and are subject to the natural selection pressure. The question of network robustness against point mutations is addressed, and we conclude that only a small part of connections defined as '' essential '' for cell's existence is fragile. Additionally, the obtained networks are sparse with narrow in-degree and broad out-degree, properties well known from experimental study of biological regulatory networks. Furthermore, during sampling procedure we observe that significantly different genotypes can emerge under mutation-selection balance. All the preceding features hold for the model parameters which lay in the experimentally relevant range. (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

Transmutation Fuel Performance Code Thermal Model Verification

Energy Technology Data Exchange (ETDEWEB)

Gregory K. Miller; Pavel G. Medvedev

2007-09-01

FRAPCON fuel performance code is being modified to be able to model performance of the nuclear fuels of interest to the Global Nuclear Energy Partnership (GNEP). The present report documents the effort for verification of the FRAPCON thermal model. It was found that, with minor modifications, FRAPCON thermal model temperature calculation agrees with that of the commercial software ABAQUS (Version 6.4-4). This report outlines the methodology of the verification, code input, and calculation results.

Improving the quality of clinical coding: a comprehensive audit model

Directory of Open Access Journals (Sweden)

Hamid Moghaddasi

2014-04-01

Full Text Available Introduction: The review of medical records with the aim of assessing the quality of codes has long been conducted in different countries. Auditing medical coding, as an instructive approach, could help to review the quality of codes objectively using defined attributes, and this in turn would lead to improvement of the quality of codes. Method: The current study aimed to present a model for auditing the quality of clinical codes. The audit model was formed after reviewing other audit models, considering their strengths and weaknesses. A clear definition was presented for each quality attribute and more detailed criteria were then set for assessing the quality of codes. Results: The audit tool (based on the quality attributes included legibility, relevancy, completeness, accuracy, definition and timeliness; led to development of an audit model for assessing the quality of medical coding. Delphi technique was then used to reassure the validity of the model. Conclusion: The inclusive audit model designed could provide a reliable and valid basis for assessing the quality of codes considering more quality attributes and their clear definition. The inter-observer check suggested in the method of auditing is of particular importance to reassure the reliability of coding.

Operating nuclear plant feedback to ASME and French codes

International Nuclear Information System (INIS)

Journet, J.; O'Donnell, W.J.

1996-01-01

The French have an advantage in nuclear plant operating experience feedback due to the highly centralized nature of their nuclear industry. There is only one utility in charge of design as well as operations (EDF) and only one reactor vendor (Framatome). The ASME Code has played a key role in resolving technical issues in the design and operation of nuclear plants since the inception of nuclear power. The committee structure of the Code brings an ideal combination of senior technical people with both broad and specialized experience to bear on complex how safe is safe enough technical issues. The authors now see an even greater role for the ASME Code in a proposed new regulatory era for the US nuclear industry. The current legalistic confrontational regulatory era has been quite destructive. There now appears to be a real opportunity to begin a new era of technical consensus as the primary means for resolving safety issues. This change can quickly be brought about by having the industry take operating plant problems and regulatory technical issues directly to the ASME Code for timely resolution. Surprisingly, there is no institution in the US nuclear industry with such a mandate. In fact, the industry is organized to feedback through the Nuclear Regulatory Commission issues which could be far better resolved through the ASME Code. Major regulatory benefits can be achieved by closing this loop and providing systematic interaction with the ASME Code. The essential elements of a new regulatory era and ideas for organizing US institutional industry responsibilities, taken from the French experience, are described in this paper

Proceedings of the workshop on advanced thermal-hydraulic and neutronic codes: current and future applications

International Nuclear Information System (INIS)

2001-01-01

An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for

Review of SFR Design Safety using Preliminary Regulatory PSA Model

International Nuclear Information System (INIS)

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

2013-01-01

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

Assessment of compliance with regulatory requirements for a best estimate methodology for evaluation of ECCS

Energy Technology Data Exchange (ETDEWEB)

Lee, Un Chul; Jang, Jin Wook; Lim, Ho Gon; Jeong, Ik [Seoul National Univ., Seoul (Korea, Republic of); Sim, Suk Ku [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2000-03-15

Best estimate methodology for evaluation of ECCS proposed by KEPCO(KREM) os using thermal-hydraulic best-estimate code and the topical report for the methodology is described that it meets the regulatory requirement of USNRC regulatory guide. In this research the assessment of compliance with regulatory guide. In this research the assessment of compliance with regulatory requirements for the methodology is performed. The state of licensing procedure of other countries and best-estimate evaluation methodologies of Europe is also investigated, The applicability of models and propriety of procedure of uncertainty analysis of KREM are appraised and compliance with USNRC regulatory guide is assessed.

Conservation of concrete structures according to fib Model Code 2010

NARCIS (Netherlands)

Matthews, S.; Bigaj-Van Vliet, A.; Ueda, T.

2013-01-01

Conservation of concrete structures forms an essential part of the fib Model Code for Concrete Structures 2010 (fib Model Code 2010). In particular, Chapter 9 of fib Model Code 2010 addresses issues concerning conservation strategies and tactics, conservation management, condition surveys, condition

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

Computer code selection criteria for flow and transport code(s) to be used in undisturbed vadose zone calculations for TWRS environmental analyses

International Nuclear Information System (INIS)

Mann, F.M.

1998-01-01

The Tank Waste Remediation System (TWRS) is responsible for the safe storage, retrieval, and disposal of waste currently being held in 177 underground tanks at the Hanford Site. In order to successfully carry out its mission, TWRS must perform environmental analyses describing the consequences of tank contents leaking from tanks and associated facilities during the storage, retrieval, or closure periods and immobilized low-activity tank waste contaminants leaving disposal facilities. Because of the large size of the facilities and the great depth of the dry zone (known as the vadose zone) underneath the facilities, sophisticated computer codes are needed to model the transport of the tank contents or contaminants. This document presents the code selection criteria for those vadose zone analyses (a subset of the above analyses) where the hydraulic properties of the vadose zone are constant in time the geochemical behavior of the contaminant-soil interaction can be described by simple models, and the geologic or engineered structures are complicated enough to require a two-or three dimensional model. Thus, simple analyses would not need to use the fairly sophisticated codes which would meet the selection criteria in this document. Similarly, those analyses which involve complex chemical modeling (such as those analyses involving large tank leaks or those analyses involving the modeling of contaminant release from glass waste forms) are excluded. The analyses covered here are those where the movement of contaminants can be relatively simply calculated from the moisture flow. These code selection criteria are based on the information from the low-level waste programs of the US Department of Energy (DOE) and of the US Nuclear Regulatory Commission as well as experience gained in the DOE Complex in applying these criteria. Appendix table A-1 provides a comparison between the criteria in these documents and those used here. This document does not define the models (that

Methodology, status, and plans for development and assessment of the RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Johnson, G.W.; Riemke, R.A. [Idaho National Engineering Laboratory, Idaho Falls, ID (United States)

1997-07-01

RELAP/MOD3 is a computer code used for the simulation of transients and accidents in light-water nuclear power plants. The objective of the program to develop and maintain RELAP5 was and is to provide the U.S. Nuclear Regulatory Commission with an independent tool for assessing reactor safety. This paper describes code requirements, models, solution scheme, language and structure, user interface validation, and documentation. The paper also describes the current and near term development program and provides an assessment of the code`s strengths and limitations.

Fuel rod modelling during transients: The TOUTATIS code

International Nuclear Information System (INIS)

Bentejac, F.; Bourreau, S.; Brochard, J.; Hourdequin, N.; Lansiart, S.

2001-01-01

The TOUTATIS code is devoted to the PCI local phenomena simulation, in correlation with the METEOR code for the global behaviour of the fuel rod. More specifically, the TOUTATIS objective is to evaluate the mechanical constraints on the cladding during a power transient thus predicting its behaviour in term of stress corrosion cracking. Based upon the finite element computation code CASTEM 2000, TOUTATIS is a set of modules written in a macro language. The aim of this paper is to present both code modules: The axisymmetric bi-dimensional module, modeling a unique block pellet; The tri dimensional module modeling a radially fragmented pellet. Having shown the boundary conditions and the algorithms used, the application will be illustrated by: A short presentation of the bidimensional axisymmetric modeling performances as well as its limits; The enhancement due to the three dimensional modeling will be displayed by sensitivity studies to the geometry, in this case the pellet height/diameter ratio. Finally, we will show the easiness of the development inherent to the CASTEM 2000 system by depicting the process of a modeling enhancement by adding the possibility of an axial (horizontal) fissuration of the pellet. As conclusion, the future improvements planned for the code are depicted. (author)

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

Human population doses: Comparative analysis of CREAM code results with currently computer codes of Nuclear Regulatory Authority; Dosis en la poblacion: comparacion de los resultados del codigo CREAM con resultados de modelos vigentes en la ARN

Energy Technology Data Exchange (ETDEWEB)

Alonso Jimenez, Maria Teresa; Curti, Adriana [Autoridad Regulatoria Nuclear, Buenos Aires (Argentina)]. E-mail: mtalonso@sede.arn.gov.ar; acurti@sede.arn.gov.ar

2001-07-01

The Nuclear Regulatory Authority is performing an analysis with PC CREAM, developed at the NRPB, for updating computer programs and models used for calculating the transfer of radionuclides through the environment. For CREAM dose assessment verification for local scenarios, this paper presents a comparison of population doses assessed with the computer codes used nowadays and with CREAM, for unitary releases of main radionuclides in nuclear power plant discharges. The results of atmospheric dispersion processes and the transfer of radionuclides through the environment for local scenarios are analysed. The programs used are PLUME for atmospheric dispersion, FARMLAND for the transfer of radionuclides into foodstuffs following atmospheric deposition in the terrestrial environment and ASSESSOR for individual and collective dose assessments.This paper presents the general assumptions made for dose assessments. The results show some differences between doses due to differences in models, in the complexity level of the same models, or in parameters. (author)

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....

Properties of non-coding DNA and identification of putative cis-regulatory elements in Theileria parva

Directory of Open Access Journals (Sweden)

Guo Xiang

2008-12-01

Full Text Available Abstract Background Parasites in the genus Theileria cause lymphoproliferative diseases in cattle, resulting in enormous socio-economic losses. The availability of the genome sequences and annotation for T. parva and T. annulata has facilitated the study of parasite biology and their relationship with host cell transformation and tropism. However, the mechanism of transcriptional regulation in this genus, which may be key to understanding fundamental aspects of its parasitology, remains poorly understood. In this study, we analyze the evolution of non-coding sequences in the Theileria genome and identify conserved sequence elements that may be involved in gene regulation of these parasitic species. Results Intergenic regions and introns in Theileria are short, and their length distributions are considerably right-skewed. Intergenic regions flanked by genes in 5'-5' orientation tend to be longer and slightly more AT-rich than those flanked by two stop codons; intergenic regions flanked by genes in 3'-5' orientation have intermediate values of length and AT composition. Intron position is negatively correlated with intron length, and positively correlated with GC content. Using stringent criteria, we identified a set of high-quality orthologous non-coding sequences between T. parva and T. annulata, and determined the distribution of selective constraints across regions, which are shown to be higher close to translation start sites. A positive correlation between constraint and length in both intergenic regions and introns suggests a tight control over length expansion of non-coding regions. Genome-wide searches for functional elements revealed several conserved motifs in intergenic regions of Theileria genomes. Two such motifs are preferentially located within the first 60 base pairs upstream of transcription start sites in T. parva, are preferentially associated with specific protein functional categories, and have significant similarity to know

Rulemaking efforts on codes and standards

International Nuclear Information System (INIS)

Millman, G.C.

1992-01-01

Section 50.55a of the NRC regulations provides a mechanism for incorporating national codes and standards into the regulatory process. It incorporates by reference ASME Boiler and Pressure Vessel Code (ASME B and PV Code) Section 3 rules for construction and Section 11 rules for inservice inspection and inservice testing. The regulation is periodically amended to update these references. The rulemaking process, as applied to Section 50.55a amendments, is overviewed to familiarize users with associated internal activities of the NRC staff and the manner in which public comments are integrated into the process. The four ongoing rulemaking actions that would individually amend Section 50.55a are summarized. Two of the actions would directly impact requirements for inservice testing. Benefits accrued with NRC endorsement of the ASME B and PV Code, and possible future endorsement of the ASME Operations and Maintenance Code (ASME OM Code), are identified. Emphasis is placed on the need for code writing committees to be especially sensitive to user feedback on code rules incorporated into the regulatory process to ensure that the rules are complete, technically accurate, clear, practical, and enforceable

Mutual information and the fidelity of response of gene regulatory models

International Nuclear Information System (INIS)

Tabbaa, Omar P; Jayaprakash, C

2014-01-01

We investigate cellular response to extracellular signals by using information theory techniques motivated by recent experiments. We present results for the steady state of the following gene regulatory models found in both prokaryotic and eukaryotic cells: a linear transcription-translation model and a positive or negative auto-regulatory model. We calculate both the information capacity and the mutual information exactly for simple models and approximately for the full model. We find that (1) small changes in mutual information can lead to potentially important changes in cellular response and (2) there are diminishing returns in the fidelity of response as the mutual information increases. We calculate the information capacity using Gillespie simulations of a model for the TNF-α-NF-κ B network and find good agreement with the measured value for an experimental realization of this network. Our results provide a quantitative understanding of the differences in cellular response when comparing experimentally measured mutual information values of different gene regulatory models. Our calculations demonstrate that Gillespie simulations can be used to compute the mutual information of more complex gene regulatory models, providing a potentially useful tool in synthetic biology. (paper)

Lnc2Meth: a manually curated database of regulatory relationships between long non-coding RNAs and DNA methylation associated with human disease.

Science.gov (United States)

Zhi, Hui; Li, Xin; Wang, Peng; Gao, Yue; Gao, Baoqing; Zhou, Dianshuang; Zhang, Yan; Guo, Maoni; Yue, Ming; Shen, Weitao; Ning, Shangwei; Jin, Lianhong; Li, Xia

2018-01-04

Lnc2Meth (http://www.bio-bigdata.com/Lnc2Meth/), an interactive resource to identify regulatory relationships between human long non-coding RNAs (lncRNAs) and DNA methylation, is not only a manually curated collection and annotation of experimentally supported lncRNAs-DNA methylation associations but also a platform that effectively integrates tools for calculating and identifying the differentially methylated lncRNAs and protein-coding genes (PCGs) in diverse human diseases. The resource provides: (i) advanced search possibilities, e.g. retrieval of the database by searching the lncRNA symbol of interest, DNA methylation patterns, regulatory mechanisms and disease types; (ii) abundant computationally calculated DNA methylation array profiles for the lncRNAs and PCGs; (iii) the prognostic values for each hit transcript calculated from the patients clinical data; (iv) a genome browser to display the DNA methylation landscape of the lncRNA transcripts for a specific type of disease; (v) tools to re-annotate probes to lncRNA loci and identify the differential methylation patterns for lncRNAs and PCGs with user-supplied external datasets; (vi) an R package (LncDM) to complete the differentially methylated lncRNAs identification and visualization with local computers. Lnc2Meth provides a timely and valuable resource that can be applied to significantly expand our understanding of the regulatory relationships between lncRNAs and DNA methylation in various human diseases. © The Author(s) 2017. Published by Oxford University Press on behalf of Nucleic Acids Research.

Structure of fuel performance audit code for SFR metal fuel

Energy Technology Data Exchange (ETDEWEB)

Yang, Yong Sik; Kim, Hyo Chan [KAERI, Daejeon (Korea, Republic of); Jeong, Hye Dong; Shin, An Dong; Suh, Nam Duk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2012-10-15

A Sodium Cooled Fast Reactor (SFR) is a promising option to solve the spent fuel problems, but, there are still much technical issues to commercialize a SFR. One of issues is a development of advanced fuel which can solve the safety and the economic issues at the same time. Since a nuclear fuel is the first barrier to protect radioactive isotope release, the fuel's integrity must be secured. In Korea Institute of Nuclear Safety (KINS), the new project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. To develop the new code system, the code structure design and its requirements need to be studied. Various performance models and code systems are reviewed and their characteristics are analyzed in this paper. Based on this study, the fundamental performance models are deduced and basic code requirements and structure are established.

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.

Regulatory inspection of BARC facilities

International Nuclear Information System (INIS)

Rajdeep; Jayarajan, K.

2017-01-01

Nuclear and radiation facilities are sited, constructed, commissioned, operated and decommissioned, in conformity with the current safety standards and codes. Regulatory bodies follow different means to ensure compliance of the standards for the safety of the personnel, the public and the environment. Regulatory Inspection (RI) is one of the important measures employed by regulatory bodies to obtain the safety status of a facility or project and to verify the fulfilment of the conditions stipulated in the consent

Final Technical Report for GO17004 Regulatory Logic: Codes and Standards for the Hydrogen Economy

Energy Technology Data Exchange (ETDEWEB)

Nakarado, Gary L. [Regulatory Logic LLC, Golden, CO (United States)

2017-02-22

The objectives of this project are to: develop a robust supporting research and development program to provide critical hydrogen behavior data and a detailed understanding of hydrogen combustion and safety across a range of scenarios, needed to establish setback distances in building codes and minimize the overall data gaps in code development; support and facilitate the completion of technical specifications by the International Organization for Standardization (ISO) for gaseous hydrogen refueling (TS 20012) and standards for on-board liquid (ISO 13985) and gaseous or gaseous blend (ISO 15869) hydrogen storage by 2007; support and facilitate the effort, led by the NFPA, to complete the draft Hydrogen Technologies Code (NFPA 2) by 2008; with experimental data and input from Technology Validation Program element activities, support and facilitate the completion of standards for bulk hydrogen storage (e.g., NFPA 55) by 2008; facilitate the adoption of the most recently available model codes (e.g., from the International Code Council [ICC]) in key regions; complete preliminary research and development on hydrogen release scenarios to support the establishment of setback distances in building codes and provide a sound basis for model code development and adoption; support and facilitate the development of Global Technical Regulations (GTRs) by 2010 for hydrogen vehicle systems under the United Nations Economic Commission for Europe, World Forum for Harmonization of Vehicle Regulations and Working Party on Pollution and Energy Program (ECE-WP29/GRPE); and to Support and facilitate the completion by 2012 of necessary codes and standards needed for the early commercialization and market entry of hydrogen energy technologies.

MELCOR code modeling for APR1400

Energy Technology Data Exchange (ETDEWEB)

Choi, Young; Park, S. Y.; Kim, D. H.; Ahn, K. I.; Song, Y. M.; Kim, S. D.; Park, J. H

2001-11-01

The severe accident phenomena of nuclear power plant have large uncertainties. For the retention of the containment integrity and improvement of nuclear reactor safety against severe accident, it is essential to understand severe accident phenomena and be able to access the accident progression accurately using computer code. Furthermore, it is important to attain a capability for developing technique and assessment tools for an advanced nuclear reactor design as well as for the severe accident prevention and mitigation. The objective of this report is to establish technical bases for an application of the MELCOR code to the Korean Next Generation Reactor (APR1400) by modeling the plant and analyzing plant steady state. This report shows the data and the input preparation for MELCOR code as well as state-state assessment results using MELCOR code.

Development of the PRO-LOCA Probabilistic Fracture Mechanics Code, MERIT Final Report

International Nuclear Information System (INIS)

Scott, Paul; Kurth, Robert; Cox, Andrew; Olson, Rick; Rudland, Dave

2010-12-01

The MERIT project has been an internationally financed program with the main purpose of developing probabilistic models for piping failure of nuclear components and to include these models in a probabilistic code named PRO-LOCA. The principal objective of the project has been to develop probabilistic models for piping failure of nuclear components and to include these models in a probabilistic code. The MERIT program has produced a code named PRO-LOCA with the following features: - Crack initiation models for fatigue or stress corrosion cracking for previously unflawed material. - Subcritical crack growth models for fatigue and stress corrosion cracking for both initiated and pre-existing circumferential defects. - Models for flaw detection by inspections and leak detection. - Crack stability. The PRO-LOCA code can thus predict the leak or break frequency for the whole sequence of initiation, subcritical crack growth until wall penetration and leakage, instability of the through-wall crack (pipe rupture). The outcome of the PRO-LOCA code are a sequence of failure frequencies which represents the probability of surface crack developing, a through-wall crack developing and six different sizes of crack opening areas corresponding to different leak flow rates or LOCA categories. Note that the level of quality assurance of the PRO-LOCA code is such that the code in its current state of development is considered to be more of a research code than a regulatory tool.

Development of the PRO-LOCA Probabilistic Fracture Mechanics Code, MERIT Final Report

Energy Technology Data Exchange (ETDEWEB)

Scott, Paul; Kurth, Robert; Cox, Andrew; Olson, Rick (Battelle Columbus (United States)); Rudland, Dave (Nuclear Regulatory Commission (United States))

2010-12-15

The MERIT project has been an internationally financed program with the main purpose of developing probabilistic models for piping failure of nuclear components and to include these models in a probabilistic code named PRO-LOCA. The principal objective of the project has been to develop probabilistic models for piping failure of nuclear components and to include these models in a probabilistic code. The MERIT program has produced a code named PRO-LOCA with the following features: - Crack initiation models for fatigue or stress corrosion cracking for previously unflawed material. - Subcritical crack growth models for fatigue and stress corrosion cracking for both initiated and pre-existing circumferential defects. - Models for flaw detection by inspections and leak detection. - Crack stability. The PRO-LOCA code can thus predict the leak or break frequency for the whole sequence of initiation, subcritical crack growth until wall penetration and leakage, instability of the through-wall crack (pipe rupture). The outcome of the PRO-LOCA code are a sequence of failure frequencies which represents the probability of surface crack developing, a through-wall crack developing and six different sizes of crack opening areas corresponding to different leak flow rates or LOCA categories. Note that the level of quality assurance of the PRO-LOCA code is such that the code in its current state of development is considered to be more of a research code than a regulatory tool.

Fuel behavior modeling using the MARS computer code

International Nuclear Information System (INIS)

Faya, S.C.S.; Faya, A.J.G.

1983-01-01

The fuel behaviour modeling code MARS against experimental data, was evaluated. Two cases were selected: an early comercial PWR rod (Maine Yankee rod) and an experimental rod from the Canadian BWR program (Canadian rod). The MARS predictions are compared with experimental data and predictions made by other fuel modeling codes. Improvements are suggested for some fuel behaviour models. Mars results are satisfactory based on the data available. (Author) [pt

Genetic coding and gene expression - new Quadruplet genetic coding model

Science.gov (United States)

Shankar Singh, Rama

2012-07-01

Successful demonstration of human genome project has opened the door not only for developing personalized medicine and cure for genetic diseases, but it may also answer the complex and difficult question of the origin of life. It may lead to making 21st century, a century of Biological Sciences as well. Based on the central dogma of Biology, genetic codons in conjunction with tRNA play a key role in translating the RNA bases forming sequence of amino acids leading to a synthesized protein. This is the most critical step in synthesizing the right protein needed for personalized medicine and curing genetic diseases. So far, only triplet codons involving three bases of RNA, transcribed from DNA bases, have been used. Since this approach has several inconsistencies and limitations, even the promise of personalized medicine has not been realized. The new Quadruplet genetic coding model proposed and developed here involves all four RNA bases which in conjunction with tRNA will synthesize the right protein. The transcription and translation process used will be the same, but the Quadruplet codons will help overcome most of the inconsistencies and limitations of the triplet codes. Details of this new Quadruplet genetic coding model and its subsequent potential applications including relevance to the origin of life will be presented.

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.

Genome-wide prediction of cis-regulatory regions using supervised deep learning methods.

Science.gov (United States)

Li, Yifeng; Shi, Wenqiang; Wasserman, Wyeth W

2018-05-31

In the human genome, 98% of DNA sequences are non-protein-coding regions that were previously disregarded as junk DNA. In fact, non-coding regions host a variety of cis-regulatory regions which precisely control the expression of genes. Thus, Identifying active cis-regulatory regions in the human genome is critical for understanding gene regulation and assessing the impact of genetic variation on phenotype. The developments of high-throughput sequencing and machine learning technologies make it possible to predict cis-regulatory regions genome wide. Based on rich data resources such as the Encyclopedia of DNA Elements (ENCODE) and the Functional Annotation of the Mammalian Genome (FANTOM) projects, we introduce DECRES based on supervised deep learning approaches for the identification of enhancer and promoter regions in the human genome. Due to their ability to discover patterns in large and complex data, the introduction of deep learning methods enables a significant advance in our knowledge of the genomic locations of cis-regulatory regions. Using models for well-characterized cell lines, we identify key experimental features that contribute to the predictive performance. Applying DECRES, we delineate locations of 300,000 candidate enhancers genome wide (6.8% of the genome, of which 40,000 are supported by bidirectional transcription data), and 26,000 candidate promoters (0.6% of the genome). The predicted annotations of cis-regulatory regions will provide broad utility for genome interpretation from functional genomics to clinical applications. The DECRES model demonstrates potentials of deep learning technologies when combined with high-throughput sequencing data, and inspires the development of other advanced neural network models for further improvement of genome annotations.

Review on the seismic safety of JRR-3 according to the revised regulatory code on seismic design for nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Kobayashi, Tetsuya; Araki, Masaaki; Ohba, Toshinobu; Torii, Yoshiya [Japan Atomic Energy Agency, Tokai, Ibaraki (Japan); Takeuchi, Masaki [Nuclear Safety Commission (Japan)

2012-03-15

JRR-3(Japan Research Reactor No.3) with the thermal power of 20MW is a light water moderated and cooled, swimming pool type research reactor. JRR-3 has been operated without major troubles. This paper presents about review on the seismic safety of JRR-3 according to the revised regulatory code on seismic design for nuclear reactors. In addition, some topics concerning damages in JRR-3 due to the Great East Japan Earthquake are presented. (author)

Model uncertainty from a regulatory point of view

International Nuclear Information System (INIS)

Abramson, L.R.

1994-01-01

This paper discusses model uncertainty in the larger context of knowledge and random uncertainty. It explores some regulatory implications of model uncertainty and argues that, from a regulator's perspective, a conservative approach must be taken. As a consequence of this perspective, averaging over model results is ruled out

Transcriptomic Analysis of Long Non-Coding RNAs and Coding Genes Uncovers a Complex Regulatory Network That Is Involved in Maize Seed Development

Directory of Open Access Journals (Sweden)

Ming Zhu

2017-10-01

Full Text Available Long non-coding RNAs (lncRNAs have been reported to be involved in the development of maize plant. However, few focused on seed development of maize. Here, we identified 753 lncRNA candidates in maize genome from six seed samples. Similar to the mRNAs, lncRNAs showed tissue developmental stage specific and differential expression, indicating their putative role in seed development. Increasing evidence shows that crosstalk among RNAs mediated by shared microRNAs (miRNAs represents a novel layer of gene regulation, which plays important roles in plant development. Functional roles and regulatory mechanisms of lncRNAs as competing endogenous RNAs (ceRNA in plants, particularly in maize seed development, are unclear. We combined analyses of consistently altered 17 lncRNAs, 840 mRNAs and known miRNA to genome-wide investigate potential lncRNA-mediated ceRNA based on “ceRNA hypothesis”. The results uncovered seven novel lncRNAs as potential functional ceRNAs. Functional analyses based on their competitive coding-gene partners by Gene Ontology (GO and KEGG biological pathway demonstrated that combined effects of multiple ceRNAs can have major impacts on general developmental and metabolic processes in maize seed. These findings provided a useful platform for uncovering novel mechanisms of maize seed development and may provide opportunities for the functional characterization of individual lncRNA in future studies.

Code manual for MACCS2: Volume 1, user's guide

International Nuclear Information System (INIS)

Chanin, D.I.; Young, M.L.

1997-03-01

This report describes the use of the MACCS2 code. The document is primarily a user's guide, though some model description information is included. MACCS2 represents a major enhancement of its predecessor MACCS, the MELCOR Accident Consequence Code System. MACCS, distributed by government code centers since 1990, was developed to evaluate the impacts of severe accidents at nuclear power plants on the surrounding public. The principal phenomena considered are atmospheric transport and deposition under time-variant meteorology, short- and long-term mitigative actions and exposure pathways, deterministic and stochastic health effects, and economic costs. No other U.S. code that is publicly available at present offers all these capabilities. MACCS2 was developed as a general-purpose tool applicable to diverse reactor and nonreactor facilities licensed by the Nuclear Regulatory Commission or operated by the Department of Energy or the Department of Defense. The MACCS2 package includes three primary enhancements: (1) a more flexible emergency-response model, (2) an expanded library of radionuclides, and (3) a semidynamic food-chain model. Other improvements are in the areas of phenomenological modeling and new output options. Initial installation of the code, written in FORTRAN 77, requires a 486 or higher IBM-compatible PC with 8 MB of RAM

Improvement of a combustion model in MELCOR code

International Nuclear Information System (INIS)

Ogino, Masao; Hashimoto, Takashi

1999-01-01

NUPEC has been improving a hydrogen combustion model in MELCOR code for severe accident analysis. In the proposed combustion model, the flame velocity in a node was predicted using five different flame front shapes of fireball, prism, bubble, spherical jet, and plane jet. For validation of the proposed model, the results of the Battelle multi-compartment hydrogen combustion test were used. The selected test cases for the study were Hx-6, 13, 14, 20 and Ix-2 which had two, three or four compartments under homogeneous hydrogen concentration of 5 to 10 vol%. The proposed model could predict well the combustion behavior in multi-compartment containment geometry on the whole. MELCOR code, incorporating the present combustion model, can simulate combustion behavior during severe accident with acceptable computing time and some degree of accuracy. The applicability study of the improved MELCOR code to the actual reactor plants will be further continued. (author)

Development of CAP code for nuclear power plant containment: Lumped model

Energy Technology Data Exchange (ETDEWEB)

Hong, Soon Joon, E-mail: sjhong90@fnctech.com [FNC Tech. Co. Ltd., Heungdeok 1 ro 13, Giheung-gu, Yongin-si, Gyeonggi-do 446-908 (Korea, Republic of); Choo, Yeon Joon; Hwang, Su Hyun; Lee, Byung Chul [FNC Tech. Co. Ltd., Heungdeok 1 ro 13, Giheung-gu, Yongin-si, Gyeonggi-do 446-908 (Korea, Republic of); Ha, Sang Jun [Central Research Institute, Korea Hydro & Nuclear Power Company, Ltd., 70, 1312-gil, Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of)

2015-09-15

Highlights: • State-of-art containment analysis code, CAP, has been developed. • CAP uses 3-field equations, water level oriented upwind scheme, local head model. • CAP has a function of linked calculation with reactor coolant system code. • CAP code assessments showed appropriate prediction capabilities. - Abstract: CAP (nuclear Containment Analysis Package) code has been developed in Korean nuclear society for the analysis of nuclear containment thermal hydraulic behaviors including pressure and temperature trends and hydrogen concentration. Lumped model of CAP code uses 2-phase, 3-field equations for fluid behaviors, and has appropriate constitutive equations, 1-dimensional heat conductor model, component models, trip and control models, and special process models. CAP can run in a standalone mode or a linked mode with a reactor coolant system analysis code. The linked mode enables the more realistic calculation of a containment response and is expected to be applicable to a more complicated advanced plant design calculation. CAP code assessments were carried out by gradual approaches: conceptual problems, fundamental phenomena, component and principal phenomena, experimental validation, and finally comparison with other code calculations on the base of important phenomena identifications. The assessments showed appropriate prediction capabilities of CAP.

Development of CAP code for nuclear power plant containment: Lumped model

International Nuclear Information System (INIS)

Hong, Soon Joon; Choo, Yeon Joon; Hwang, Su Hyun; Lee, Byung Chul; Ha, Sang Jun

2015-01-01

Highlights: • State-of-art containment analysis code, CAP, has been developed. • CAP uses 3-field equations, water level oriented upwind scheme, local head model. • CAP has a function of linked calculation with reactor coolant system code. • CAP code assessments showed appropriate prediction capabilities. - Abstract: CAP (nuclear Containment Analysis Package) code has been developed in Korean nuclear society for the analysis of nuclear containment thermal hydraulic behaviors including pressure and temperature trends and hydrogen concentration. Lumped model of CAP code uses 2-phase, 3-field equations for fluid behaviors, and has appropriate constitutive equations, 1-dimensional heat conductor model, component models, trip and control models, and special process models. CAP can run in a standalone mode or a linked mode with a reactor coolant system analysis code. The linked mode enables the more realistic calculation of a containment response and is expected to be applicable to a more complicated advanced plant design calculation. CAP code assessments were carried out by gradual approaches: conceptual problems, fundamental phenomena, component and principal phenomena, experimental validation, and finally comparison with other code calculations on the base of important phenomena identifications. The assessments showed appropriate prediction capabilities of CAP

MARS CODE MANUAL VOLUME V: Models and Correlations

International Nuclear Information System (INIS)

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

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 models and correlations manual provides a complete list of detailed information of the thermal-hydraulic models used in MARS, so that 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

Challenges for modeling global gene regulatory networks during development: insights from Drosophila.

Science.gov (United States)

Wilczynski, Bartek; Furlong, Eileen E M

2010-04-15

Development is regulated by dynamic patterns of gene expression, which are orchestrated through the action of complex gene regulatory networks (GRNs). Substantial progress has been made in modeling transcriptional regulation in recent years, including qualitative "coarse-grain" models operating at the gene level to very "fine-grain" quantitative models operating at the biophysical "transcription factor-DNA level". Recent advances in genome-wide studies have revealed an enormous increase in the size and complexity or GRNs. Even relatively simple developmental processes can involve hundreds of regulatory molecules, with extensive interconnectivity and cooperative regulation. This leads to an explosion in the number of regulatory functions, effectively impeding Boolean-based qualitative modeling approaches. At the same time, the lack of information on the biophysical properties for the majority of transcription factors within a global network restricts quantitative approaches. In this review, we explore the current challenges in moving from modeling medium scale well-characterized networks to more poorly characterized global networks. We suggest to integrate coarse- and find-grain approaches to model gene regulatory networks in cis. We focus on two very well-studied examples from Drosophila, which likely represent typical developmental regulatory modules across metazoans. Copyright (c) 2009 Elsevier Inc. All rights reserved.

RELAP5/MOD2 code assessment

International Nuclear Information System (INIS)

Nithianandan, C.K.; Shah, N.H.; Schomaker, R.J.; Miller, F.R.

1985-01-01

Babcock and Wilcox (B and W) has been working with the code developers at EG and G and the US Nuclear Regulatory Commission in assessing the RELAP5/MOD2 computer code for the past year by simulating selected separate-effects tests. The purpose of this assessment has been to evaluate the code for use in MIST (Ref. 2) and OTIS integral system tests simulations and in the prediction of pressurized water reactor transients. B and W evaluated various versions of the code and made recommendations to improve code performance. As a result, the currently released version (cycle 36.1) has been improved considerably over earlier versions. However, further refinements to some of the constitutive models may still be needed to further improve the predictive capability of RELAP5/MOD2. The following versions of the code were evaluated. (1) RELAP/MOD2/Cycle 22 - first released version; (2) YELAP5/Cycle 32 - EG and G test version of RELAP5/MOD2/Cycle 32; (3) RELAP5/MOD2/Cycle 36 - frozen cycle for international code assessment; (4) updates to cycle 36 based on recommendations developed by B and W during the simulation of a Massachusetts Institute of Technology (MIT) pressurizer test; and (5) cycle 36.1 updates received from EG and G

RELAP5/MOD2 code assessment

Energy Technology Data Exchange (ETDEWEB)

Nithianandan, C.K.; Shah, N.H.; Schomaker, R.J.; Miller, F.R.

1985-11-01

Babcock and Wilcox (B and W) has been working with the code developers at EG and G and the US Nuclear Regulatory Commission in assessing the RELAP5/MOD2 computer code for the past year by simulating selected separate-effects tests. The purpose of this assessment has been to evaluate the code for use in MIST (Ref. 2) and OTIS integral system tests simulations and in the prediction of pressurized water reactor transients. B and W evaluated various versions of the code and made recommendations to improve code performance. As a result, the currently released version (cycle 36.1) has been improved considerably over earlier versions. However, further refinements to some of the constitutive models may still be needed to further improve the predictive capability of RELAP5/MOD2. The following versions of the code were evaluated. (1) RELAP/MOD2/Cycle 22 - first released version; (2) YELAP5/Cycle 32 - EG and G test version of RELAP5/MOD2/Cycle 32; (3) RELAP5/MOD2/Cycle 36 - frozen cycle for international code assessment; (4) updates to cycle 36 based on recommendations developed by B and W during the simulation of a Massachusetts Institute of Technology (MIT) pressurizer test; and (5) cycle 36.1 updates received from EG and G.

Modeling peripheral olfactory coding in Drosophila larvae.

Directory of Open Access Journals (Sweden)

Derek J Hoare

Full Text Available The Drosophila larva possesses just 21 unique and identifiable pairs of olfactory sensory neurons (OSNs, enabling investigation of the contribution of individual OSN classes to the peripheral olfactory code. We combined electrophysiological and computational modeling to explore the nature of the peripheral olfactory code in situ. We recorded firing responses of 19/21 OSNs to a panel of 19 odors. This was achieved by creating larvae expressing just one functioning class of odorant receptor, and hence OSN. Odor response profiles of each OSN class were highly specific and unique. However many OSN-odor pairs yielded variable responses, some of which were statistically indistinguishable from background activity. We used these electrophysiological data, incorporating both responses and spontaneous firing activity, to develop a bayesian decoding model of olfactory processing. The model was able to accurately predict odor identity from raw OSN responses; prediction accuracy ranged from 12%-77% (mean for all odors 45.2% but was always significantly above chance (5.6%. However, there was no correlation between prediction accuracy for a given odor and the strength of responses of wild-type larvae to the same odor in a behavioral assay. We also used the model to predict the ability of the code to discriminate between pairs of odors. Some of these predictions were supported in a behavioral discrimination (masking assay but others were not. We conclude that our model of the peripheral code represents basic features of odor detection and discrimination, yielding insights into the information available to higher processing structures in the brain.

Mining Gene Regulatory Networks by Neural Modeling of Expression Time-Series.

Science.gov (United States)

Rubiolo, Mariano; Milone, Diego H; Stegmayer, Georgina

2015-01-01

Discovering gene regulatory networks from data is one of the most studied topics in recent years. Neural networks can be successfully used to infer an underlying gene network by modeling expression profiles as times series. This work proposes a novel method based on a pool of neural networks for obtaining a gene regulatory network from a gene expression dataset. They are used for modeling each possible interaction between pairs of genes in the dataset, and a set of mining rules is applied to accurately detect the subjacent relations among genes. The results obtained on artificial and real datasets confirm the method effectiveness for discovering regulatory networks from a proper modeling of the temporal dynamics of gene expression profiles.

Data model description for the DESCARTES and CIDER codes

International Nuclear Information System (INIS)

Miley, T.B.; Ouderkirk, S.J.; Nichols, W.E.; Eslinger, P.W.

1993-01-01

The primary objective of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate the radiation dose that individuals could have received as a result of emissions since 1944 from the US Department of Energy's (DOE) Hanford Site near Richland, Washington. One of the major objectives of the HEDR Project is to develop several computer codes to model the airborne releases. transport and envirorunental accumulation of radionuclides resulting from Hanford operations from 1944 through 1972. In July 1992, the HEDR Project Manager determined that the computer codes being developed (DESCARTES, calculation of environmental accumulation from airborne releases, and CIDER, dose calculations from environmental accumulation) were not sufficient to create accurate models. A team of HEDR staff members developed a plan to assure that computer codes would meet HEDR Project goals. The plan consists of five tasks: (1) code requirements definition. (2) scoping studies, (3) design specifications, (4) benchmarking, and (5) data modeling. This report defines the data requirements for the DESCARTES and CIDER codes

Partitioning of genetic variation between regulatory and coding gene segments: the predominance of software variation in genes encoding introvert proteins.

Science.gov (United States)

Mitchison, A

1997-01-01

In considering genetic variation in eukaryotes, a fundamental distinction can be made between variation in regulatory (software) and coding (hardware) gene segments. For quantitative traits the bulk of variation, particularly that near the population mean, appears to reside in regulatory segments. The main exceptions to this rule concern proteins which handle extrinsic substances, here termed extrovert proteins. The immune system includes an unusually large proportion of this exceptional category, but even so its chief source of variation may well be polymorphism in regulatory gene segments. The main evidence for this view emerges from genome scanning for quantitative trait loci (QTL), which in the case of the immune system points to a major contribution of pro-inflammatory cytokine genes. Further support comes from sequencing of major histocompatibility complex (Mhc) class II promoters, where a high level of polymorphism has been detected. These Mhc promoters appear to act, in part at least, by gating the back-signal from T cells into antigen-presenting cells. Both these forms of polymorphism are likely to be sustained by the need for flexibility in the immune response. Future work on promoter polymorphism is likely to benefit from the input from genome informatics.

LMFBR models for the ORIGEN2 computer code

International Nuclear Information System (INIS)

Croff, A.G.; McAdoo, J.W.; Bjerke, M.A.

1981-10-01

Reactor physics calculations have led to the development of nine liquid-metal fast breeder reactor (LMFBR) models for the ORIGEN2 computer code. Four of the models are based on the U-Pu fuel cycle, two are based on the Th-U-Pu fuel cycle, and three are based on the Th- 238 U fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST are given

Verification of the LWRARC code for light-water-reactor afterheat rate calculations

International Nuclear Information System (INIS)

Murphy, B.D.

1998-02-01

This report describes verification studies carried out on the LWRARC (Light-Water-Reactor Afterheat Rate Calculations) computer code. The LWRARC code is proposed for automating the implementation of procedures specified in Draft Revision 1 of the U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 3.54, open-quotes Spent-Fuel Heat Generation in an Independent Spent-Fuel Storage Installation,close quotes which gives guidelines on the calculation of decay heat for spent nuclear fuel. Draft Regulatory Guide 3.54 allows one to estimate decay-heat values by means of a table lookup procedure with interpolation performed between table-entry values. The tabulated values of the relevant parameters span ranges that are appropriate for spent fuel from a boiling-water reactor (BWR) or a pressurized-water reactor (PWR), as the case may be, and decay-heat rates are obtained for spent fuel whose properties are within those parameter limits. In some instances, where these limits are either exceeded or where they approach critical regions, adjustments are invoked following table lookup. The LWRARC computer code is intended to replicate the manual process just described. In the code, the table lookup is done by entering a database and carrying out interpolations. The code then determines if adjustments apply, and, if this is the case, adjustment factors are calculated separately. The manual procedures in the Draft Regulatory Guide have been validated (i.e., they produce results that are good estimates of reality). The work reported in this document verifies that the LWRARC code replicates the manual procedures of the Draft Regulatory Guide, and that the code, taken together with the Draft Regulatory Guide, can support both verification and validation processes

Toward a Probabilistic Automata Model of Some Aspects of Code-Switching.

Science.gov (United States)

Dearholt, D. W.; Valdes-Fallis, G.

1978-01-01

The purpose of the model is to select either Spanish or English as the language to be used; its goals at this stage of development include modeling code-switching for lexical need, apparently random code-switching, dependency of code-switching upon sociolinguistic context, and code-switching within syntactic constraints. (EJS)

Repairing business process models as retrieved from source code

NARCIS (Netherlands)

Fernández-Ropero, M.; Reijers, H.A.; Pérez-Castillo, R.; Piattini, M.; Nurcan, S.; Proper, H.A.; Soffer, P.; Krogstie, J.; Schmidt, R.; Halpin, T.; Bider, I.

2013-01-01

The static analysis of source code has become a feasible solution to obtain underlying business process models from existing information systems. Due to the fact that not all information can be automatically derived from source code (e.g., consider manual activities), such business process models

Promoter Analysis Reveals Globally Differential Regulation of Human Long Non-Coding RNA and Protein-Coding Genes

KAUST Repository

Alam, Tanvir

2014-10-02

Transcriptional regulation of protein-coding genes is increasingly well-understood on a global scale, yet no comparable information exists for long non-coding RNA (lncRNA) genes, which were recently recognized to be as numerous as protein-coding genes in mammalian genomes. We performed a genome-wide comparative analysis of the promoters of human lncRNA and protein-coding genes, finding global differences in specific genetic and epigenetic features relevant to transcriptional regulation. These two groups of genes are hence subject to separate transcriptional regulatory programs, including distinct transcription factor (TF) proteins that significantly favor lncRNA, rather than coding-gene, promoters. We report a specific signature of promoter-proximal transcriptional regulation of lncRNA genes, including several distinct transcription factor binding sites (TFBS). Experimental DNase I hypersensitive site profiles are consistent with active configurations of these lncRNA TFBS sets in diverse human cell types. TFBS ChIP-seq datasets confirm the binding events that we predicted using computational approaches for a subset of factors. For several TFs known to be directly regulated by lncRNAs, we find that their putative TFBSs are enriched at lncRNA promoters, suggesting that the TFs and the lncRNAs may participate in a bidirectional feedback loop regulatory network. Accordingly, cells may be able to modulate lncRNA expression levels independently of mRNA levels via distinct regulatory pathways. Our results also raise the possibility that, given the historical reliance on protein-coding gene catalogs to define the chromatin states of active promoters, a revision of these chromatin signature profiles to incorporate expressed lncRNA genes is warranted in the future.

WDEC: A Code for Modeling White Dwarf Structure and Pulsations

Science.gov (United States)

Bischoff-Kim, Agnès; Montgomery, Michael H.

2018-05-01

The White Dwarf Evolution Code (WDEC), written in Fortran, makes models of white dwarf stars. It is fast, versatile, and includes the latest physics. The code evolves hot (∼100,000 K) input models down to a chosen effective temperature by relaxing the models to be solutions of the equations of stellar structure. The code can also be used to obtain g-mode oscillation modes for the models. WDEC has a long history going back to the late 1960s. Over the years, it has been updated and re-packaged for modern computer architectures and has specifically been used in computationally intensive asteroseismic fitting. Generations of white dwarf astronomers and dozens of publications have made use of the WDEC, although the last true instrument paper is the original one, published in 1975. This paper discusses the history of the code, necessary to understand why it works the way it does, details the physics and features in the code today, and points the reader to where to find the code and a user guide.

Model-Driven Engineering of Machine Executable Code

Science.gov (United States)

Eichberg, Michael; Monperrus, Martin; Kloppenburg, Sven; Mezini, Mira

Implementing static analyses of machine-level executable code is labor intensive and complex. We show how to leverage model-driven engineering to facilitate the design and implementation of programs doing static analyses. Further, we report on important lessons learned on the benefits and drawbacks while using the following technologies: using the Scala programming language as target of code generation, using XML-Schema to express a metamodel, and using XSLT to implement (a) transformations and (b) a lint like tool. Finally, we report on the use of Prolog for writing model transformations.

Characteristics of a third generation regulatory models

International Nuclear Information System (INIS)

Kallos, G.; Pilinis, C.; Kassomenos, P.; Hatzakis, G.

1992-01-01

A new class of air pollution dispersion models has to be used for regulatory purposes in the future and they should have the following capabilities: Adequate spatial and temporal characterization of the PBL, wind and turbulence fields, enhancement and regionalization of existing air quality models to yield a model applicable on both urban and regional scales, adequate characterization of the chemical transformation of the gaseous pollutants in the atmosphere, adequate description of the dispersion and the physical changes of the toxic material released (i.e. evaporation and condensation), good description of the chemistry in aqueous phases, like fog and cloud chemistry, and accurate predictions of the removal processes of the gas and particulate matter (i.e. wet an dry deposition). Some of these model requirements are already satisfied in some research oriented models. In general, there is no unique model that satisfies all of the above, and even if it existed there is not the necessary database for such application and testing. The two major areas of development of regulatory modeling in the future, i.e. development of a complete meteorological and air quality model and development of the database capabilities for the routine application of the model, in the urban and rural areas of interest. With the computer power available in the near future, this kind of models will be able to be used even for emergency cases. (AB) (20 refs.)

Preliminary Consideration for the Development of Regulatory Level 2 PSA Model

International Nuclear Information System (INIS)

Lee, Chang-Ju

2006-01-01

In order to assess the validity of PSA (probabilistic safety assessment) results and to establish regulatory requirements for relevant safety issues most of the regulators want to develop an independent and convenient risk assessment model including Level 2 PSA area. As this model and framework should be implicitly independent on the licensee's PSA model, it has a primary objective directly for applying to the risk-informed regulatory affairs and for supporting those kinds of works. According this, the regulator can take an objective view for the uncertainty of risk information made by the licensee and keep up the capability and decision-making framework for overall risk assessment results. In addition, the regulatory model may be used to verify and validate the operational risk levels of all engineered safety features of nuclear power plants (NPPs). An issue for plant-specific application of safety goals was previously identified in the US NRC's risk-informed regulatory guidance development activities, and discussed in many Commission papers, e. g. SECY-97-287, which identifies the goal for large early release frequency (LERF). LERF defines a containment performance criteria derived from the quantitative health objectives. As we know, the LERF was chosen to assess risk significance in Regulatory Guide 1.174 (2002) again, which provides one measure of the performance of the containment barrier, and represents a surrogate for early health effects

Code manual for MACCS2: Volume 1, user`s guide

Energy Technology Data Exchange (ETDEWEB)

Chanin, D.I.; Young, M.L.

1997-03-01

This report describes the use of the MACCS2 code. The document is primarily a user`s guide, though some model description information is included. MACCS2 represents a major enhancement of its predecessor MACCS, the MELCOR Accident Consequence Code System. MACCS, distributed by government code centers since 1990, was developed to evaluate the impacts of severe accidents at nuclear power plants on the surrounding public. The principal phenomena considered are atmospheric transport and deposition under time-variant meteorology, short- and long-term mitigative actions and exposure pathways, deterministic and stochastic health effects, and economic costs. No other U.S. code that is publicly available at present offers all these capabilities. MACCS2 was developed as a general-purpose tool applicable to diverse reactor and nonreactor facilities licensed by the Nuclear Regulatory Commission or operated by the Department of Energy or the Department of Defense. The MACCS2 package includes three primary enhancements: (1) a more flexible emergency-response model, (2) an expanded library of radionuclides, and (3) a semidynamic food-chain model. Other improvements are in the areas of phenomenological modeling and new output options. Initial installation of the code, written in FORTRAN 77, requires a 486 or higher IBM-compatible PC with 8 MB of RAM.

The Role of the International Code Council in the U.S. Building Regukation System and Green Building Contruction

OpenAIRE

David Walls

2015-01-01

This paper will address the components of the International Code Council (ICC) as one of the most important organizations in terms of developing the model building codes for the US: the international codes. This membership-driven organization has the task of providing the building industry and all its stakeholders with the necessary regulatory documents, training, certification, plan check, product evaluation, and accreditation services to achieve safer and more sustainable building construct...

Preliminary Development of Regulatory PSA Models for SFR

International Nuclear Information System (INIS)

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

2013-01-01

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

Reactor safety computer code development at INEL

International Nuclear Information System (INIS)

Johnsen, G.W.

1985-01-01

This report provides a brief overview of the computer code development programs being conducted at EG and G Idaho, Inc. on behalf of US Nuclear Regulatory Commission and the Department of Energy, Idaho Operations Office. Included are descriptions of the codes being developed, their development status as of the date of this report, and resident code development expertise

Code Development and Analysis Program: developmental checkout of the BEACON/MOD2A code

International Nuclear Information System (INIS)

Ramsthaler, J.A.; Lime, J.F.; Sahota, M.S.

1978-12-01

A best-estimate transient containment code, BEACON, is being developed by EG and G Idaho, Inc. for the Nuclear Regulatory Commission's reactor safety research program. This is an advanced, two-dimensional fluid flow code designed to predict temperatures and pressures in a dry PWR containment during a hypothetical loss-of-coolant accident. The most recent version of the code, MOD2A, is presently in the final stages of production prior to being released to the National Energy Software Center. As part of the final code checkout, seven sample problems were selected to be run with BEACON/MOD2A

Developmental assessment of the SCDAP/RELAP5 code

International Nuclear Information System (INIS)

Harvego, E.A.; Slefken, L.J.; Coryell, E.W.

1997-01-01

The development and assessment of the late-phase damage progression models in the current version (designated MOD3.2) of the SCDAP/RELAP5 code are described. The SCDAP/RELAP5 code is being developed at the Idaho National Engineering and Environmental Laboratory under the primary sponsorship of the US Nuclear Regulatory Commission (NRC) to provide best-estimate transient simulations of light water reactor coolant systems (RCS) during severe accident conditions. Recent modeling improvements made to the MOD3.2 version of the code include (1) molten pool formation and heat up, including the transient start-up of natural circulation heat transfer, (2) in-core molten pool thermal-mechanical crust failure, (3) the melting and relocation of upper plenum structures, and (4) improvements in the modeling of lower plenum debris behavior and the potential for failure of the lower head. Finally, to eliminate abrupt transitions between core damage states and provide more realistic predictions of late phase accident progression phenomena, a transition smoothing methodology was developed and implemented that results in the calculation of a gradual transition from an intact core geometry through the different core damage states leading to molten pool formation. A wide range of experiments and modeling tools were used to assess the capabilities of MOD3.2. The results of the SCDAP/RELAP5/MOD3.2 assessment indicate that modeling improvements have significantly enhanced the code capabilities and performance in several areas compared to the earlier code version. New models for transition smoothing between core damage states, and modeling improvements/additions for cladding oxide failure, molten pool behavior, and molten pool crust failure have significantly improved the code usability for a wide range of applications and have significantly improved the prediction of hydrogen production, molten pool melt mass and core melt relocation time

Pervasive hitchhiking at coding and regulatory sites in humans.

Directory of Open Access Journals (Sweden)

James J Cai

2009-01-01

Full Text Available Much effort and interest have focused on assessing the importance of natural selection, particularly positive natural selection, in shaping the human genome. Although scans for positive selection have identified candidate loci that may be associated with positive selection in humans, such scans do not indicate whether adaptation is frequent in general in humans. Studies based on the reasoning of the MacDonald-Kreitman test, which, in principle, can be used to evaluate the extent of positive selection, suggested that adaptation is detectable in the human genome but that it is less common than in Drosophila or Escherichia coli. Both positive and purifying natural selection at functional sites should affect levels and patterns of polymorphism at linked nonfunctional sites. Here, we search for these effects by analyzing patterns of neutral polymorphism in humans in relation to the rates of recombination, functional density, and functional divergence with chimpanzees. We find that the levels of neutral polymorphism are lower in the regions of lower recombination and in the regions of higher functional density or divergence. These correlations persist after controlling for the variation in GC content, density of simple repeats, selective constraint, mutation rate, and depth of sequencing coverage. We argue that these results are most plausibly explained by the effects of natural selection at functional sites -- either recurrent selective sweeps or background selection -- on the levels of linked neutral polymorphism. Natural selection at both coding and regulatory sites appears to affect linked neutral polymorphism, reducing neutral polymorphism by 6% genome-wide and by 11% in the gene-rich half of the human genome. These findings suggest that the effects of natural selection at linked sites cannot be ignored in the study of neutral human polymorphism.

Cross-band noise model refinement for transform domain Wyner–Ziv video coding

DEFF Research Database (Denmark)

Huang, Xin; Forchhammer, Søren

2012-01-01

TDWZ video coding trails that of conventional video coding solutions, mainly due to the quality of side information, inaccurate noise modeling and loss in the final coding step. The major goal of this paper is to enhance the accuracy of the noise modeling, which is one of the most important aspects...... influencing the coding performance of DVC. A TDWZ video decoder with a novel cross-band based adaptive noise model is proposed, and a noise residue refinement scheme is introduced to successively update the estimated noise residue for noise modeling after each bit-plane. Experimental results show...... that the proposed noise model and noise residue refinement scheme can improve the rate-distortion (RD) performance of TDWZ video coding significantly. The quality of the side information modeling is also evaluated by a measure of the ideal code length....

Development of the Level 1 PSA Model for PGSFR Regulatory

International Nuclear Information System (INIS)

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

2014-01-01

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

MELCOR computer code manuals

Energy Technology Data Exchange (ETDEWEB)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L. [Sandia National Labs., Albuquerque, NM (United States); Hodge, S.A.; Hyman, C.R.; Sanders, R.L. [Oak Ridge National Lab., TN (United States)

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.

MELCOR computer code manuals

International Nuclear Information System (INIS)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L.; Hodge, S.A.; Hyman, C.R.; Sanders, R.L.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR's phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package

CONTAIN 2.0 code release and the transition to licensing

International Nuclear Information System (INIS)

Murata, K.K.; Griffith, R.O.; Bergeron, K.D.; Tills, J.

1997-10-01

CONTAIN is a reactor accident simulation code developed by Sandia National Laboratories under US Nuclear Regulatory Commission (USNRC) sponsorship to provide integrated analysis of containment phenomena, including those related to nuclear reactor containment loads and radiological source terms. The recently released CONTAIN 2.0 code version represents a significant advance in CONTAIN modeling capabilities over the last major code release (CONTAIN 1.12V). The new modeling capabilities are discussed here. The principal motivation for many of the recent model improvements has been to allow CONTAIN to model the special features in advanced light water reactor (ALWR) designs. The work done in this area is also summarized. In addition to the ALWR work, the USNRC is currently engaged in an effort to qualify CONTAIN for more general use in licensing, with the intent of supplementing or possibly replacing traditional licensing codes. To qualify the CONTAIN code for licensing applications, studies utilizing CONTAIN 2.0 are in progress. A number of results from this effort are presented in this paper to illustrate the code capabilities. In particular, CONTAIN calculations of the NUPEC M-8-1 and ISP-23 experiments and CVTR test number-sign 3 are presented to illustrate (1) the ability of CONTAIN to model non-uniform gas density and/or temperature distributions, and (2) the relationship between such gas distributions and containment loads. CONTAIN and CONTEMPT predictions for a large break loss of coolant accident scenario in the San Onofre plant are also compared

A system-level model for the microbial regulatory genome.

Science.gov (United States)

Brooks, Aaron N; Reiss, David J; Allard, Antoine; Wu, Wei-Ju; Salvanha, Diego M; Plaisier, Christopher L; Chandrasekaran, Sriram; Pan, Min; Kaur, Amardeep; Baliga, Nitin S

2014-07-15

Microbes can tailor transcriptional responses to diverse environmental challenges despite having streamlined genomes and a limited number of regulators. Here, we present data-driven models that capture the dynamic interplay of the environment and genome-encoded regulatory programs of two types of prokaryotes: Escherichia coli (a bacterium) and Halobacterium salinarum (an archaeon). The models reveal how the genome-wide distributions of cis-acting gene regulatory elements and the conditional influences of transcription factors at each of those elements encode programs for eliciting a wide array of environment-specific responses. We demonstrate how these programs partition transcriptional regulation of genes within regulons and operons to re-organize gene-gene functional associations in each environment. The models capture fitness-relevant co-regulation by different transcriptional control mechanisms acting across the entire genome, to define a generalized, system-level organizing principle for prokaryotic gene regulatory networks that goes well beyond existing paradigms of gene regulation. An online resource (http://egrin2.systemsbiology.net) has been developed to facilitate multiscale exploration of conditional gene regulation in the two prokaryotes. © 2014 The Authors. Published under the terms of the CC BY 4.0 license.

Coupling a Basin Modeling and a Seismic Code using MOAB

KAUST Repository

Yan, Mi; Jordan, Kirk; Kaushik, Dinesh; Perrone, Michael; Sachdeva, Vipin; Tautges, Timothy J.; Magerlein, John

2012-01-01

We report on a demonstration of loose multiphysics coupling between a basin modeling code and a seismic code running on a large parallel machine. Multiphysics coupling, which is one critical capability for a high performance computing (HPC) framework, was implemented using the MOAB open-source mesh and field database. MOAB provides for code coupling by storing mesh data and input and output field data for the coupled analysis codes and interpolating the field values between different meshes used by the coupled codes. We found it straightforward to use MOAB to couple the PBSM basin modeling code and the FWI3D seismic code on an IBM Blue Gene/P system. We describe how the coupling was implemented and present benchmarking results for up to 8 racks of Blue Gene/P with 8192 nodes and MPI processes. The coupling code is fast compared to the analysis codes and it scales well up to at least 8192 nodes, indicating that a mesh and field database is an efficient way to implement loose multiphysics coupling for large parallel machines.

Coupling a Basin Modeling and a Seismic Code using MOAB

KAUST Repository

Yan, Mi

2012-06-02

We report on a demonstration of loose multiphysics coupling between a basin modeling code and a seismic code running on a large parallel machine. Multiphysics coupling, which is one critical capability for a high performance computing (HPC) framework, was implemented using the MOAB open-source mesh and field database. MOAB provides for code coupling by storing mesh data and input and output field data for the coupled analysis codes and interpolating the field values between different meshes used by the coupled codes. We found it straightforward to use MOAB to couple the PBSM basin modeling code and the FWI3D seismic code on an IBM Blue Gene/P system. We describe how the coupling was implemented and present benchmarking results for up to 8 racks of Blue Gene/P with 8192 nodes and MPI processes. The coupling code is fast compared to the analysis codes and it scales well up to at least 8192 nodes, indicating that a mesh and field database is an efficient way to implement loose multiphysics coupling for large parallel machines.

Cis-regulatory somatic mutations and gene-expression alteration in B-cell lymphomas.

Science.gov (United States)

Mathelier, Anthony; Lefebvre, Calvin; Zhang, Allen W; Arenillas, David J; Ding, Jiarui; Wasserman, Wyeth W; Shah, Sohrab P

2015-04-23

With the rapid increase of whole-genome sequencing of human cancers, an important opportunity to analyze and characterize somatic mutations lying within cis-regulatory regions has emerged. A focus on protein-coding regions to identify nonsense or missense mutations disruptive to protein structure and/or function has led to important insights; however, the impact on gene expression of mutations lying within cis-regulatory regions remains under-explored. We analyzed somatic mutations from 84 matched tumor-normal whole genomes from B-cell lymphomas with accompanying gene expression measurements to elucidate the extent to which these cancers are disrupted by cis-regulatory mutations. We characterize mutations overlapping a high quality set of well-annotated transcription factor binding sites (TFBSs), covering a similar portion of the genome as protein-coding exons. Our results indicate that cis-regulatory mutations overlapping predicted TFBSs are enriched in promoter regions of genes involved in apoptosis or growth/proliferation. By integrating gene expression data with mutation data, our computational approach culminates with identification of cis-regulatory mutations most likely to participate in dysregulation of the gene expression program. The impact can be measured along with protein-coding mutations to highlight key mutations disrupting gene expression and pathways in cancer. Our study yields specific genes with disrupted expression triggered by genomic mutations in either the coding or the regulatory space. It implies that mutated regulatory components of the genome contribute substantially to cancer pathways. Our analyses demonstrate that identifying genomically altered cis-regulatory elements coupled with analysis of gene expression data will augment biological interpretation of mutational landscapes of cancers.

Development of regulatory techniques for operational performance evaluation of nuclear power plants

International Nuclear Information System (INIS)

Sung, K. Y.; Lee, C. J.

2005-03-01

For the purpose of keeping up the capability for independent regulatory audit of risk information, we developed an PSA model, independent on that of the licensee. The developed regulatory PSA model, as named MPAS (Multi-purpose Probabilistic Analysis of Safety), is primarily based on the result of feasibility study about ASME PSA standard and done by the cooperation work with KAERI PSA research team. Current MPAS model is limited to the work scope on Level 1 internal full power analysis, which is finally implemented for the KSNP. The development of MPAS model consists of plenty of works establishing overall items which comprehensively implements all subjects such as detailed logic and input data for each analysis fields. In addition, relevant computational framework utilizing PSA and risk monitoring codes is also developed for users' convenience

RELAP5/MOD3 code manual. Volume 4, Models and correlations

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 presents modeling theory and associated numerical schemes; Volume II details instructions for code application and input data preparation; Volume III presents the results of developmental assessment cases that demonstrate and verify the models used in the code; Volume IV discusses in detail RELAP5 models and correlations; Volume V presents guidelines that have evolved over the past several years through the use of the RELAP5 code; Volume VI discusses the numerical scheme used in RELAP5; and Volume VII presents a collection of independent assessment calculations

A vectorized Monte Carlo code for modeling photon transport in SPECT

International Nuclear Information System (INIS)

Smith, M.F.; Floyd, C.E. Jr.; Jaszczak, R.J.

1993-01-01

A vectorized Monte Carlo computer code has been developed for modeling photon transport in single photon emission computed tomography (SPECT). The code models photon transport in a uniform attenuating region and photon detection by a gamma camera. It is adapted from a history-based Monte Carlo code in which photon history data are stored in scalar variables and photon histories are computed sequentially. The vectorized code is written in FORTRAN77 and uses an event-based algorithm in which photon history data are stored in arrays and photon history computations are performed within DO loops. The indices of the DO loops range over the number of photon histories, and these loops may take advantage of the vector processing unit of our Stellar GS1000 computer for pipelined computations. Without the use of the vector processor the event-based code is faster than the history-based code because of numerical optimization performed during conversion to the event-based algorithm. When only the detection of unscattered photons is modeled, the event-based code executes 5.1 times faster with the use of the vector processor than without; when the detection of scattered and unscattered photons is modeled the speed increase is a factor of 2.9. Vectorization is a valuable way to increase the performance of Monte Carlo code for modeling photon transport in SPECT

The analysis of thermal-hydraulic models in MELCOR code

Energy Technology Data Exchange (ETDEWEB)

Kim, M H; Hur, C; Kim, D K; Cho, H J [POhang Univ., of Science and TECHnology, Pohang (Korea, Republic of)

1996-07-15

The objective of the present work is to verify the prediction and analysis capability of MELCOR code about the progression of severe accidents in light water reactor and also to evaluate appropriateness of thermal-hydraulic models used in MELCOR code. Comparing the results of experiment and calculation with MELCOR code is carried out to achieve the above objective. Specially, the comparison between the CORA-13 experiment and the MELCOR code calculation was performed.

Modeling Guidelines for Code Generation in the Railway Signaling Context

Science.gov (United States)

Ferrari, Alessio; Bacherini, Stefano; Fantechi, Alessandro; Zingoni, Niccolo

2009-01-01

Modeling guidelines constitute one of the fundamental cornerstones for Model Based Development. Their relevance is essential when dealing with code generation in the safety-critical domain. This article presents the experience of a railway signaling systems manufacturer on this issue. Introduction of Model-Based Development (MBD) and code generation in the industrial safety-critical sector created a crucial paradigm shift in the development process of dependable systems. While traditional software development focuses on the code, with MBD practices the focus shifts to model abstractions. The change has fundamental implications for safety-critical systems, which still need to guarantee a high degree of confidence also at code level. Usage of the Simulink/Stateflow platform for modeling, which is a de facto standard in control software development, does not ensure by itself production of high-quality dependable code. This issue has been addressed by companies through the definition of modeling rules imposing restrictions on the usage of design tools components, in order to enable production of qualified code. The MAAB Control Algorithm Modeling Guidelines (MathWorks Automotive Advisory Board)[3] is a well established set of publicly available rules for modeling with Simulink/Stateflow. This set of recommendations has been developed by a group of OEMs and suppliers of the automotive sector with the objective of enforcing and easing the usage of the MathWorks tools within the automotive industry. The guidelines have been published in 2001 and afterwords revisited in 2007 in order to integrate some additional rules developed by the Japanese division of MAAB [5]. The scope of the current edition of the guidelines ranges from model maintainability and readability to code generation issues. The rules are conceived as a reference baseline and therefore they need to be tailored to comply with the characteristics of each industrial context. Customization of these

Modelling of fluid-solid interaction using two stand-alone codes

CSIR Research Space (South Africa)

Grobler, Jan H

2010-01-01

Full Text Available A method is proposed for the modelling of fluid-solid interaction in applications where fluid forces dominate. Data are transferred between two stand-alone codes: a dedicated computational fluid dynamics (CFD) code capable of free surface modelling...

Survey of computer codes applicable to waste facility performance evaluations

International Nuclear Information System (INIS)

Alsharif, M.; Pung, D.L.; Rivera, A.L.; Dole, L.R.

1988-01-01

This study is an effort to review existing information that is useful to develop an integrated model for predicting the performance of a radioactive waste facility. A summary description of 162 computer codes is given. The identified computer programs address the performance of waste packages, waste transport and equilibrium geochemistry, hydrological processes in unsaturated and saturated zones, and general waste facility performance assessment. Some programs also deal with thermal analysis, structural analysis, and special purposes. A number of these computer programs are being used by the US Department of Energy, the US Nuclear Regulatory Commission, and their contractors to analyze various aspects of waste package performance. Fifty-five of these codes were identified as being potentially useful on the analysis of low-level radioactive waste facilities located above the water table. The code summaries include authors, identification data, model types, and pertinent references. 14 refs., 5 tabs

Universal Regularizers For Robust Sparse Coding and Modeling

OpenAIRE

Ramirez, Ignacio; Sapiro, Guillermo

2010-01-01

Sparse data models, where data is assumed to be well represented as a linear combination of a few elements from a dictionary, have gained considerable attention in recent years, and their use has led to state-of-the-art results in many signal and image processing tasks. It is now well understood that the choice of the sparsity regularization term is critical in the success of such models. Based on a codelength minimization interpretation of sparse coding, and using tools from universal coding...

Phenomenological optical potentials and optical model computer codes

International Nuclear Information System (INIS)

Prince, A.

1980-01-01

An introduction to the Optical Model is presented. Starting with the purpose and nature of the physical problems to be analyzed, a general formulation and the various phenomenological methods of solution are discussed. This includes the calculation of observables based on assumed potentials such as local and non-local and their forms, e.g. Woods-Saxon, folded model etc. Also discussed are the various calculational methods and model codes employed to describe nuclear reactions in the spherical and deformed regions (e.g. coupled-channel analysis). An examination of the numerical solutions and minimization techniques associated with the various codes, is briefly touched upon. Several computer programs are described for carrying out the calculations. The preparation of input, (formats and options), determination of model parameters and analysis of output are described. The class is given a series of problems to carry out using the available computer. Interpretation and evaluation of the samples includes the effect of varying parameters, and comparison of calculations with the experimental data. Also included is an intercomparison of the results from the various model codes, along with their advantages and limitations. (author)

Anthropomorphic Coding of Speech and Audio: A Model Inversion Approach

Directory of Open Access Journals (Sweden)

W. Bastiaan Kleijn

2005-06-01

Full Text Available Auditory modeling is a well-established methodology that provides insight into human perception and that facilitates the extraction of signal features that are most relevant to the listener. The aim of this paper is to provide a tutorial on perceptual speech and audio coding using an invertible auditory model. In this approach, the audio signal is converted into an auditory representation using an invertible auditory model. The auditory representation is quantized and coded. Upon decoding, it is then transformed back into the acoustic domain. This transformation converts a complex distortion criterion into a simple one, thus facilitating quantization with low complexity. We briefly review past work on auditory models and describe in more detail the components of our invertible model and its inversion procedure, that is, the method to reconstruct the signal from the output of the auditory model. We summarize attempts to use the auditory representation for low-bit-rate coding. Our approach also allows the exploitation of the inherent redundancy of the human auditory system for the purpose of multiple description (joint source-channel coding.

Dual coding: a cognitive model for psychoanalytic research.

Science.gov (United States)

Bucci, W

1985-01-01

Four theories of mental representation derived from current experimental work in cognitive psychology have been discussed in relation to psychoanalytic theory. These are: verbal mediation theory, in which language determines or mediates thought; perceptual dominance theory, in which imagistic structures are dominant; common code or propositional models, in which all information, perceptual or linguistic, is represented in an abstract, amodal code; and dual coding, in which nonverbal and verbal information are each encoded, in symbolic form, in separate systems specialized for such representation, and connected by a complex system of referential relations. The weight of current empirical evidence supports the dual code theory. However, psychoanalysis has implicitly accepted a mixed model-perceptual dominance theory applying to unconscious representation, and verbal mediation characterizing mature conscious waking thought. The characterization of psychoanalysis, by Schafer, Spence, and others, as a domain in which reality is constructed rather than discovered, reflects the application of this incomplete mixed model. The representations of experience in the patient's mind are seen as without structure of their own, needing to be organized by words, thus vulnerable to distortion or dissolution by the language of the analyst or the patient himself. In these terms, hypothesis testing becomes a meaningless pursuit; the propositions of the theory are no longer falsifiable; the analyst is always more or less "right." This paper suggests that the integrated dual code formulation provides a more coherent theoretical framework for psychoanalysis than the mixed model, with important implications for theory and technique. In terms of dual coding, the problem is not that the nonverbal representations are vulnerable to distortion by words, but that the words that pass back and forth between analyst and patient will not affect the nonverbal schemata at all. Using the dual code

Ecological models for regulatory risk assessments of pesticides: Developing a strategy for the future.

NARCIS (Netherlands)

Thorbek, P.; Forbes, V.; Heimbach, F.; Hommen, U.; Thulke, H.H.; Brink, van den P.J.

2010-01-01

Ecological Models for Regulatory Risk Assessments of Pesticides: Developing a Strategy for the Future provides a coherent, science-based view on ecological modeling for regulatory risk assessments. It discusses the benefits of modeling in the context of registrations, identifies the obstacles that

Constitutive model development needs for reactor safety thermal-hydraulic codes

International Nuclear Information System (INIS)

Kelly, J.M.

1998-01-01

This paper discusses the constitutive model development needs for our current and future generation of reactor safety thermal-hydraulic analysis codes. Rather than provide a simple 'shopping list' of models to be improved, a detailed description is given of how a constitutive model works within the computational framework of a current reactor safety code employing the two-fluid model of two-phase flow. The intent is to promote a better understanding of both the types of experiments and the instrumentation needs that will be required in the USNRCs code improvement program. First, a summary is given of the modeling considerations that need to be taken into account when developing constitutive models for use in reactor safety thermal-hydraulic codes. Specifically, the two-phase flow model should be applicable to a control volume formulation employing computational volumes with dimensions on the order of meters but containing embedded structure with a dimension on the order of a centimeter. The closure relations are then required to be suitable when averaged over such large volumes containing millions or even tens of millions of discrete fluid particles (bubbles/drops). This implies a space and time averaging procedure that neglects the intermittency observed in slug and chum turbulent two-phase flows. Furthermore, the geometries encountered in reactor systems are complex, the constitutive relations should therefore be component specific (e.g., interfacial shear in a tube does not represent that in a rod bundle nor in the downcomer). When practicable, future modeling efforts should be directed towards resolving the spatial evolution of two-phase flow patterns through the introduction of interfacial area transport equations and by modeling the individual physical processes responsible for the creation or destruction of interfacial area. Then the example of the implementation and assessment of a subcooled boiling model in a two-fluid code is given. The primary parameter

An object-oriented framework for magnetic-fusion modeling and analysis codes

International Nuclear Information System (INIS)

Cohen, R H; Yang, T Y Brian.

1999-01-01

The magnetic-fusion energy (MFE) program, like many other scientific and engineering activities, has a need to efficiently develop complex modeling codes which combine detailed models of components to make an integrated model of a device, as well as a rich supply of legacy code that could provide the component models. There is also growing recognition in many technical fields of the desirability of steerable software: computer programs whose functionality can be changed by the user as it is run. This project had as its goals the development of two key pieces of infrastructure that are needed to combine existing code modules, written mainly in Fortran, into flexible, steerable, object-oriented integrated modeling codes for magnetic- fusion applications. These two pieces are (1) a set of tools to facilitate the interfacing of Fortran code with a steerable object-oriented framework (which we have chosen to be based on PythonlW3, an object-oriented interpreted language), and (2) a skeleton for the integrated modeling code which defines the relationships between the modules. The first of these activities obviously has immediate applicability to a spectrum of projects; the second is more focussed on the MFE application, but may be of value as an example for other applications

A MODEL BUILDING CODE ARTICLE ON FALLOUT SHELTERS WITH RECOMMENDATIONS FOR INCLUSION OF REQUIREMENTS FOR FALLOUT SHELTER CONSTRUCTION IN FOUR NATIONAL MODEL BUILDING CODES.

Science.gov (United States)

American Inst. of Architects, Washington, DC.

A MODEL BUILDING CODE FOR FALLOUT SHELTERS WAS DRAWN UP FOR INCLUSION IN FOUR NATIONAL MODEL BUILDING CODES. DISCUSSION IS GIVEN OF FALLOUT SHELTERS WITH RESPECT TO--(1) NUCLEAR RADIATION, (2) NATIONAL POLICIES, AND (3) COMMUNITY PLANNING. FALLOUT SHELTER REQUIREMENTS FOR SHIELDING, SPACE, VENTILATION, CONSTRUCTION, AND SERVICES SUCH AS ELECTRICAL…

Computational challenges in modeling gene regulatory events.

Science.gov (United States)

Pataskar, Abhijeet; Tiwari, Vijay K

2016-10-19

Cellular transcriptional programs driven by genetic and epigenetic mechanisms could be better understood by integrating "omics" data and subsequently modeling the gene-regulatory events. Toward this end, computational biology should keep pace with evolving experimental procedures and data availability. This article gives an exemplified account of the current computational challenges in molecular biology.

MIG version 0.0 model interface guidelines: Rules to accelerate installation of numerical models into any compliant parent code

Energy Technology Data Exchange (ETDEWEB)

Brannon, R.M.; Wong, M.K.

1996-08-01

A set of model interface guidelines, called MIG, is presented as a means by which any compliant numerical material model can be rapidly installed into any parent code without having to modify the model subroutines. Here, {open_quotes}model{close_quotes} usually means a material model such as one that computes stress as a function of strain, though the term may be extended to any numerical operation. {open_quotes}Parent code{close_quotes} means a hydrocode, finite element code, etc. which uses the model and enforces, say, the fundamental laws of motion and thermodynamics. MIG requires the model developer (who creates the model package) to specify model needs in a standardized but flexible way. MIG includes a dictionary of technical terms that allows developers and parent code architects to share a common vocabulary when specifying field variables. For portability, database management is the responsibility of the parent code. Input/output occurs via structured calling arguments. As much model information as possible (such as the lists of required inputs, as well as lists of precharacterized material data and special needs) is supplied by the model developer in an ASCII text file. Every MIG-compliant model also has three required subroutines to check data, to request extra field variables, and to perform model physics. To date, the MIG scheme has proven flexible in beta installations of a simple yield model, plus a more complicated viscodamage yield model, three electromechanical models, and a complicated anisotropic microcrack constitutive model. The MIG yield model has been successfully installed using identical subroutines in three vectorized parent codes and one parallel C++ code, all predicting comparable results. By maintaining one model for many codes, MIG facilitates code-to-code comparisons and reduces duplication of effort, thereby reducing the cost of installing and sharing models in diverse new codes.

The RADionuclide Transport, Removal, and Dose (RADTRAD) code

International Nuclear Information System (INIS)

Miller, L.A.; Chanin, D.I.; Lee, J.

1993-01-01

The RADionuclide Transport, Removal, And Dose (RADTRAD) code is designed for US Nuclear Regulatory Commission (USNRC) use to calculate the radiological consequences to the offsite population and to control room operators following a design-basis accident at Light Water Reactor (LWR) power plants. This code utilizes updated reactor accident source terms published in draft NUREG-1465, ''Accident Source Terms for Light-Water Nuclear Power Plants.'' The code will track the transport of radionuclides as they are released from the reactor pressure vessel, travel through the primary containment and other buildings, and are released to the environment. As the radioactive material is transported through the primary containment and other buildings, credit for several removal mechanisms may be taken including sprays, suppression pools, overlying pools, filters, and natural deposition. Simple models are available for these different removal mechanisms that use, as input, information about the conditions in the plant and predict either a removal coefficient (λ) or decontamination factor. The user may elect to use these models or input a single value for a removal coefficient or decontamination factor

Regulatory Safety Issues in the Structural Design Criteria of ASME Section III Subsection NH and for Very High Temperatures for VHTR & GEN IV

Energy Technology Data Exchange (ETDEWEB)

William J. O’Donnell; Donald S. Griffin

2007-05-07

The objective of this task is to identify issues relevant to ASME Section III, Subsection NH [1], and related Code Cases that must be resolved for licensing purposes for VHTGRs (Very High Temperature Gas Reactor concepts such as those of PBMR, Areva, and GA); and to identify the material models, design criteria, and analysis methods that need to be added to the ASME Code to cover the unresolved safety issues. Subsection NH was originally developed to provide structural design criteria and limits for elevated-temperature design of Liquid Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Breeder Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTGRs, and identified issues that need to be resolved to provide a regulatory basis for licensing. This Report describes: (1) NRC and ACRS safety concerns raised during the licensing process of CRBR , (2) how some of these issues are addressed by the current Subsection NH of the ASME Code; and (3) the material models, design criteria, and analysis methods that need to be added to the ASME Code and Code Cases to cover unresolved regulatory issues for very high temperature service.

A Perceptual Model for Sinusoidal Audio Coding Based on Spectral Integration

NARCIS (Netherlands)

Van de Par, S.; Kohlrausch, A.; Heusdens, R.; Jensen, J.; Holdt Jensen, S.

2005-01-01

Psychoacoustical models have been used extensively within audio coding applications over the past decades. Recently, parametric coding techniques have been applied to general audio and this has created the need for a psychoacoustical model that is specifically suited for sinusoidal modelling of

A perceptual model for sinusoidal audio coding based on spectral integration

NARCIS (Netherlands)

Van de Par, S.; Kohlrauch, A.; Heusdens, R.; Jensen, J.; Jensen, S.H.

2005-01-01

Psychoacoustical models have been used extensively within audio coding applications over the past decades. Recently, parametric coding techniques have been applied to general audio and this has created the need for a psychoacoustical model that is specifically suited for sinusoidal modelling of

Case studies in Gaussian process modelling of computer codes

International Nuclear Information System (INIS)

Kennedy, Marc C.; Anderson, Clive W.; Conti, Stefano; O'Hagan, Anthony

2006-01-01

In this paper we present a number of recent applications in which an emulator of a computer code is created using a Gaussian process model. Tools are then applied to the emulator to perform sensitivity analysis and uncertainty analysis. Sensitivity analysis is used both as an aid to model improvement and as a guide to how much the output uncertainty might be reduced by learning about specific inputs. Uncertainty analysis allows us to reflect output uncertainty due to unknown input parameters, when the finished code is used for prediction. The computer codes themselves are currently being developed within the UK Centre for Terrestrial Carbon Dynamics

Selection on Coding and Regulatory Variation Maintains Individuality in Major Urinary Protein Scent Marks in Wild Mice.

Directory of Open Access Journals (Sweden)

Michael J Sheehan

2016-03-01

Full Text Available Recognition of individuals by scent is widespread across animal taxa. Though animals can often discriminate chemical blends based on many compounds, recent work shows that specific protein pheromones are necessary and sufficient for individual recognition via scent marks in mice. The genetic nature of individuality in scent marks (e.g. coding versus regulatory variation and the evolutionary processes that maintain diversity are poorly understood. The individual signatures in scent marks of house mice are the protein products of a group of highly similar paralogs in the major urinary protein (Mup gene family. Using the offspring of wild-caught mice, we examine individuality in the major urinary protein (MUP scent marks at the DNA, RNA and protein levels. We show that individuality arises through a combination of variation at amino acid coding sites and differential transcription of central Mup genes across individuals, and we identify eSNPs in promoters. There is no evidence of post-transcriptional processes influencing phenotypic diversity as transcripts accurately predict the relative abundance of proteins in urine samples. The match between transcripts and urine samples taken six months earlier also emphasizes that the proportional relationships across central MUP isoforms in urine is stable. Balancing selection maintains coding variants at moderate frequencies, though pheromone diversity appears limited by interactions with vomeronasal receptors. We find that differential transcription of the central Mup paralogs within and between individuals significantly increases the individuality of pheromone blends. Balancing selection on gene regulation allows for increased individuality via combinatorial diversity in a limited number of pheromones.

21 CFR 201.25 - Bar code label requirements.

Science.gov (United States)

2010-04-01

... Number/Uniform Code Council (EAN.UCC) or Health Industry Business Communications Council (HIBCC... alternative regulatory program or method of product use renders the bar code unnecessary for patient safety... Evaluation and Research, Food and Drug Administration, 5600 Fishers Lane, Rockville, MD 20857 (requests...

A New Algorithm for Identifying Cis-Regulatory Modules Based on Hidden Markov Model

Directory of Open Access Journals (Sweden)

Haitao Guo

2017-01-01

Full Text Available The discovery of cis-regulatory modules (CRMs is the key to understanding mechanisms of transcription regulation. Since CRMs have specific regulatory structures that are the basis for the regulation of gene expression, how to model the regulatory structure of CRMs has a considerable impact on the performance of CRM identification. The paper proposes a CRM discovery algorithm called ComSPS. ComSPS builds a regulatory structure model of CRMs based on HMM by exploring the rules of CRM transcriptional grammar that governs the internal motif site arrangement of CRMs. We test ComSPS on three benchmark datasets and compare it with five existing methods. Experimental results show that ComSPS performs better than them.

A New Algorithm for Identifying Cis-Regulatory Modules Based on Hidden Markov Model

Science.gov (United States)

2017-01-01

The discovery of cis-regulatory modules (CRMs) is the key to understanding mechanisms of transcription regulation. Since CRMs have specific regulatory structures that are the basis for the regulation of gene expression, how to model the regulatory structure of CRMs has a considerable impact on the performance of CRM identification. The paper proposes a CRM discovery algorithm called ComSPS. ComSPS builds a regulatory structure model of CRMs based on HMM by exploring the rules of CRM transcriptional grammar that governs the internal motif site arrangement of CRMs. We test ComSPS on three benchmark datasets and compare it with five existing methods. Experimental results show that ComSPS performs better than them. PMID:28497059

Noise Residual Learning for Noise Modeling in Distributed Video Coding

DEFF Research Database (Denmark)

Luong, Huynh Van; Forchhammer, Søren

2012-01-01

Distributed video coding (DVC) is a coding paradigm which exploits the source statistics at the decoder side to reduce the complexity at the encoder. The noise model is one of the inherently difficult challenges in DVC. This paper considers Transform Domain Wyner-Ziv (TDWZ) coding and proposes...

A method for modeling co-occurrence propensity of clinical codes with application to ICD-10-PCS auto-coding.

Science.gov (United States)

Subotin, Michael; Davis, Anthony R

2016-09-01

Natural language processing methods for medical auto-coding, or automatic generation of medical billing codes from electronic health records, generally assign each code independently of the others. They may thus assign codes for closely related procedures or diagnoses to the same document, even when they do not tend to occur together in practice, simply because the right choice can be difficult to infer from the clinical narrative. We propose a method that injects awareness of the propensities for code co-occurrence into this process. First, a model is trained to estimate the conditional probability that one code is assigned by a human coder, given than another code is known to have been assigned to the same document. Then, at runtime, an iterative algorithm is used to apply this model to the output of an existing statistical auto-coder to modify the confidence scores of the codes. We tested this method in combination with a primary auto-coder for International Statistical Classification of Diseases-10 procedure codes, achieving a 12% relative improvement in F-score over the primary auto-coder baseline. The proposed method can be used, with appropriate features, in combination with any auto-coder that generates codes with different levels of confidence. The promising results obtained for International Statistical Classification of Diseases-10 procedure codes suggest that the proposed method may have wider applications in auto-coding. © The Author 2016. Published by Oxford University Press on behalf of the American Medical Informatics Association. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Improved virtual channel noise model for transform domain Wyner-Ziv video coding

DEFF Research Database (Denmark)

Huang, Xin; Forchhammer, Søren

2009-01-01

Distributed video coding (DVC) has been proposed as a new video coding paradigm to deal with lossy source coding using side information to exploit the statistics at the decoder to reduce computational demands at the encoder. A virtual channel noise model is utilized at the decoder to estimate...... the noise distribution between the side information frame and the original frame. This is one of the most important aspects influencing the coding performance of DVC. Noise models with different granularity have been proposed. In this paper, an improved noise model for transform domain Wyner-Ziv video...... coding is proposed, which utilizes cross-band correlation to estimate the Laplacian parameters more accurately. Experimental results show that the proposed noise model can improve the rate-distortion (RD) performance....

Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2

Energy Technology Data Exchange (ETDEWEB)

Coryell, E.W.; Siefken, L.J.; Harvego, E.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States)] [and others

1997-07-01

The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures. The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds.

Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2

International Nuclear Information System (INIS)

Coryell, E.W.; Siefken, L.J.; Harvego, E.A.

1997-01-01

The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures. The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds

Development of the three dimensional flow model in the SPACE code

International Nuclear Information System (INIS)

Oh, Myung Taek; Park, Chan Eok; Kim, Shin Whan

2014-01-01

SPACE (Safety and Performance Analysis CodE) is a nuclear plant safety analysis code, which has been developed in the Republic of Korea through a joint research between the Korean nuclear industry and research institutes. The SPACE code has been developed with multi-dimensional capabilities as a requirement of the next generation safety code. It allows users to more accurately model the multi-dimensional flow behavior that can be exhibited in components such as the core, lower plenum, upper plenum and downcomer region. Based on generalized models, the code can model any configuration or type of fluid system. All the geometric quantities of mesh are described in terms of cell volume, centroid, face area, and face center, so that it can naturally represent not only the one dimensional (1D) or three dimensional (3D) Cartesian system, but also the cylindrical mesh system. It is possible to simulate large and complex domains by modelling the complex parts with a 3D approach and the rest of the system with a 1D approach. By 1D/3D co-simulation, more realistic conditions and component models can be obtained, providing a deeper understanding of complex systems, and it is expected to overcome the shortcomings of 1D system codes. (author)

The drift flux model in the ASSERT subchannel code

International Nuclear Information System (INIS)

Carver, M.B.; Judd, R.A.; Kiteley, J.C.; Tahir, A.

1987-01-01

The ASSERT subchannel code has been developed specifically to model flow and phase distributions within CANDU fuel bundles. ASSERT uses a drift-flux model that permits the phases to have unequal velocities, and can thus model phase separation tendencies that may occur in horizontal flow. The basic principles of ASSERT are outlined, and computed results are compared against data from various experiments for validation purposes. The paper concludes with an example of the use of the code to predict critical heat flux in CANDU geometries

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

Synchronous versus asynchronous modeling of gene regulatory networks.

Science.gov (United States)

Garg, Abhishek; Di Cara, Alessandro; Xenarios, Ioannis; Mendoza, Luis; De Micheli, Giovanni

2008-09-01

In silico modeling of gene regulatory networks has gained some momentum recently due to increased interest in analyzing the dynamics of biological systems. This has been further facilitated by the increasing availability of experimental data on gene-gene, protein-protein and gene-protein interactions. The two dynamical properties that are often experimentally testable are perturbations and stable steady states. Although a lot of work has been done on the identification of steady states, not much work has been reported on in silico modeling of cellular differentiation processes. In this manuscript, we provide algorithms based on reduced ordered binary decision diagrams (ROBDDs) for Boolean modeling of gene regulatory networks. Algorithms for synchronous and asynchronous transition models have been proposed and their corresponding computational properties have been analyzed. These algorithms allow users to compute cyclic attractors of large networks that are currently not feasible using existing software. Hereby we provide a framework to analyze the effect of multiple gene perturbation protocols, and their effect on cell differentiation processes. These algorithms were validated on the T-helper model showing the correct steady state identification and Th1-Th2 cellular differentiation process. The software binaries for Windows and Linux platforms can be downloaded from http://si2.epfl.ch/~garg/genysis.html.

RCS modeling with the TSAR FDTD code

Energy Technology Data Exchange (ETDEWEB)

Pennock, S.T.; Ray, S.L.

1992-03-01

The TSAR electromagnetic modeling system consists of a family of related codes that have been designed to work together to provide users with a practical way to set up, run, and interpret the results from complex 3-D finite-difference time-domain (FDTD) electromagnetic simulations. The software has been in development at the Lawrence Livermore National Laboratory (LLNL) and at other sites since 1987. Active internal use of the codes began in 1988 with limited external distribution and use beginning in 1991. TSAR was originally developed to analyze high-power microwave and EMP coupling problems. However, the general-purpose nature of the tools has enabled us to use the codes to solve a broader class of electromagnetic applications and has motivated the addition of new features. In particular a family of near-to-far field transformation routines have been added to the codes, enabling TSAR to be used for radar-cross section and antenna analysis problems.

A computer code to estimate accidental fire and radioactive airborne releases in nuclear fuel cycle facilities: User's manual for FIRIN

International Nuclear Information System (INIS)

Chan, M.K.; Ballinger, M.Y.; Owczarski, P.C.

1989-02-01

This manual describes the technical bases and use of the computer code FIRIN. This code was developed to estimate the source term release of smoke and radioactive particles from potential fires in nuclear fuel cycle facilities. FIRIN is a product of a broader study, Fuel Cycle Accident Analysis, which Pacific Northwest Laboratory conducted for the US Nuclear Regulatory Commission. The technical bases of FIRIN consist of a nonradioactive fire source term model, compartment effects modeling, and radioactive source term models. These three elements interact with each other in the code affecting the course of the fire. This report also serves as a complete FIRIN user's manual. Included are the FIRIN code description with methods/algorithms of calculation and subroutines, code operating instructions with input requirements, and output descriptions. 40 refs., 5 figs., 31 tabs

Regulatory Safety Issues in the Structural Design Criteria of ASME Section III Subsection NH and for Very High Temperatures for VHTR and GEN IV

International Nuclear Information System (INIS)

O'Donnell, William J.; Griffin, Donald S.

2007-01-01

The objective of this task is to identify issues relevant to ASME Section III, Subsection NH [1], and related Code Cases that must be resolved for licensing purposes for VHTGRs (Very High Temperature Gas Reactor concepts such as those of PBMR, Areva, and GA); and to identify the material models, design criteria, and analysis methods that need to be added to the ASME Code to cover the unresolved safety issues. Subsection NH was originally developed to provide structural design criteria and limits for elevated-temperature design of Liquid Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Breeder Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTGRs, and identified issues that need to be resolved to provide a regulatory basis for licensing. This Report describes: (1) NRC and ACRS safety concerns raised during the licensing process of CRBR , (2) how some of these issues are addressed by the current Subsection NH of the ASME Code; and (3) the material models, design criteria, and analysis methods that need to be added to the ASME Code and Code Cases to cover unresolved regulatory issues for very high temperature service.

XcisClique: analysis of regulatory bicliques

Directory of Open Access Journals (Sweden)

Grene Ruth

2006-04-01

Full Text Available Abstract Background Modeling of cis-elements or regulatory motifs in promoter (upstream regions of genes is a challenging computational problem. In this work, set of regulatory motifs simultaneously present in the promoters of a set of genes is modeled as a biclique in a suitably defined bipartite graph. A biologically meaningful co-occurrence of multiple cis-elements in a gene promoter is assessed by the combined analysis of genomic and gene expression data. Greater statistical significance is associated with a set of genes that shares a common set of regulatory motifs, while simultaneously exhibiting highly correlated gene expression under given experimental conditions. Methods XcisClique, the system developed in this work, is a comprehensive infrastructure that associates annotated genome and gene expression data, models known cis-elements as regular expressions, identifies maximal bicliques in a bipartite gene-motif graph; and ranks bicliques based on their computed statistical significance. Significance is a function of the probability of occurrence of those motifs in a biclique (a hypergeometric distribution, and on the new sum of absolute values statistic (SAV that uses Spearman correlations of gene expression vectors. SAV is a statistic well-suited for this purpose as described in the discussion. Results XcisClique identifies new motif and gene combinations that might indicate as yet unidentified involvement of sets of genes in biological functions and processes. It currently supports Arabidopsis thaliana and can be adapted to other organisms, assuming the existence of annotated genomic sequences, suitable gene expression data, and identified regulatory motifs. A subset of Xcis Clique functionalities, including the motif visualization component MotifSee, source code, and supplementary material are available at https://bioinformatics.cs.vt.edu/xcisclique/.

Site-specific Probabilistic Analysis of DCGLs Using RESRAD Code

Energy Technology Data Exchange (ETDEWEB)

Kim, Jeongju; Yoon, Suk Bon; Sohn, Wook [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

In general, DCGLs can be conservative (screening DCGL) if they do not take into account site specific factors. Use of such conservative DCGLs can lead to additional remediation that would not be required if the effort was made to develop site-specific DCGLs. Therefore, the objective of this work is to provide an example on the use of the RESRAD 6.0 probabilistic (site-specific) dose analysis to compare with the screening DCGL. Site release regulations state that a site will be considered acceptable for unrestricted use if the residual radioactivity that is distinguishable from background radiation results in a Total Effective Dose Equivalent (TEDE) to an average member of the critical group of less than the site release criteria, for example 0.25 mSv per year in U.S. Utilities use computer dose modeling codes to establish an acceptable level of contamination, the derived concentration guideline level (DCGL) that will meet this regulatory limit. Since the DCGL value is the principal measure of residual radioactivity, it is critical to understand the technical basis of these dose modeling codes. The objective this work was to provide example on nuclear power plant decommissioning dose analysis in a probabilistic analysis framework. The focus was on the demonstration of regulatory compliance for surface soil contamination using the RESRAD 6.0 code. Both the screening and site-specific probabilistic dose analysis methodologies were examined. Example analyses performed with the screening probabilistic dose analysis confirmed the conservatism of the NRC screening values and indicated the effectiveness of probabilistic dose analysis in reducing the conservatism in DCGL derivation.

GASFLOW computer code (physical models and input data)

International Nuclear Information System (INIS)

Muehlbauer, Petr

2007-11-01

The GASFLOW computer code was developed jointly by the Los Alamos National Laboratory, USA, and Forschungszentrum Karlsruhe, Germany. The code is primarily intended for calculations of the transport, mixing, and combustion of hydrogen and other gases in nuclear reactor containments and in other facilities. The physical models and the input data are described, and a commented simple calculation is presented

A primer on thermodynamic-based models for deciphering transcriptional regulatory logic.

Science.gov (United States)

Dresch, Jacqueline M; Richards, Megan; Ay, Ahmet

2013-09-01

A rigorous analysis of transcriptional regulation at the DNA level is crucial to the understanding of many biological systems. Mathematical modeling has offered researchers a new approach to understanding this central process. In particular, thermodynamic-based modeling represents the most biophysically informed approach aimed at connecting DNA level regulatory sequences to the expression of specific genes. The goal of this review is to give biologists a thorough description of the steps involved in building, analyzing, and implementing a thermodynamic-based model of transcriptional regulation. The data requirements for this modeling approach are described, the derivation for a specific regulatory region is shown, and the challenges and future directions for the quantitative modeling of gene regulation are discussed. Copyright © 2013 Elsevier B.V. All rights reserved.

Learning-Based Just-Noticeable-Quantization- Distortion Modeling for Perceptual Video Coding.

Science.gov (United States)

Ki, Sehwan; Bae, Sung-Ho; Kim, Munchurl; Ko, Hyunsuk

2018-07-01

Conventional predictive video coding-based approaches are reaching the limit of their potential coding efficiency improvements, because of severely increasing computation complexity. As an alternative approach, perceptual video coding (PVC) has attempted to achieve high coding efficiency by eliminating perceptual redundancy, using just-noticeable-distortion (JND) directed PVC. The previous JNDs were modeled by adding white Gaussian noise or specific signal patterns into the original images, which were not appropriate in finding JND thresholds due to distortion with energy reduction. In this paper, we present a novel discrete cosine transform-based energy-reduced JND model, called ERJND, that is more suitable for JND-based PVC schemes. Then, the proposed ERJND model is extended to two learning-based just-noticeable-quantization-distortion (JNQD) models as preprocessing that can be applied for perceptual video coding. The two JNQD models can automatically adjust JND levels based on given quantization step sizes. One of the two JNQD models, called LR-JNQD, is based on linear regression and determines the model parameter for JNQD based on extracted handcraft features. The other JNQD model is based on a convolution neural network (CNN), called CNN-JNQD. To our best knowledge, our paper is the first approach to automatically adjust JND levels according to quantization step sizes for preprocessing the input to video encoders. In experiments, both the LR-JNQD and CNN-JNQD models were applied to high efficiency video coding (HEVC) and yielded maximum (average) bitrate reductions of 38.51% (10.38%) and 67.88% (24.91%), respectively, with little subjective video quality degradation, compared with the input without preprocessing applied.

Utility experience in code updating of equipment built to 1974 code, Section 3, Subsection NF

International Nuclear Information System (INIS)

Rao, K.R.; Deshpande, N.

1990-01-01

This paper addresses changes to ASME Code Subsection NF and reconciles the differences between the updated codes and the as built construction code, of ASME Section III, 1974 to which several nuclear plants have been built. Since Section III is revised every three years and replacement parts complying with the construction code are invariably not available from the plant stock inventory, parts must be procured from vendors who comply with the requirements of the latest codes. Aspects of the ASME code which reflect Subsection NF are identified and compared with the later Code editions and addenda, especially up to and including the 1974 ASME code used as the basis for the plant qualification. The concern of the regulatory agencies is that if later code allowables and provisions are adopted it is possible to reduce the safety margins of the construction code. Areas of concern are highlighted and the specific changes of later codes are discerned; adoption of which, would not sacrifice the intended safety margins of the codes to which plants are licensed

Collection of regulatory texts related to radiation protection (collection of legal and regulatory measures related to radiation protection). Part 1: laws and decrees (Extracts of the Public Health Code and of the Labour Code dealing with the protection of population, patients and workers against the hazards of ionizing radiations); Part 2: orders, decisions, non codified decrees (Orders and decisions taken in application of the Public Health Code and of the Labour Code dealing with the protection of population, patients and workers against the hazards of ionizing radiations)

International Nuclear Information System (INIS)

Rivas, R.; Saad, N.; Niel, X.; Cottin, V.; Lachaume, J.L.; Feries, J.

2011-01-01

The first part contains legal and regulatory texts extracted from the Public Health Code and related to health general protection and to health products (medical devices), from the Social Security Code, and from the Labour Code related to individual work relationships, to health and safety at work, to work places, to work equipment and means of protection, to the prevention of some exposure risks and of risks related to some activities. The second part gathers texts extracted from the Public Health Code and related to ionizing radiations (general measures for the protection of the population, exposure to natural radiations, general regime of authorizations and declarations, purchase, retailing, importation, exportation, transfer and elimination of radioactive sources, protection of persons exposed to ionizing radiations for medical or forensics purposes, situations of radiological emergency and of sustained exposure to ionizing radiations, control), to the safety of waters and food products, and to the control of medical devices, to the protection of patients. It also contains extracts for the Labour Code related to workers protection

Modeling ion exchange in clinoptilolite using the EQ3/6 geochemical modeling code

International Nuclear Information System (INIS)

Viani, B.E.; Bruton, C.J.

1992-06-01

Assessing the suitability of Yucca Mtn., NV as a potential repository for high-level nuclear waste requires the means to simulate ion-exchange behavior of zeolites. Vanselow and Gapon convention cation-exchange models have been added to geochemical modeling codes EQ3NR/EQ6, allowing exchange to be modeled for up to three exchangers or a single exchanger with three independent sites. Solid-solution models that are numerically equivalent to the ion-exchange models were derived and also implemented in the code. The Gapon model is inconsistent with experimental adsorption isotherms of trace components in clinoptilolite. A one-site Vanselow model can describe adsorption of Cs or Sr on clinoptilolite, but a two-site Vanselow exchange model is necessary to describe K contents of natural clinoptilolites

MIDAS/PK code development using point kinetics model

International Nuclear Information System (INIS)

Song, Y. M.; Park, S. H.

1999-01-01

In this study, a MIDAS/PK code has been developed for analyzing the ATWS (Anticipated Transients Without Scram) which can be one of severe accident initiating events. The MIDAS is an integrated computer code based on the MELCOR code to develop a severe accident risk reduction strategy by Korea Atomic Energy Research Institute. In the mean time, the Chexal-Layman correlation in the current MELCOR, which was developed under a BWR condition, is appeared to be inappropriate for a PWR. So as to provide ATWS analysis capability to the MIDAS code, a point kinetics module, PKINETIC, has first been developed as a stand-alone code whose reference model was selected from the current accident analysis codes. In the next step, the MIDAS/PK code has been developed via coupling PKINETIC with the MIDAS code by inter-connecting several thermal hydraulic parameters between the two codes. Since the major concern in the ATWS analysis is the primary peak pressure during the early few minutes into the accident, the peak pressure from the PKINETIC module and the MIDAS/PK are compared with the RETRAN calculations showing a good agreement between them. The MIDAS/PK code is considered to be valuable for analyzing the plant response during ATWS deterministically, especially for the early domestic Westinghouse plants which rely on the operator procedure instead of an AMSAC (ATWS Mitigating System Actuation Circuitry) against ATWS. This capability of ATWS analysis is also important from the view point of accident management and mitigation

Compilation of documented computer codes applicable to environmental assessment of radioactivity releases

International Nuclear Information System (INIS)

Hoffman, F.O.; Miller, C.W.; Shaeffer, D.L.; Garten, C.T. Jr.; Shor, R.W.; Ensminger, J.T.

1977-04-01

The objective of this paper is to present a compilation of computer codes for the assessment of accidental or routine releases of radioactivity to the environment from nuclear power facilities. The capabilities of 83 computer codes in the areas of environmental transport and radiation dosimetry are summarized in tabular form. This preliminary analysis clearly indicates that the initial efforts in assessment methodology development have concentrated on atmospheric dispersion, external dosimetry, and internal dosimetry via inhalation. The incorporation of terrestrial and aquatic food chain pathways has been a more recent development and reflects the current requirements of environmental legislation and the needs of regulatory agencies. The characteristics of the conceptual models employed by these codes are reviewed. The appendixes include abstracts of the codes and indexes by author, key words, publication description, and title

Tol2 transposon-mediated transgenesis in the Midas cichlid (Amphilophus citrinellus) - towards understanding gene function and regulatory evolution in an ecological model system for rapid phenotypic diversification.

Science.gov (United States)

Kratochwil, Claudius F; Sefton, Maggie M; Liang, Yipeng; Meyer, Axel

2017-11-23

The Midas cichlid species complex (Amphilophus spp.) is widely known among evolutionary biologists as a model system for sympatric speciation and adaptive phenotypic divergence within extremely short periods of time (a few hundred generations). The repeated parallel evolution of adaptive phenotypes in this radiation, combined with their near genetic identity, makes them an excellent model for studying phenotypic diversification. While many ecological and evolutionary studies have been performed on Midas cichlids, the molecular basis of specific phenotypes, particularly adaptations, and their underlying coding and cis-regulatory changes have not yet been studied thoroughly. For the first time in any New World cichlid, we use Tol2 transposon-mediated transgenesis in the Midas cichlid (Amphilophus citrinellus). By adapting existing microinjection protocols, we established an effective protocol for transgenesis in Midas cichlids. Embryos were injected with a Tol2 plasmid construct that drives enhanced green fluorescent protein (eGFP) expression under the control of the ubiquitin promoter. The transgene was successfully integrated into the germline, driving strong ubiquitous expression of eGFP in the first transgenic Midas cichlid line. Additionally, we show transient expression of two further transgenic constructs, ubiquitin::tdTomato and mitfa::eGFP. Transgenesis in Midas cichlids will facilitate further investigation of the genetic basis of species-specific traits, many of which are adaptations. Transgenesis is a versatile tool not only for studying regulatory elements such as promoters and enhancers, but also for testing gene function through overexpression of allelic gene variants. As such, it is an important first step in establishing the Midas cichlid as a powerful model for studying adaptive coding and non-coding changes in an ecological and evolutionary context.

Implementation of JAERI's reflood model into TRAC-PF1/MOD1 code

International Nuclear Information System (INIS)

Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio

1993-02-01

Selected physical models of REFLA code, that is a reflood analysis code developed at JAERI, were implemented into the TRAC-PF1/MOD1 code in order to improve the predictive capability of the TRAC-PF1/MOD1 code for the core thermal hydraulic behaviors during the reflood phase in a PWR LOCA. Through comparisons of physical models between both codes, (1) Murao-Iguchi void fraction correlation, (2) the drag coefficient correlation acting to drops, (3) the correlation for wall heat transfer coefficient in the film boiling regime, (4) the quench velocity correlation and (5) heat transfer correlations for the dispersed flow regime were selected from the REFLA code to be implemented into the TRAC-PF1/MOD1 code. A method for the transformation of the void fraction correlation to the equivalent interfacial friction model was developed and the effect of the transformation method on the stability of the solution was discussed. Through assessment calculation using data from CCTF (Cylindrical Core Test Facility) flat power test, it was confirmed that the predictive capability of the TRAC code for the core thermal hydraulic behaviors during the reflood can be improved by the implementation of selected physical models of the REFLA code. Several user guidelines for the modified TRAC code were proposed based on the sensitivity studies on fluid cell number in the hydraulic calculation and on node number and effect of axial heat conduction in the heat conduction calculation of fuel rod. (author)

HCPCS Coding: An Integral Part of Your Reimbursement Strategy.

Science.gov (United States)

Nusgart, Marcia

2013-12-01

The first step to a successful reimbursement strategy is to ensure that your wound care product has the most appropriate Healthcare Common Procedure Coding System (HCPCS) code (or billing) for your product. The correct HCPCS code plays an essential role in patient access to new and existing technologies. When devising a strategy to obtain a HCPCS code for its product, companies must consider a number of factors as follows: (1) Has the product gone through the Food and Drug Administration (FDA) regulatory process or does it need to do so? Will the FDA code designation impact which HCPCS code will be assigned to your product? (2) In what "site of service" do you intend to market your product? Where will your customers use the product? Which coding system (CPT ® or HCPCS) applies to your product? (3) Does a HCPCS code for a similar product already exist? Does your product fit under the existing HCPCS code? (4) Does your product need a new HCPCS code? What is the linkage, if any, between coding, payment, and coverage for the product? Researchers and companies need to start early and place the same emphasis on a reimbursement strategy as it does on a regulatory strategy. Your reimbursement strategy staff should be involved early in the process, preferably during product research and development and clinical trial discussions.

Lost opportunities: Modeling commercial building energy code adoption in the United States

International Nuclear Information System (INIS)

Nelson, Hal T.

2012-01-01

This paper models the adoption of commercial building energy codes in the US between 1977 and 2006. Energy code adoption typically results in an increase in aggregate social welfare by cost effectively reducing energy expenditures. Using a Cox proportional hazards model, I test if relative state funding, a new, objective, multivariate regression-derived measure of government capacity, as well as a vector of control variables commonly used in comparative state research, predict commercial building energy code adoption. The research shows little political influence over historical commercial building energy code adoption in the sample. Colder climates and higher electricity prices also do not predict more frequent code adoptions. I do find evidence of high government capacity states being 60 percent more likely than low capacity states to adopt commercial building energy codes in the following year. Wealthier states are also more likely to adopt commercial codes. Policy recommendations to increase building code adoption include increasing access to low cost capital for the private sector and providing noncompetitive block grants to the states from the federal government. - Highlights: ► Model the adoption of commercial building energy codes from 1977–2006 in the US. ► Little political influence over historical building energy code adoption. ► High capacity states are over 60 percent more likely than low capacity states to adopt codes. ► Wealthier states are more likely to adopt commercial codes. ► Access to capital and technical assistance is critical to increase code adoption.

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.

Organization of Risk Analysis Codes for Living Evaluations (ORACLE)

International Nuclear Information System (INIS)

Batt, D.L.; MacDonald, P.E.; Sattison, M.B.; Vesely, E.

1987-01-01

ORACLE (Organization of Risk Analysis Codes for Living Evaluations) is an integration concept for using risk-based information in United States Nuclear Regulatory Commission (USNRC) applications. Portions of ORACLE are being developed at the Idaho Nationale Engineering Laboratory for the USNRC. The ORACLE concept consists of related databases, software, user interfaces, processes, and quality control checks allowing a wide variety of regulatory problems and activities to be addressed using current, updated PRA information. The ORACLE concept provides for smooth transitions between one code and the next without pre- or post-processing. (orig.)

ADVANCED ELECTRIC AND MAGNETIC MATERIAL MODELS FOR FDTD ELECTROMAGNETIC CODES

Energy Technology Data Exchange (ETDEWEB)

Poole, B R; Nelson, S D; Langdon, S

2005-05-05

The modeling of dielectric and magnetic materials in the time domain is required for pulse power applications, pulsed induction accelerators, and advanced transmission lines. For example, most induction accelerator modules require the use of magnetic materials to provide adequate Volt-sec during the acceleration pulse. These models require hysteresis and saturation to simulate the saturation wavefront in a multipulse environment. In high voltage transmission line applications such as shock or soliton lines the dielectric is operating in a highly nonlinear regime, which require nonlinear models. Simple 1-D models are developed for fast parameterization of transmission line structures. In the case of nonlinear dielectrics, a simple analytic model describing the permittivity in terms of electric field is used in a 3-D finite difference time domain code (FDTD). In the case of magnetic materials, both rate independent and rate dependent Hodgdon magnetic material models have been implemented into 3-D FDTD codes and 1-D codes.

ADVANCED ELECTRIC AND MAGNETIC MATERIAL MODELS FOR FDTD ELECTROMAGNETIC CODES

International Nuclear Information System (INIS)

Poole, B R; Nelson, S D; Langdon, S

2005-01-01

The modeling of dielectric and magnetic materials in the time domain is required for pulse power applications, pulsed induction accelerators, and advanced transmission lines. For example, most induction accelerator modules require the use of magnetic materials to provide adequate Volt-sec during the acceleration pulse. These models require hysteresis and saturation to simulate the saturation wavefront in a multipulse environment. In high voltage transmission line applications such as shock or soliton lines the dielectric is operating in a highly nonlinear regime, which require nonlinear models. Simple 1-D models are developed for fast parameterization of transmission line structures. In the case of nonlinear dielectrics, a simple analytic model describing the permittivity in terms of electric field is used in a 3-D finite difference time domain code (FDTD). In the case of magnetic materials, both rate independent and rate dependent Hodgdon magnetic material models have been implemented into 3-D FDTD codes and 1-D codes

Two-phase wall friction model for the trace computer code

International Nuclear Information System (INIS)

Wang Weidong

2005-01-01

The wall drag model in the TRAC/RELAP5 Advanced Computational Engine computer code (TRACE) has certain known deficiencies. For example, in an annular flow regime, the code predicts an unphysical high liquid velocity compared to the experimental data. To address those deficiencies, a new wall frictional drag package has been developed and implemented in the TRACE code to model the wall drag for two-phase flow system code. The modeled flow regimes are (1) annular/mist, (2) bubbly/slug, and (3) bubbly/slug with wall nucleation. The new models use void fraction (instead of flow quality) as the correlating variable to minimize the calculation oscillation. In addition, the models allow for transitions between the three regimes. The annular/mist regime is subdivided into three separate regimes for pure annular flow, annular flow with entrainment, and film breakdown. For adiabatic two-phase bubbly/slug flows, the vapor phase primarily exists outside of the boundary layer, and the wall shear uses single-phase liquid velocity for friction calculation. The vapor phase wall friction drag is set to zero for bubbly/slug flows. For bubbly/slug flows with wall nucleation, the bubbles are presented within the hydrodynamic boundary layer, and the two-phase wall friction drag is significantly higher with a pronounced mass flux effect. An empirical correlation has been studied and applied to account for nucleate boiling. Verification and validation tests have been performed, and the test results showed a significant code improvement. (authors)

Code-To-Code Benchmarking Of The Porflow And GoldSim Contaminant Transport Models Using A Simple 1-D Domain - 11191

International Nuclear Information System (INIS)

Hiergesell, R.; Taylor, G.

2010-01-01

An investigation was conducted to compare and evaluate contaminant transport results of two model codes, GoldSim and Porflow, using a simple 1-D string of elements in each code. Model domains were constructed to be identical with respect to cell numbers and dimensions, matrix material, flow boundary and saturation conditions. One of the codes, GoldSim, does not simulate advective movement of water; therefore the water flux term was specified as a boundary condition. In the other code, Porflow, a steady-state flow field was computed and contaminant transport was simulated within that flow-field. The comparisons were made solely in terms of the ability of each code to perform contaminant transport. The purpose of the investigation was to establish a basis for, and to validate follow-on work that was conducted in which a 1-D GoldSim model developed by abstracting information from Porflow 2-D and 3-D unsaturated and saturated zone models and then benchmarked to produce equivalent contaminant transport results. A handful of contaminants were selected for the code-to-code comparison simulations, including a non-sorbing tracer and several long- and short-lived radionuclides exhibiting both non-sorbing to strongly-sorbing characteristics with respect to the matrix material, including several requiring the simulation of in-growth of daughter radionuclides. The same diffusion and partitioning coefficients associated with each contaminant and the half-lives associated with each radionuclide were incorporated into each model. A string of 10-elements, having identical spatial dimensions and properties, were constructed within each code. GoldSim's basic contaminant transport elements, Mixing cells, were utilized in this construction. Sand was established as the matrix material and was assigned identical properties (e.g. bulk density, porosity, saturated hydraulic conductivity) in both codes. Boundary conditions applied included an influx of water at the rate of 40 cm/yr at one

GASFLOW: A Computational Fluid Dynamics Code for Gases, Aerosols, and Combustion, Volume 1: Theory and Computational Model

International Nuclear Information System (INIS)

Nichols, B.D.; Mueller, C.; Necker, G.A.; Travis, J.R.; Spore, J.W.; Lam, K.L.; Royl, P.; Redlinger, R.; Wilson, T.L.

1998-01-01

Los Alamos National Laboratory (LANL) and Forschungszentrum Karlsruhe (FzK) are developing GASFLOW, a three-dimensional (3D) fluid dynamics field code as a best-estimate tool to characterize local phenomena within a flow field. Examples of 3D phenomena include circulation patterns; flow stratification; hydrogen distribution mixing and stratification; combustion and flame propagation; effects of noncondensable gas distribution on local condensation and evaporation; and aerosol entrainment, transport, and deposition. An analysis with GASFLOW will result in a prediction of the gas composition and discrete particle distribution in space and time throughout the facility and the resulting pressure and temperature loadings on the walls and internal structures with or without combustion. A major application of GASFLOW is for predicting the transport, mixing, and combustion of hydrogen and other gases in nuclear reactor containments and other facilities. It has been applied to situations involving transporting and distributing combustible gas mixtures. It has been used to study gas dynamic behavior (1) in low-speed, buoyancy-driven flows, as well as sonic flows or diffusion dominated flows; and (2) during chemically reacting flows, including deflagrations. The effects of controlling such mixtures by safety systems can be analyzed. The code version described in this manual is designated GASFLOW 2.1, which combines previous versions of the United States Nuclear Regulatory Commission code HMS (for Hydrogen Mixing Studies) and the Department of Energy and FzK versions of GASFLOW. The code was written in standard Fortran 90. This manual comprises three volumes. Volume I describes the governing physical equations and computational model. Volume II describes how to use the code to set up a model geometry, specify gas species and material properties, define initial and boundary conditions, and specify different outputs, especially graphical displays. Sample problems are included

Development of health effect assessment software using MACCS2 code

International Nuclear Information System (INIS)

Hwang, Seok-Won; Park, Jong-Woon; Kang, Kyung Min; Jae, Moosung

2008-01-01

The extended regulatory interests in severe accidents management and enhanced safety regulatory requirements raise a need of more accurate analysis of the effect to the public health by users with diverse disciplines. This facilitates this work to develop web-based radiation health effect assessment software, RASUM, by using the MACCS2 code and HTML language to provide diverse users (regulators, operators, and public) with easy understanding, modeling, calculating, analyzing, documenting and reporting of the radiation health effect under hypothetical severe accidents. The engine of the web-based RASUM uses the MACCS2 as a base code developed by NRC and is composed of five modules such as development module, PSA training module, output module, input data module (source term, population distribution, meteorological data, etc.), and MACCS2 run module. For verification and demonstration of the RASUM, the offsite consequence analysis using the RASUM frame is performed for such as early fatality risk, organ does, and whole body does for two selected scenarios. Moreover, CCDF results from the RASUM for KSNP and CANDU type reactors are presented and compared. (author)

Development of Regulatory Thermal-Hydraulic Analysis System (RETAS)

Energy Technology Data Exchange (ETDEWEB)

Ahn, Seung-Hoon; Kim, In-Goo; Kim, Hho-Jung; Cho, Yong Jin [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2007-10-15

A review is provided of the reasons why the Korea Institute of Nuclear Safety needs improvement of the existing codes employed for a regulatory audit. The proposed new organization of the codes, developed or to be developed, is presented together with illustrative applications. Inspection of the quality assurance activities is planned to ensure the robustness of MARS (Multi-dimensional Analysis for Reactor Safety) code, served as a pivot of the organization.

Development of Regulatory Thermal-Hydraulic Analysis System (RETAS)

International Nuclear Information System (INIS)

Ahn, Seung-Hoon; Kim, In-Goo; Kim, Hho-Jung; Cho, Yong Jin

2007-01-01

A review is provided of the reasons why the Korea Institute of Nuclear Safety needs improvement of the existing codes employed for a regulatory audit. The proposed new organization of the codes, developed or to be developed, is presented together with illustrative applications. Inspection of the quality assurance activities is planned to ensure the robustness of MARS (Multi-dimensional Analysis for Reactor Safety) code, served as a pivot of the organization

Fuel analysis code FAIR and its high burnup modelling capabilities

International Nuclear Information System (INIS)

Prasad, P.S.; Dutta, B.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

1995-01-01

A computer code FAIR has been developed for analysing performance of water cooled reactor fuel pins. It is capable of analysing high burnup fuels. This code has recently been used for analysing ten high burnup fuel rods irradiated at Halden reactor. In the present paper, the code FAIR and its various high burnup models are described. The performance of code FAIR in analysing high burnup fuels and its other applications are highlighted. (author). 21 refs., 12 figs

The 2010 fib Model Code for Structural Concrete: A new approach to structural engineering

NARCIS (Netherlands)

Walraven, J.C.; Bigaj-Van Vliet, A.

2011-01-01

The fib Model Code is a recommendation for the design of reinforced and prestressed concrete which is intended to be a guiding document for future codes. Model Codes have been published before, in 1978 and 1990. The draft for fib Model Code 2010 was published in May 2010. The most important new

Regulatory Framework and Radiation Protection as Basis for Evaluation

International Nuclear Information System (INIS)

Elegba, S.B.

2010-01-01

Regulatory Framework for Nuclear Safety and Radiation Protection International Instruments: Conventions; Safety Fundamentals; Codes of Conduct; Safety Requirements and Guide, and National Instruments:-Legislation; Regulations; Guidance Documents. The Sustainable Development Principle recognizes a duty to prevent undue burden and degradation of the environment for future generations. The prime responsibility for safety must rest with the person or organization responsible for facilities…that give rise to radiation risks” (IAEA Safety Fundamentals – SF-1). Compliance with regulations and requirements imposed by the Regulatory Body shall not relieve the organization of its prime responsibility for safety. The regulatory body shall establish and implement appropriate arrangements for a systematic approach to quality management which extend throughout the range of responsibilities and functions undertaken.”. The IAEA self-assessment model for a Regulatory Body is based on a three tier approach. This model is modular and can be used and adopted for implementation by any regulator at any stage of maturity “Start up”, “On the way”, “Mature” Small, medium size, big. The IAEA is required by its Statute to promote international cooperation while regulating safety is a national responsibility. However, radiation risks may transcend national borders and international cooperation that serves to promote and enhance safety globally by exchanging experience and by improving capabilities to control hazards, to prevent accidents, to respond to emergencies and to mitigate any harmful consequences

Development and validation of the ENIGMA code for MOX fuel performance modelling

International Nuclear Information System (INIS)

Palmer, I.; Rossiter, G.; White, R.J.

2000-01-01

The ENIGMA fuel performance code has been under development in the UK since the mid-1980s with contributions made by both the fuel vendor (BNFL) and the utility (British Energy). In recent years it has become the principal code for UO 2 fuel licensing for both PWR and AGR reactor systems in the UK and has also been used by BNFL in support of overseas UO 2 and MOX fuel business. A significant new programme of work has recently been initiated by BNFL to further develop the code specifically for MOX fuel application. Model development is proceeding hand in hand with a major programme of MOX fuel testing and PIE studies, with the objective of producing a fuel modelling code suitable for mechanistic analysis, as well as for licensing applications. This paper gives an overview of the model developments being undertaken and of the experimental data being used to underpin and to validate the code. The paper provides a summary of the code development programme together with specific examples of new models produced. (author)

7 CFR Exhibit E to Subpart A of... - Voluntary National Model Building Codes

Science.gov (United States)

2010-01-01

... 7 Agriculture 12 2010-01-01 2010-01-01 false Voluntary National Model Building Codes E Exhibit E... National Model Building Codes The following documents address the health and safety aspects of buildings and related structures and are voluntary national model building codes as defined in § 1924.4(h)(2) of...

Three-dimensional modeling with finite element codes

Energy Technology Data Exchange (ETDEWEB)

Druce, R.L.

1986-01-17

This paper describes work done to model magnetostatic field problems in three dimensions. Finite element codes, available at LLNL, and pre- and post-processors were used in the solution of the mathematical model, the output from which agreed well with the experimentally obtained data. The geometry used in this work was a cylinder with ports in the periphery and no current sources in the space modeled. 6 refs., 8 figs.

Uncertainty Methods Framework Development for the TRACE Thermal-Hydraulics Code by the U.S.NRC

International Nuclear Information System (INIS)

Bajorek, Stephen M.; Gingrich, Chester

2013-01-01

The Code of Federal Regulations, Title 10, Part 50.46 requires that the Emergency Core Cooling System (ECCS) performance be evaluated for a number of postulated Loss-Of-Coolant-Accidents (LOCAs). The rule allows two methods for calculation of the acceptance criteria; using a realistic model in the so-called 'Best Estimate' approach, or the more prescriptive following Appendix K to Part 50. Because of the conservatism of Appendix K, recent Evaluation Model submittals to the NRC used the realistic approach. With this approach, the Evaluation Model must demonstrate that the Peak Cladding Temperature (PCT), the Maximum Local Oxidation (MLO) and Core-Wide Oxidation (CWO) remain below their regulatory limits with a 'high probability'. Guidance for Best Estimate calculations following 50.46(a)(1) was provided by Regulatory Guide 1.157. This Guide identified a 95% probability level as being acceptable for comparisons of best-estimate predictions to the applicable regulatory limits, but was vague with respect to acceptable methods in which to determine the code uncertainty. Nor, did it specify if a confidence level should be determined. As a result, vendors have developed Evaluation Models utilizing several different methods to combine uncertainty parameters and determine the PCT and other variables to a high probability. In order to quantify the accuracy of TRACE calculations for a wide variety of applications and to audit Best Estimate calculations made by industry, the NRC is developing its own independent methodology to determine the peak cladding temperature and other parameters of regulatory interest to a high probability. Because several methods are in use, and each vendor's methodology ranges different parameters, the NRC method must be flexible and sufficiently general. Not only must the method apply to LOCA analysis for conventional light-water reactors, it must also be extendable to new reactor designs and type of analyses where the acceptance criteria are less

INDOSE V2.1.1, Internal Dosimetry Code Using Biokinetics Models

International Nuclear Information System (INIS)

Silverman, Ido

2002-01-01

A - Description of program or function: InDose is an internal dosimetry code developed to enable dose estimations using the new biokinetic models (presented in ICRP-56 to ICRP71) as well as the old ones. The code is written in FORTRAN90 and uses the ICRP-66 respiratory tract model and the ICRP-30 gastrointestinal tract model as well as the new and old biokinetic models. The code has been written in such a way that the user is able to change any of the parameters of any one of the models without recompiling the code. All the parameters are given in well annotated parameters files that the user may change and the code reads during invocation. As default, these files contains the values listed in ICRP publications. The full InDose code is planed to have three parts: 1) the main part includes the uptake and systemic models and is used to calculate the activities in the body tissues and excretion as a function of time for a given intake. 2) An optimization module for automatic estimation of the intake for a specific exposure case. 3) A module to calculate the dose due to the estimated intake. Currently, the code is able to perform only its main task (part 1) while the other two have to be done externally using other tools. In the future we would like to add these modules in order to provide a complete solution for the people in the laboratory. The code has been tested extensively to verify the accuracy of its results. The verification procedure was divided into three parts: 1) verification of the implementation of each model, 2) verification of the integrity of the whole code, and 3) usability test. The first two parts consisted of comparing results obtained with InDose to published results for the same cases. For example ICRP-78 monitoring data. The last part consisted of participating in the 3. EIE-IDA and assessing some of the scenarios provided in this exercise. These tests where presented in a few publications. It has been found that there is very good agreement

Modeling radiation belt dynamics using a 3-D layer method code

Science.gov (United States)

Wang, C.; Ma, Q.; Tao, X.; Zhang, Y.; Teng, S.; Albert, J. M.; Chan, A. A.; Li, W.; Ni, B.; Lu, Q.; Wang, S.

2017-08-01

A new 3-D diffusion code using a recently published layer method has been developed to analyze radiation belt electron dynamics. The code guarantees the positivity of the solution even when mixed diffusion terms are included. Unlike most of the previous codes, our 3-D code is developed directly in equatorial pitch angle (α0), momentum (p), and L shell coordinates; this eliminates the need to transform back and forth between (α0,p) coordinates and adiabatic invariant coordinates. Using (α0,p,L) is also convenient for direct comparison with satellite data. The new code has been validated by various numerical tests, and we apply the 3-D code to model the rapid electron flux enhancement following the geomagnetic storm on 17 March 2013, which is one of the Geospace Environment Modeling Focus Group challenge events. An event-specific global chorus wave model, an AL-dependent statistical plasmaspheric hiss wave model, and a recently published radial diffusion coefficient formula from Time History of Events and Macroscale Interactions during Substorms (THEMIS) statistics are used. The simulation results show good agreement with satellite observations, in general, supporting the scenario that the rapid enhancement of radiation belt electron flux for this event results from an increased level of the seed population by radial diffusion, with subsequent acceleration by chorus waves. Our results prove that the layer method can be readily used to model global radiation belt dynamics in three dimensions.

Hamming Code Based Watermarking Scheme for 3D Model Verification

Directory of Open Access Journals (Sweden)

Jen-Tse Wang

2014-01-01

Full Text Available Due to the explosive growth of the Internet and maturing of 3D hardware techniques, protecting 3D objects becomes a more and more important issue. In this paper, a public hamming code based fragile watermarking technique is proposed for 3D objects verification. An adaptive watermark is generated from each cover model by using the hamming code technique. A simple least significant bit (LSB substitution technique is employed for watermark embedding. In the extraction stage, the hamming code based watermark can be verified by using the hamming code checking without embedding any verification information. Experimental results shows that 100% vertices of the cover model can be watermarked, extracted, and verified. It also shows that the proposed method can improve security and achieve low distortion of stego object.

Recent SCDAP/RELAP5 code applications and improvements

International Nuclear Information System (INIS)

Harvego, E.A.; Ghan, L.S.; Knudson, D.L.; Siefken, L.J.

1998-01-01

This paper summarizes (1) a recent application of the severe accident analysis code SCDAP/RELAP5/MOD3.1, and (2) development and assessment activities associated with the release of SACDAP/RELAP5/MOD3.2. The Nuclear Regulatory Commission (NRC) has been evaluating the integrity of steam generator tubes during severe accidents. MOD3.1 has been used to support that evaluation. Studies indicate that the pressurizer surge line will fail before any steam generator tubes are damaged. Thus, core decay energy would be released as steam through the surge line and the tube wall would be spared from exposure to prolonged flow of high temperature steam. The latest code version, MOD3.2, contains several improvements to models that address both the early phase and late phase of a severe accident. The impact of these improvements to the overall code capabilities has been assessed. Results of the assessment are summarized in this paper

Evolution of Cis-Regulatory Elements and Regulatory Networks in Duplicated Genes of Arabidopsis.

Science.gov (United States)

Arsovski, Andrej A; Pradinuk, Julian; Guo, Xu Qiu; Wang, Sishuo; Adams, Keith L

2015-12-01

Plant genomes contain large numbers of duplicated genes that contribute to the evolution of new functions. Following duplication, genes can exhibit divergence in their coding sequence and their expression patterns. Changes in the cis-regulatory element landscape can result in changes in gene expression patterns. High-throughput methods developed recently can identify potential cis-regulatory elements on a genome-wide scale. Here, we use a recent comprehensive data set of DNase I sequencing-identified cis-regulatory binding sites (footprints) at single-base-pair resolution to compare binding sites and network connectivity in duplicated gene pairs in Arabidopsis (Arabidopsis thaliana). We found that duplicated gene pairs vary greatly in their cis-regulatory element architecture, resulting in changes in regulatory network connectivity. Whole-genome duplicates (WGDs) have approximately twice as many footprints in their promoters left by potential regulatory proteins than do tandem duplicates (TDs). The WGDs have a greater average number of footprint differences between paralogs than TDs. The footprints, in turn, result in more regulatory network connections between WGDs and other genes, forming denser, more complex regulatory networks than shown by TDs. When comparing regulatory connections between duplicates, WGDs had more pairs in which the two genes are either partially or fully diverged in their network connections, but fewer genes with no network connections than the TDs. There is evidence of younger TDs and WGDs having fewer unique connections compared with older duplicates. This study provides insights into cis-regulatory element evolution and network divergence in duplicated genes. © 2015 American Society of Plant Biologists. All Rights Reserved.

Self-shielding models of MICROX-2 code: Review and updates

International Nuclear Information System (INIS)

Hou, J.; Choi, H.; Ivanov, K.N.

2014-01-01

Highlights: • The MICROX-2 code has been improved to expand its application to advanced reactors. • New fine-group cross section libraries based on ENDF/B-VII have been generated. • Resonance self-shielding and spatial self-shielding models have been improved. • The improvements were assessed by a series of benchmark calculations against MCNPX. - Abstract: The MICROX-2 is a transport theory code that solves for the neutron slowing-down and thermalization equations of a two-region lattice cell. The MICROX-2 code has been updated to expand its application to advanced reactor concepts and fuel cycle simulations, including generation of new fine-group cross section libraries based on ENDF/B-VII. In continuation of previous work, the MICROX-2 methods are reviewed and updated in this study, focusing on its resonance self-shielding and spatial self-shielding models for neutron spectrum calculations. The improvement of self-shielding method was assessed by a series of benchmark calculations against the Monte Carlo code, using homogeneous and heterogeneous pin cell models. The results have shown that the implementation of the updated self-shielding models is correct and the accuracy of physics calculation is improved. Compared to the existing models, the updates reduced the prediction error of the infinite multiplication factor by ∼0.1% and ∼0.2% for the homogeneous and heterogeneous pin cell models, respectively, considered in this study

Application of the thermal-hydraulic codes in VVER-440 steam generators modelling

Energy Technology Data Exchange (ETDEWEB)

Matejovic, P.; Vranca, L.; Vaclav, E. [Nuclear Power Plant Research Inst. VUJE (Slovakia)

1995-12-31

Performances with the CATHARE2 V1.3U and RELAP5/MOD3.0 application to the VVER-440 SG modelling during normal conditions and during transient with secondary water lowering are described. Similar recirculation model was chosen for both codes. In the CATHARE calculation, no special measures were taken with the aim to optimize artificially flow rate distribution coefficients for the junction between SG riser and steam dome. Contrary to RELAP code, the CATHARE code is able to predict reasonable the secondary swell level in nominal conditions. Both codes are able to model properly natural phase separation on the SG water level. 6 refs.

Application of the thermal-hydraulic codes in VVER-440 steam generators modelling

Energy Technology Data Exchange (ETDEWEB)

Matejovic, P; Vranca, L; Vaclav, E [Nuclear Power Plant Research Inst. VUJE (Slovakia)

1996-12-31

Performances with the CATHARE2 V1.3U and RELAP5/MOD3.0 application to the VVER-440 SG modelling during normal conditions and during transient with secondary water lowering are described. Similar recirculation model was chosen for both codes. In the CATHARE calculation, no special measures were taken with the aim to optimize artificially flow rate distribution coefficients for the junction between SG riser and steam dome. Contrary to RELAP code, the CATHARE code is able to predict reasonable the secondary swell level in nominal conditions. Both codes are able to model properly natural phase separation on the SG water level. 6 refs.

The new French code for PWR in service inspection

Energy Technology Data Exchange (ETDEWEB)

Noel, R; Hutin, J P [Electricite de France (EDF), 75 - Paris (France)

1988-12-31

This document presents the new french code for pressured water reactor in service inspection. The historic regulatory basis is presented, together with the new regulatory act (dating back to the 26 february 1974) and the major guidelines of the french practice for in service inspection of PWR components. (TEC).

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.

Improvement of blow down model for LEAP code

International Nuclear Information System (INIS)

Itooka, Satoshi; Fujimata, Kazuhiro

2003-03-01

In Japan Nuclear Cycle Development Institute, the improvement of analysis method for overheating tube rapture was studied for the accident of sodium-water reactions in the steam generator of a fast breeder reactor and the evaluation of heat transfer condition in the tube were carried out based on study of critical heat flux (CHF) and post-CHF heat transfer equation in Light Water Reactors. In this study, the improvement of blow down model for the LEAP code was carried out taking into consideration the above-mentioned evaluation of heat transfer condition. Improvements of the LEAP code were following items. Calculations and verification were performed with the improved LEAP code in order to confirm the code functions. The addition of critical heat flux (CHF) by the formula of Katto and the formula of Tong. The addition of post-CHF heat transfer equation by the formula of Condie-BengstonIV and the formula of Groeneveld 5.9. The physical properties of the water and steam are expanded to the critical conditions of the water. The expansion of the total number of section and the improvement of the input form. The addition of the function to control the valve setting by the PID control model. (author)

Documentation for grants equal to tax model: Volume 3, Source code

International Nuclear Information System (INIS)

Boryczka, M.K.

1986-01-01

The GETT model is capable of forecasting the amount of tax liability associated with all property owned and all activities undertaken by the US Department of Energy (DOE) in site characterization and repository development. The GETT program is a user-friendly, menu-driven model developed using dBASE III/trademark/, a relational data base management system. The data base for GETT consists primarily of eight separate dBASE III/trademark/ files corresponding to each of the eight taxes (real property, personal property, corporate income, franchise, sales, use, severance, and excise) levied by State and local jurisdictions on business property and activity. Additional smaller files help to control model inputs and reporting options. Volume 3 of the GETT model documentation is the source code. The code is arranged primarily by the eight tax types. Other code files include those for JURISDICTION, SIMULATION, VALIDATION, TAXES, CHANGES, REPORTS, GILOT, and GETT. The code has been verified through hand calculations

Properties of Sequence Conservation in Upstream Regulatory and Protein Coding Sequences among Paralogs in Arabidopsis thaliana

Science.gov (United States)

Richardson, Dale N.; Wiehe, Thomas

Whole genome duplication (WGD) has catalyzed the formation of new species, genes with novel functions, altered expression patterns, complexified signaling pathways and has provided organisms a level of genetic robustness. We studied the long-term evolution and interrelationships of 5’ upstream regulatory sequences (URSs), protein coding sequences (CDSs) and expression correlations (EC) of duplicated gene pairs in Arabidopsis. Three distinct methods revealed significant evolutionary conservation between paralogous URSs and were highly correlated with microarray-based expression correlation of the respective gene pairs. Positional information on exact matches between sequences unveiled the contribution of micro-chromosomal rearrangements on expression divergence. A three-way rank analysis of URS similarity, CDS divergence and EC uncovered specific gene functional biases. Transcription factor activity was associated with gene pairs exhibiting conserved URSs and divergent CDSs, whereas a broad array of metabolic enzymes was found to be associated with gene pairs showing diverged URSs but conserved CDSs.

Identification of functional elements and regulatory circuits by Drosophila modENCODE.

Science.gov (United States)

Roy, Sushmita; Ernst, Jason; Kharchenko, Peter V; Kheradpour, Pouya; Negre, Nicolas; Eaton, Matthew L; Landolin, Jane M; Bristow, Christopher A; Ma, Lijia; Lin, Michael F; Washietl, Stefan; Arshinoff, Bradley I; Ay, Ferhat; Meyer, Patrick E; Robine, Nicolas; Washington, Nicole L; Di Stefano, Luisa; Berezikov, Eugene; Brown, Christopher D; Candeias, Rogerio; Carlson, Joseph W; Carr, Adrian; Jungreis, Irwin; Marbach, Daniel; Sealfon, Rachel; Tolstorukov, Michael Y; Will, Sebastian; Alekseyenko, Artyom A; Artieri, Carlo; Booth, Benjamin W; Brooks, Angela N; Dai, Qi; Davis, Carrie A; Duff, Michael O; Feng, Xin; Gorchakov, Andrey A; Gu, Tingting; Henikoff, Jorja G; Kapranov, Philipp; Li, Renhua; MacAlpine, Heather K; Malone, John; Minoda, Aki; Nordman, Jared; Okamura, Katsutomo; Perry, Marc; Powell, Sara K; Riddle, Nicole C; Sakai, Akiko; Samsonova, Anastasia; Sandler, Jeremy E; Schwartz, Yuri B; Sher, Noa; Spokony, Rebecca; Sturgill, David; van Baren, Marijke; Wan, Kenneth H; Yang, Li; Yu, Charles; Feingold, Elise; Good, Peter; Guyer, Mark; Lowdon, Rebecca; Ahmad, Kami; Andrews, Justen; Berger, Bonnie; Brenner, Steven E; Brent, Michael R; Cherbas, Lucy; Elgin, Sarah C R; Gingeras, Thomas R; Grossman, Robert; Hoskins, Roger A; Kaufman, Thomas C; Kent, William; Kuroda, Mitzi I; Orr-Weaver, Terry; Perrimon, Norbert; Pirrotta, Vincenzo; Posakony, James W; Ren, Bing; Russell, Steven; Cherbas, Peter; Graveley, Brenton R; Lewis, Suzanna; Micklem, Gos; Oliver, Brian; Park, Peter J; Celniker, Susan E; Henikoff, Steven; Karpen, Gary H; Lai, Eric C; MacAlpine, David M; Stein, Lincoln D; White, Kevin P; Kellis, Manolis

2010-12-24

To gain insight into how genomic information is translated into cellular and developmental programs, the Drosophila model organism Encyclopedia of DNA Elements (modENCODE) project is comprehensively mapping transcripts, histone modifications, chromosomal proteins, transcription factors, replication proteins and intermediates, and nucleosome properties across a developmental time course and in multiple cell lines. We have generated more than 700 data sets and discovered protein-coding, noncoding, RNA regulatory, replication, and chromatin elements, more than tripling the annotated portion of the Drosophila genome. Correlated activity patterns of these elements reveal a functional regulatory network, which predicts putative new functions for genes, reveals stage- and tissue-specific regulators, and enables gene-expression prediction. Our results provide a foundation for directed experimental and computational studies in Drosophila and related species and also a model for systematic data integration toward comprehensive genomic and functional annotation.

COCOA code for creating mock observations of star cluster models

Science.gov (United States)

Askar, Abbas; Giersz, Mirek; Pych, Wojciech; Dalessandro, Emanuele

2018-04-01

We introduce and present results from the COCOA (Cluster simulatiOn Comparison with ObservAtions) code that has been developed to create idealized mock photometric observations using results from numerical simulations of star cluster evolution. COCOA is able to present the output of realistic numerical simulations of star clusters carried out using Monte Carlo or N-body codes in a way that is useful for direct comparison with photometric observations. In this paper, we describe the COCOA code and demonstrate its different applications by utilizing globular cluster (GC) models simulated with the MOCCA (MOnte Carlo Cluster simulAtor) code. COCOA is used to synthetically observe these different GC models with optical telescopes, perform point spread function photometry, and subsequently produce observed colour-magnitude diagrams. We also use COCOA to compare the results from synthetic observations of a cluster model that has the same age and metallicity as the Galactic GC NGC 2808 with observations of the same cluster carried out with a 2.2 m optical telescope. We find that COCOA can effectively simulate realistic observations and recover photometric data. COCOA has numerous scientific applications that maybe be helpful for both theoreticians and observers that work on star clusters. Plans for further improving and developing the code are also discussed in this paper.

Annual Report 2010. Nuclear Regulatory Authority

International Nuclear Information System (INIS)

2010-01-01

The present Annual Report of Activities of the Nuclear Regulatory Authority (ARN), prepared regularly from the creation as independent institution, describes across six chapters and seven annexes the activities developed by the organism during 2010. The main topic are: institutional issues; regulatory guides and standards; argentinean nuclear regulatory system; quality assurance of the ARN; the institutional communications; the licensing and inspection of nuclear power plants and critical facilities; the emergency systems; the safeguards and the physical protection; the environmental control; the institutional relations; the training and the public information. Also, this publication have annexes with the following content: the regulatory framework; regulatory documents; inspections to medical, industrial and training installations; measurement and evaluation of the drinking water of Ezeiza; international expert's report on the application of the international standards of radiological protection of the public in the zone of the Ezeiza Atomic Center; ethical code

Annual Report 2011. Nuclear Regulatory Authority

International Nuclear Information System (INIS)

2011-01-01

The present Annual Report of Activities of the Nuclear Regulatory Authority (ARN), prepared regularly from the creation as independent institution, describes across six chapters and seven annexes the activities developed by the organism during 2011. The main topic are: institutional issues; regulatory guides and standards; argentinean nuclear regulatory system; quality assurance of the ARN; the institutional communications; the licensing and inspection of nuclear power plants and critical facilities; the emergency systems; the safeguards and the physical protection; the environmental control; the institutional relations; the training and the public information. Also, this publication have annexes with the following content: the regulatory framework; regulatory documents; inspections to medical, industrial and training installations; measurement and evaluation of the drinking water of Ezeiza; international expert's report on the application of the international standards of radiological protection of the public in the zone of the Ezeiza Atomic Center; ethical code

Inspection and enforcement by the regulatory body for nuclear power plants

International Nuclear Information System (INIS)

1980-01-01

This Safety Guide was prepared as part of the Agency's programme, referred to as the NUSS programme, for establishing Codes of Practice and Safety Guides relating to nuclear power plants. It supplements the Code of Practice on Governmental Organization for the Regulation of Nuclear Power Plants, IAEA Safety Series No.50-C-G and should be used in conjunction with that document. The purpose of this Guide is to provide information, guidance and recommendations to assist Member States in (1) establishing and conducting a regulatory inspection programme for nuclear power plants, (2) establishing requirements for the applicant/licensee in regard to regulatory inspection, (3) establishing a system for enforcing compliance with the requirements and decisions of the regulatory body

EMPIRE-II statistical model code for nuclear reaction calculations

Energy Technology Data Exchange (ETDEWEB)

Herman, M [International Atomic Energy Agency, Vienna (Austria)

2001-12-15

EMPIRE II is a nuclear reaction code, comprising various nuclear models, and designed for calculations in the broad range of energies and incident particles. A projectile can be any nucleon or Heavy Ion. The energy range starts just above the resonance region, in the case of neutron projectile, and extends up to few hundreds of MeV for Heavy Ion induced reactions. The code accounts for the major nuclear reaction mechanisms, such as optical model (SCATB), Multistep Direct (ORION + TRISTAN), NVWY Multistep Compound, and the full featured Hauser-Feshbach model. Heavy Ion fusion cross section can be calculated within the simplified coupled channels approach (CCFUS). A comprehensive library of input parameters covers nuclear masses, optical model parameters, ground state deformations, discrete levels and decay schemes, level densities, fission barriers (BARFIT), moments of inertia (MOMFIT), and {gamma}-ray strength functions. Effects of the dynamic deformation of a fast rotating nucleus can be taken into account in the calculations. The results can be converted into the ENDF-VI format using the accompanying code EMPEND. The package contains the full EXFOR library of experimental data. Relevant EXFOR entries are automatically retrieved during the calculations. Plots comparing experimental results with the calculated ones can be produced using X4TOC4 and PLOTC4 codes linked to the rest of the system through bash-shell (UNIX) scripts. The graphic user interface written in Tcl/Tk is provided. (author)

Verification of reactor safety codes

International Nuclear Information System (INIS)

Murley, T.E.

1978-01-01

The safety evaluation of nuclear power plants requires the investigation of wide range of potential accidents that could be postulated to occur. Many of these accidents deal with phenomena that are outside the range of normal engineering experience. Because of the expense and difficulty of full scale tests covering the complete range of accident conditions, it is necessary to rely on complex computer codes to assess these accidents. The central role that computer codes play in safety analyses requires that the codes be verified, or tested, by comparing the code predictions with a wide range of experimental data chosen to span the physical phenomena expected under potential accident conditions. This paper discusses the plans of the Nuclear Regulatory Commission for verifying the reactor safety codes being developed by NRC to assess the safety of light water reactors and fast breeder reactors. (author)

Evolution of cichlid vision via trans-regulatory divergence

Directory of Open Access Journals (Sweden)

O’Quin Kelly E

2012-12-01

Full Text Available Abstract Background Phenotypic evolution may occur through mutations that affect either the structure or expression of protein-coding genes. Although the evolution of color vision has historically been attributed to structural mutations within the opsin genes, recent research has shown that opsin regulatory mutations can also tune photoreceptor sensitivity and color vision. Visual sensitivity in African cichlid fishes varies as a result of the differential expression of seven opsin genes. We crossed cichlid species that express different opsin gene sets and scanned their genome for expression Quantitative Trait Loci (eQTL responsible for these differences. Our results shed light on the role that different structural, cis-, and trans-regulatory mutations play in the evolution of color vision. Results We identified 11 eQTL that contribute to the divergent expression of five opsin genes. On three linkage groups, several eQTL formed regulatory “hotspots” associated with the expression of multiple opsins. Importantly, however, the majority of the eQTL we identified (8/11 or 73% occur on linkage groups located trans to the opsin genes, suggesting that cichlid color vision has evolved primarily via trans-regulatory divergence. By modeling the impact of just two of these trans-regulatory eQTL, we show that opsin regulatory mutations can alter cichlid photoreceptor sensitivity and color vision at least as much as opsin structural mutations can. Conclusions Combined with previous work, we demonstrate that the evolution of cichlid color vision results from the interplay of structural, cis-, and especially trans-regulatory loci. Although there are numerous examples of structural and cis-regulatory mutations that contribute to phenotypic evolution, our results suggest that trans-regulatory mutations could contribute to phenotypic divergence more commonly than previously expected, especially in systems like color vision, where compensatory changes in the

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

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

Gamma spectroscopy modelization intercomparison of the modelization results using two different codes (MCNP, and Pascalys-mercure)

International Nuclear Information System (INIS)

Luneville, L.; Chiron, M.; Toubon, H.; Dogny, S.; Huver, M.; Berger, L.

2001-01-01

The research performed in common these last 3 years by the French Atomic Commission CEA, COGEMA and Eurisys Mesures had for main subject the realization of a complete tool of modelization for the largest range of realistic cases, the Pascalys modelization software. The main purpose of the modelization was to calculate the global measurement efficiency, which delivers the most accurate relationship between the photons emitted by the nuclear source in volume, punctual or deposited form and the germanium hyper pure detector, which detects and analyzes the received photons. It has been stated since long time that experimental global measurement efficiency becomes more and more difficult to address especially for complex scene as we can find in decommissioning and dismantling or in case of high activities for which the use of high activity reference sources become difficult to use for both health physics point of view and regulations. The choice of a calculation code is fundamental if accurate modelization is searched. MCNP represents the reference code but its use is long time calculation consuming and then not practicable in line on the field. Direct line-of-sight point kernel code as the French Atomic Commission 3-D analysis Mercure code can represent the practicable compromise between the most accurate MCNP reference code and the realistic performances needed in modelization. The comparison between the results of Pascalys-Mercure and MCNP code taking in account the last improvements of Mercure in the low energy range where the most important errors can occur, is presented in this paper, Mercure code being supported in line by the recent Pascalys 3-D modelization scene software. The incidence of the intrinsic efficiency of the Germanium detector is also approached for the total efficiency of measurement. (authors)

Regulatory systems-based licensing guidance documentation

International Nuclear Information System (INIS)

Delligatti, M.S.

1991-01-01

The US Nuclear Regulatory Commission (NRC) has developed a series of licensing guidance documents based on the regulatory requirements in Part 60 of Title 10 of the Code of Federal Regulations (10 CFR Part 60). This regulatory systems-based approach to licensing guidance documentation relies on the definition of the high-level waste repository in 10 CFR Part 60. A document which is important for the frame-work it gives to other programmatic licensing guidance is the Draft Regulatory Guide open-quotes Format and Content for the License Application for the High-Level Waste Repositoryclose quotes (FCRG). The FCRG describes a format and content acceptable to NRC for a high-level waste repository license application pursuant to the requirements of 10 CFR Part 60. Other licensing guidance documents will be compatible with the FCRG

Annual Report 2013. Nuclear Regulatory Authority

International Nuclear Information System (INIS)

2010-01-01

The present Annual Report of Activities of the Nuclear Regulatory Authority (ARN), prepared regularly from the creation as independent institution, describes across seven parts and eight annexes the activities developed by the organism during 2013. The main topic are: the organization and the activity of the ARN; the regulatory standards; the licensing and inspection of nuclear power plants and critical facilities; the emergency systems; the environmental monitoring; the occupational surveillance; the training and the public information; improved organizational and budgetary developments. Also, this publication has annexes with the following content: regulatory documents; inspections to medical; presentations of publications from ARN staff; measurement and evaluation of the drinking water of Ezeiza; international expert report on the implementation of international standards on radiation protection in the Ezeiza Atomic Center; Code of Ethics of the Nuclear Regulatory Authority.

An evaluation model for the definition of regulatory requirements on spent fuel pool cooling systems

International Nuclear Information System (INIS)

Izquierdo, J.M.

1979-01-01

A calculation model is presented for establishing regulatory requirements in the SFPCS System. The major design factors, regulatory and design limits and key parameters are discussed. A regulatory position for internal use is proposed. Finally, associated problems and experience are presented. (author)

Experimental data bases useful for quantification of model uncertainties in best estimate codes

International Nuclear Information System (INIS)

Wilson, G.E.; Katsma, K.R.; Jacobson, J.L.; Boodry, K.S.

1988-01-01

A data base is necessary for assessment of thermal hydraulic codes within the context of the new NRC ECCS Rule. Separate effect tests examine particular phenomena that may be used to develop and/or verify models and constitutive relationships in the code. Integral tests are used to demonstrate the capability of codes to model global characteristics and sequence of events for real or hypothetical transients. The nuclear industry has developed a large experimental data base of fundamental nuclear, thermal-hydraulic phenomena for code validation. Given a particular scenario, and recognizing the scenario's important phenomena, selected information from this data base may be used to demonstrate applicability of a particular code to simulate the scenario and to determine code model uncertainties. LBLOCA experimental data bases useful to this objective are identified in this paper. 2 tabs

Nodal kinetics model upgrade in the Penn State coupled TRAC/NEM codes

International Nuclear Information System (INIS)

Beam, Tara M.; Ivanov, Kostadin N.; Baratta, Anthony J.; Finnemann, Herbert

1999-01-01

The Pennsylvania State University currently maintains and does development and verification work for its own versions of the coupled three-dimensional kinetics/thermal-hydraulics codes TRAC-PF1/NEM and TRAC-BF1/NEM. The subject of this paper is nodal model enhancements in the above mentioned codes. Because of the numerous validation studies that have been performed on almost every aspect of these codes, this upgrade is done without a major code rewrite. The upgrade consists of four steps. The first two steps are designed to improve the accuracy of the kinetics model, based on the nodal expansion method. The polynomial expansion solution of 1D transverse integrated diffusion equation is replaced with a solution, which uses a semi-analytic expansion. Further the standard parabolic polynomial representation of the transverse leakage in the above 1D equations is replaced with an improved approximation. The last two steps of the upgrade address the code efficiency by improving the solution of the time-dependent NEM equations and implementing a multi-grid solver. These four improvements are implemented into the standalone NEM kinetics code. Verification of this code was accomplished based on the original verification studies. The results show that the new methods improve the accuracy and efficiency of the code. The verification of the upgraded NEM model in the TRAC-PF1/NEM and TRAC-BF1/NEM coupled codes is underway

Applicability evaluation on the conservative metal-water reaction(MWR) model implemented into the SPACE code

International Nuclear Information System (INIS)

Lee, Suk Ho; You, Sung Chang; Kim, Han Gon

2011-01-01

The SBLOCA (Small Break Loss-of-Coolant Accident) evaluation methodology for the APR1400 (Advanced Power Reactor 1400) is under development using the SPACE code. The goal of the development of this methodology is to set up a conservative evaluation methodology in accordance with Appendix K of 10CFR50 by the end of 2012. In order to develop the Appendix K version of the SPACE code, the code modification is considered through implementation of the code on the required evaluation models. For the conservative models required in the SPACE code, the metal-water reaction (MWR) model, the critical flow model, the Critical Heat Flux (CHF) model and the post-CHF model must be implemented in the code. At present, the integration of the model to generate the Appendix K version of SPACE is in its preliminary stage. Among them, the conservative MWR model and its code applicability are introduced in this paper

In silico modeling of epigenetic-induced changes in photoreceptor cis-regulatory elements.

Science.gov (United States)

Hossain, Reafa A; Dunham, Nicholas R; Enke, Raymond A; Berndsen, Christopher E

2018-01-01

DNA methylation is a well-characterized epigenetic repressor of mRNA transcription in many plant and vertebrate systems. However, the mechanism of this repression is not fully understood. The process of transcription is controlled by proteins that regulate recruitment and activity of RNA polymerase by binding to specific cis-regulatory sequences. Cone-rod homeobox (CRX) is a well-characterized mammalian transcription factor that controls photoreceptor cell-specific gene expression. Although much is known about the functions and DNA binding specificity of CRX, little is known about how DNA methylation modulates CRX binding affinity to genomic cis-regulatory elements. We used bisulfite pyrosequencing of human ocular tissues to measure DNA methylation levels of the regulatory regions of RHO , PDE6B, PAX6 , and LINE1 retrotransposon repeats. To describe the molecular mechanism of repression, we used molecular modeling to illustrate the effect of DNA methylation on human RHO regulatory sequences. In this study, we demonstrate an inverse correlation between DNA methylation in regulatory regions adjacent to the human RHO and PDE6B genes and their subsequent transcription in human ocular tissues. Docking of CRX to the DNA models shows that CRX interacts with the grooves of these sequences, suggesting changes in groove structure could regulate binding. Molecular dynamics simulations of the RHO promoter and enhancer regions show changes in the flexibility and groove width upon epigenetic modification. Models also demonstrate changes in the local dynamics of CRX binding sites within RHO regulatory sequences which may account for the repression of CRX-dependent transcription. Collectively, these data demonstrate epigenetic regulation of CRX binding sites in human retinal tissue and provide insight into the mechanism of this mode of epigenetic regulation to be tested in future experiments.

Light water reactor fuel analysis code FEMAXI-7; model and structure

International Nuclear Information System (INIS)

Suzuki, Motoe; Udagawa, Yutaka; Saitou, Hiroaki

2011-03-01

A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in both normal conditions and anticipated transient conditions. This code is an advanced version which has been produced by incorporating the former version FEMAXI-6 with numerous functional improvements and extensions. In FEMAXI-7, many new models have been added and parameters have been clearly arranged. Also, to facilitate effective maintenance and accessibility of the code, modularization of subroutines and functions have been attained, and quality comment descriptions of variables or physical quantities have been incorporated in the source code. With these advancements, the FEMAXI-7 code has been upgraded to a versatile analytical tool for high burnup fuel behavior analyses. This report describes in detail the design, basic theory and structure, models and numerical method, and improvements and extensions. (author)

NRC model simulations in support of the hydrologic code intercomparison study (HYDROCOIN): Level 1-code verification

International Nuclear Information System (INIS)

1988-03-01

HYDROCOIN is an international study for examining ground-water flow modeling strategies and their influence on safety assessments of geologic repositories for nuclear waste. This report summarizes only the combined NRC project temas' simulation efforts on the computer code bench-marking problems. The codes used to simulate thesee seven problems were SWIFT II, FEMWATER, UNSAT2M USGS-3D, AND TOUGH. In general, linear problems involving scalars such as hydraulic head were accurately simulated by both finite-difference and finite-element solution algorithms. Both types of codes produced accurate results even for complex geometrics such as intersecting fractures. Difficulties were encountered in solving problems that invovled nonlinear effects such as density-driven flow and unsaturated flow. In order to fully evaluate the accuracy of these codes, post-processing of results using paricle tracking algorithms and calculating fluxes were examined. This proved very valuable by uncovering disagreements among code results even through the hydraulic-head solutions had been in agreement. 9 refs., 111 figs., 6 tabs

DOEZOR2: a code to assess the radiological impact of effluent discharges

International Nuclear Information System (INIS)

Martin Garcia, J.E.; Gomez Rodriguez, C.A.; Gimeno Blesa, M.E.; GARCIA ACOSTA, F.

2010-01-01

DOEZOR2 (DOsis al Exterior en ZORita v. 2) comprises a suite of models and data management tools which can be used to perform the radiological impact assessments of routine and continuous discharges from nuclear power plants and nuclear fuel cycle facilities for regulatory purposes. The Code is based on Radiation Protection 72 and the Regulatory Guide 1.109 (USNRC) methodology. The code development was carried out by SOCOIN (Gas Natural Fenosa Group) and is implemented in Jose Cabrera NPP (or Zorita NPP). DOEZOR has been operating during ten years and is about to be updated in order to consider realistic doses methodology. The new software differs from its predecessor in a number of ways, including a new user interface, new radionuclides and several new capacities (realistic dose). The model is developed to estimate radiological consequences of emissions from nuclear power plants. Internal exposure via inhalation and ingestion, external exposure from clouds and radioactivity deposited on the ground are included in the model. DOEZOR2 has been developed in Visual Fortran, using user friendly windows environment and modular architecture (easy to implement for other uses and installations). The main features of the code include: - Annual discharge to the atmospheric or river environment can be modelled. - Dynamic systems dispersion of radionuclides released to the river (two reservoirs). - A comprehensive list of exposure pathways. - A suite of environmental transfer models to estimate the transfer of radionuclides through the environment. - Site specific data (parameters of life habits). - Results in terms of individual doses (three age's groups), using effective dose as defined in Council Directive 96/29/EURATOM and dose coefficients from U.S. Federal Guidance Report 13. - Results in 'MS Word form' for periodic information to authorities. This software could be implemented anywhere, considering the particular characteristic of the site for the plant or facility

Radiation transport phenomena and modeling - part A: Codes

International Nuclear Information System (INIS)

Lorence, L.J.

1997-01-01

The need to understand how particle radiation (high-energy photons and electrons) from a variety of sources affects materials and electronics has motivated the development of sophisticated computer codes that describe how radiation with energies from 1.0 keV to 100.0 GeV propagates through matter. Predicting radiation transport is the necessary first step in predicting radiation effects. The radiation transport codes that are described here are general-purpose codes capable of analyzing a variety of radiation environments including those produced by nuclear weapons (x-rays, gamma rays, and neutrons), by sources in space (electrons and ions) and by accelerators (x-rays, gamma rays, and electrons). Applications of these codes include the study of radiation effects on electronics, nuclear medicine (imaging and cancer treatment), and industrial processes (food disinfestation, waste sterilization, manufacturing.) The primary focus will be on coupled electron-photon transport codes, with some brief discussion of proton transport. These codes model a radiation cascade in which electrons produce photons and vice versa. This coupling between particles of different types is important for radiation effects. For instance, in an x-ray environment, electrons are produced that drive the response in electronics. In an electron environment, dose due to bremsstrahlung photons can be significant once the source electrons have been stopped

Mechanistic modelling of gaseous fission product behaviour in UO2 fuel by Rtop code

International Nuclear Information System (INIS)

Kanukova, V.D.; Khoruzhii, O.V.; Kourtchatov, S.Y.; Likhanskii, V.V.; Matveew, L.V.

2002-01-01

The current status of a mechanistic modelling by the RTOP code of the fission product behaviour in polycrystalline UO 2 fuel is described. An outline of the code and implemented physical models is presented. The general approach to code validation is discussed. It is exemplified by the results of validation of the models of fuel oxidation and grain growth. The different models of intragranular and intergranular gas bubble behaviour have been tested and the sensitivity of the code in the framework of these models has been analysed. An analysis of available models of the resolution of grain face bubbles is also presented. The possibilities of the RTOP code are presented through the example of modelling behaviour of WWER fuel over the course of a comparative WWER-PWR experiment performed at Halden and by comparison with Yanagisawa experiments. (author)

Regulatory focus at work : the moderating role of regulatory focus in the job demands-resources model

NARCIS (Netherlands)

Brenninkmeijer, V.; Demerouti, E.; Blanc, Le P.M.; Emmerik, van I.J.H.

2010-01-01

Purpose – The purpose of this study is to examine the moderating role of regulatory focus in the job demands-resources model. Design/methodology/approach – A questionnaire survey was conducted among 146 teachers in secondary education. It was expected that detrimental effects of job demands (i.e.

GEMSFITS: Code package for optimization of geochemical model parameters and inverse modeling

International Nuclear Information System (INIS)

Miron, George D.; Kulik, Dmitrii A.; Dmytrieva, Svitlana V.; Wagner, Thomas

2015-01-01

Highlights: • Tool for generating consistent parameters against various types of experiments. • Handles a large number of experimental data and parameters (is parallelized). • Has a graphical interface and can perform statistical analysis on the parameters. • Tested on fitting the standard state Gibbs free energies of aqueous Al species. • Example on fitting interaction parameters of mixing models and thermobarometry. - Abstract: GEMSFITS is a new code package for fitting internally consistent input parameters of GEM (Gibbs Energy Minimization) geochemical–thermodynamic models against various types of experimental or geochemical data, and for performing inverse modeling tasks. It consists of the gemsfit2 (parameter optimizer) and gfshell2 (graphical user interface) programs both accessing a NoSQL database, all developed with flexibility, generality, efficiency, and user friendliness in mind. The parameter optimizer gemsfit2 includes the GEMS3K chemical speciation solver ( (http://gems.web.psi.ch/GEMS3K)), which features a comprehensive suite of non-ideal activity- and equation-of-state models of solution phases (aqueous electrolyte, gas and fluid mixtures, solid solutions, (ad)sorption. The gemsfit2 code uses the robust open-source NLopt library for parameter fitting, which provides a selection between several nonlinear optimization algorithms (global, local, gradient-based), and supports large-scale parallelization. The gemsfit2 code can also perform comprehensive statistical analysis of the fitted parameters (basic statistics, sensitivity, Monte Carlo confidence intervals), thus supporting the user with powerful tools for evaluating the quality of the fits and the physical significance of the model parameters. The gfshell2 code provides menu-driven setup of optimization options (data selection, properties to fit and their constraints, measured properties to compare with computed counterparts, and statistics). The practical utility, efficiency, and

Merits and difficulties in adopting codes, standards and nuclear regulations

International Nuclear Information System (INIS)

El-Saiedi, A.F.; Morsy, S.; Mariy, A.

1978-01-01

Developing countries planning for introducing nuclear power plants as a source of energy have to develop or adopt sound regulatory practices. These are necessary to help governmental authorities to assess the safety of nuclear power plants and to perform inspections needed to confirm the established safe and sound limits. The first requirement is to form an independent regulatory body capable of setting up and enforcing proper safety regulations. The formation of this body is governed by several considerations related to local conditions in the developing countries, which may not always be favourable. It is quite impractical for countries with limited experience in the nuclear power field to develop their own codes, standards and regulations required for the nuclear regulatory body to perform its tasks. A practical way is to adopt codes, standards and regulations of a well-developed country. This has merits as well as drawbacks. The latter are related to problems of personnel, software, equipment and facilities. The difficulties involved in forming a nuclear regulatory body, and the merits and difficulties in adopting foreign codes, standards and regulations required for such body to perform its tasks, are discussed in this paper. Discussions are applicable to many developing countries and particular emphasis is given to the conditions and practices in Egypt. (author)

Regulatory Audit Activities on Nuclear Design of Reactor Cores

International Nuclear Information System (INIS)

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

2016-01-01

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

Regulatory Audit Activities on Nuclear Design of Reactor Cores

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Simulation of loss of feedwater transient of MASLWR test facility by MARS-KS code

Energy Technology Data Exchange (ETDEWEB)

Park, Juyeop [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-05-15

MASLWR test facility is a mock-up of a passive integral type reactor equipped with helical coil steam generator. Since SMART reactor which is being current developed domestically also adopts helical coil steam generator, KINS has joined this ICSP to evaluate performance of domestic regulatory audit thermal-hydraulic code (MARS-KS code) in various respects including wall-to-fluid heat transfer model modification implemented in the code by independent international experiment database. In the ICSP, two types of transient experiments have been focused and they are loss of feedwater transient with subsequent ADS operation and long term cooling (SP-2) and normal operating conditions at different power levels (SP-3). In the present study, KINS simulation results by the MARS-KS code (KS-002 version) for the SP-2 experiment are presented in detail and conclusions on MARS-KS code performance drawn through this simulation is described. Performance of the MARS-KS code is evaluated through the simulation of the loss of feedwater transient of the MASLWR test facility. Steady state run shows helical coil specific heat transfer models implemented in the code is reasonable. However, through the transient run, it is also found that three-dimensional effect within the HPC and axial conduction effect through the HTP are not well reproduced by the code.

A review on the CIRCE methodology to quantify the uncertainty of the physical models of a code

International Nuclear Information System (INIS)

Jeon, Seong Su; Hong, Soon Joon; Bang, Young Seok

2012-01-01

In the field of nuclear engineering, recent regulatory audit calculations of large break loss of coolant accident (LBLOCA) have been performed with the best estimate code such as MARS, RELAP5 and CATHARE. Since the credible regulatory audit calculation is very important in the evaluation of the safety of the nuclear power plant (NPP), there have been many researches to develop rules and methodologies for the use of best estimate codes. One of the major points is to develop the best estimate plus uncertainty (BEPU) method for uncertainty analysis. As a representative BEPU method, NRC proposes the CSAU (Code scaling, applicability and uncertainty) methodology, which clearly identifies the different steps necessary for an uncertainty analysis. The general idea is 1) to determine all the sources of uncertainty in the code, also called basic uncertainties, 2) quantify them and 3) combine them in order to obtain the final uncertainty for the studied application. Using the uncertainty analysis such as CSAU methodology, an uncertainty band for the code response (calculation result), important from the safety point of view is calculated and the safety margin of the NPP is quantified. An example of such a response is the peak cladding temperature (PCT) for a LBLOCA. However, there is a problem in the uncertainty analysis with the best estimate codes. Generally, it is very difficult to determine the uncertainties due to the empiricism of closure laws (also called correlations or constitutive relationships). So far the only proposed approach is based on the expert judgment. For this case, the uncertainty range of important parameters can be wide and inaccurate so that the confidence level of the BEPU calculation results can be decreased. In order to solve this problem, recently CEA (France) proposes a statistical method of data analysis, called CIRCE. The CIRCE method is intended to quantify the uncertainties of the correlations of a code. It may replace the expert judgment

Evaluation of Advanced Models for PAFS Condensation Heat Transfer in SPACE Code

Energy Technology Data Exchange (ETDEWEB)

Bae, Byoung-Uhn; Kim, Seok; Park, Yu-Sun; Kang, Kyung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ahn, Tae-Hwan; Yun, Byong-Jo [Pusan National University, Busan (Korea, Republic of)

2015-10-15

The PAFS (Passive Auxiliary Feedwater System) is operated by the natural circulation to remove the core decay heat through the PCHX (Passive Condensation Heat Exchanger) which is composed of the nearly horizontal tubes. For validation of the cooling and operational performance of the PAFS, PASCAL (PAFS Condensing Heat Removal Assessment Loop) facility was constructed and the condensation heat transfer and natural convection phenomena in the PAFS was experimentally investigated at KAERI (Korea Atomic Energy Research Institute). From the PASCAL experimental result, it was found that conventional system analysis code underestimated the condensation heat transfer. In this study, advanced condensation heat transfer models which can treat the heat transfer mechanisms with the different flow regimes in the nearly horizontal heat exchanger tube were analyzed. The models were implemented in a thermal hydraulic safety analysis code, SPACE (Safety and Performance Analysis Code for Nuclear Power Plant), and it was evaluated with the PASCAL experimental data. With an aim of enhancing the prediction capability for the condensation phenomenon inside the PCHX tube of the PAFS, advanced models for the condensation heat transfer were implemented into the wall condensation model of the SPACE code, so that the PASCAL experimental result was utilized to validate the condensation models. Calculation results showed that the improved model for the condensation heat transfer coefficient enhanced the prediction capability of the SPACE code. This result confirms that the mechanistic modeling for the film condensation in the steam phase and the convection in the condensate liquid contributed to enhance the prediction capability of the wall condensation model of the SPACE code and reduce conservatism in prediction of condensation heat transfer.

Task-based dermal exposure models for regulatory risk assessment

NARCIS (Netherlands)

Warren, N.D.; Marquart, H.; Christopher, Y.; Laitinen, J.; Hemmen, J.J. van

2006-01-01

The regulatory risk assessment of chemicals requires the estimation of occupational dermal exposure. Until recently, the models used were either based on limited data or were specific to a particular class of chemical or application. The EU project RISKOFDERM has gathered a considerable number of

Light water reactor fuel analysis code FEMAXI-7. Model and structure

International Nuclear Information System (INIS)

Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa; Saitou, Hiroaki

2013-07-01

A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in both normal conditions and anticipated transient conditions. This code is an advanced version which has been produced by incorporating the former version FEMAXI-6 with numerous functional improvements and extensions. In FEMAXI-7, many new models have been added and parameters have been clearly arranged. Also, to facilitate effective maintenance and accessibility of the code, modularization of subroutines and functions have been attained, and quality comment descriptions of variables or physical quantities have been incorporated in the source code. With these advancements, the FEMAXI-7 code has been upgraded to a versatile analytical tool for high burnup fuel behavior analyses. This report describes in detail the design, basic theory and structure, models and numerical method of FEMAXI-7, and its improvements and extensions. (author)

Recent improvements of the TNG statistical model code

International Nuclear Information System (INIS)

Shibata, K.; Fu, C.Y.

1986-08-01

The applicability of the nuclear model code TNG to cross-section evaluations has been extended. The new TNG is capable of using variable bins for outgoing particle energies. Moreover, three additional quantities can now be calculated: capture gamma-ray spectrum, the precompound mode of the (n,γ) reaction, and fission cross section. In this report, the new features of the code are described together with some sample calculations and a brief explanation of the input data. 15 refs., 6 figs., 2 tabs

Benchmark Simulation for the Development of the Regulatory Audit Subchannel Analysis Code

Energy Technology Data Exchange (ETDEWEB)

Lee, G. H.; Song, C.; Woo, S. W. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2012-05-15

For the safe and reliable operation of a reactor, it is important to predict accurately the flow and temperature distributions in the thermal-hydraulic design of a reactor core. A subchannel approach can give the reasonable flow and temperature distributions with the short computing time. Korea Institute of Nuclear Safety (KINS) is presently reviewing new subchannel code, THALES, which will substitute for both THINC-IV and TORC code. To assess the prediction performance of THALES, KINS is developing the subchannel analysis code for the independent audit calculation. The code is based on workstation version of COBRA-IV-I. The main objective of the present study is to assess the performance of COBRA-IV-I code by comparing the simulation results with experimental ones for the sample problems

Modelling of the RA-1 reactor using a Monte Carlo code

International Nuclear Information System (INIS)

Quinteiro, Guillermo F.; Calabrese, Carlos R.

2000-01-01

It was carried out for the first time, a model of the Argentine RA-1 reactor using the MCNP Monte Carlo code. This model was validated using data for experimental neutron and gamma measurements at different energy ranges and locations. In addition, the resulting fluxes were compared with the data obtained using a 3D diffusion code. (author)

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

Background-Modeling-Based Adaptive Prediction for Surveillance Video Coding.

Science.gov (United States)

Zhang, Xianguo; Huang, Tiejun; Tian, Yonghong; Gao, Wen

2014-02-01

The exponential growth of surveillance videos presents an unprecedented challenge for high-efficiency surveillance video coding technology. Compared with the existing coding standards that were basically developed for generic videos, surveillance video coding should be designed to make the best use of the special characteristics of surveillance videos (e.g., relative static background). To do so, this paper first conducts two analyses on how to improve the background and foreground prediction efficiencies in surveillance video coding. Following the analysis results, we propose a background-modeling-based adaptive prediction (BMAP) method. In this method, all blocks to be encoded are firstly classified into three categories. Then, according to the category of each block, two novel inter predictions are selectively utilized, namely, the background reference prediction (BRP) that uses the background modeled from the original input frames as the long-term reference and the background difference prediction (BDP) that predicts the current data in the background difference domain. For background blocks, the BRP can effectively improve the prediction efficiency using the higher quality background as the reference; whereas for foreground-background-hybrid blocks, the BDP can provide a better reference after subtracting its background pixels. Experimental results show that the BMAP can achieve at least twice the compression ratio on surveillance videos as AVC (MPEG-4 Advanced Video Coding) high profile, yet with a slightly additional encoding complexity. Moreover, for the foreground coding performance, which is crucial to the subjective quality of moving objects in surveillance videos, BMAP also obtains remarkable gains over several state-of-the-art methods.

International Code Assessment and Applications Program: Summary of code assessment studies concerning RELAP5/MOD2, RELAP5/MOD3, and TRAC-B

International Nuclear Information System (INIS)

Schultz, R.R.

1993-12-01

Members of the International Code Assessment Program (ICAP) have assessed the US Nuclear Regulatory Commission (USNRC) advanced thermal-hydraulic codes over the past few years in a concerted effort to identify deficiencies, to define user guidelines, and to determine the state of each code. The results of sixty-two code assessment reviews, conducted at INEL, are summarized. Code deficiencies are discussed and user recommended nodalizations investigated during the course of conducting the assessment studies and reviews are listed. All the work that is summarized was done using the RELAP5/MOD2, RELAP5/MOD3, and TRAC-B codes

Prevalence of transcription promoters within archaeal operons and coding sequences.

Science.gov (United States)

Koide, Tie; Reiss, David J; Bare, J Christopher; Pang, Wyming Lee; Facciotti, Marc T; Schmid, Amy K; Pan, Min; Marzolf, Bruz; Van, Phu T; Lo, Fang-Yin; Pratap, Abhishek; Deutsch, Eric W; Peterson, Amelia; Martin, Dan; Baliga, Nitin S

2009-01-01

Despite the knowledge of complex prokaryotic-transcription mechanisms, generalized rules, such as the simplified organization of genes into operons with well-defined promoters and terminators, have had a significant role in systems analysis of regulatory logic in both bacteria and archaea. Here, we have investigated the prevalence of alternate regulatory mechanisms through genome-wide characterization of transcript structures of approximately 64% of all genes, including putative non-coding RNAs in Halobacterium salinarum NRC-1. Our integrative analysis of transcriptome dynamics and protein-DNA interaction data sets showed widespread environment-dependent modulation of operon architectures, transcription initiation and termination inside coding sequences, and extensive overlap in 3' ends of transcripts for many convergently transcribed genes. A significant fraction of these alternate transcriptional events correlate to binding locations of 11 transcription factors and regulators (TFs) inside operons and annotated genes-events usually considered spurious or non-functional. Using experimental validation, we illustrate the prevalence of overlapping genomic signals in archaeal transcription, casting doubt on the general perception of rigid boundaries between coding sequences and regulatory elements.

Coupling Legacy and Contemporary Deterministic Codes to Goldsim for Probabilistic Assessments of Potential Low-Level Waste Repository Sites

Science.gov (United States)

Mattie, P. D.; Knowlton, R. G.; Arnold, B. W.; Tien, N.; Kuo, M.

2006-12-01

Sandia National Laboratories (Sandia), a U.S. Department of Energy National Laboratory, has over 30 years experience in radioactive waste disposal and is providing assistance internationally in a number of areas relevant to the safety assessment of radioactive waste disposal systems. International technology transfer efforts are often hampered by small budgets, time schedule constraints, and a lack of experienced personnel in countries with small radioactive waste disposal programs. In an effort to surmount these difficulties, Sandia has developed a system that utilizes a combination of commercially available codes and existing legacy codes for probabilistic safety assessment modeling that facilitates the technology transfer and maximizes limited available funding. Numerous codes developed and endorsed by the United States Nuclear Regulatory Commission and codes developed and maintained by United States Department of Energy are generally available to foreign countries after addressing import/export control and copyright requirements. From a programmatic view, it is easier to utilize existing codes than to develop new codes. From an economic perspective, it is not possible for most countries with small radioactive waste disposal programs to maintain complex software, which meets the rigors of both domestic regulatory requirements and international peer review. Therefore, re-vitalization of deterministic legacy codes, as well as an adaptation of contemporary deterministic codes, provides a creditable and solid computational platform for constructing probabilistic safety assessment models. External model linkage capabilities in Goldsim and the techniques applied to facilitate this process will be presented using example applications, including Breach, Leach, and Transport-Multiple Species (BLT-MS), a U.S. NRC sponsored code simulating release and transport of contaminants from a subsurface low-level waste disposal facility used in a cooperative technology transfer

Verification and Validation of Heat Transfer Model of AGREE Code

Energy Technology Data Exchange (ETDEWEB)

Tak, N. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Seker, V.; Drzewiecki, T. J.; Downar, T. J. [Department of Nuclear Engineering and Radiological Sciences, Univ. of Michigan, Michigan (United States); Kelly, J. M. [US Nuclear Regulatory Commission, Washington (United States)

2013-05-15

The AGREE code was originally developed as a multi physics simulation code to perform design and safety analysis of Pebble Bed Reactors (PBR). Currently, additional capability for the analysis of Prismatic Modular Reactor (PMR) core is in progress. Newly implemented fluid model for a PMR core is based on a subchannel approach which has been widely used in the analyses of light water reactor (LWR) cores. A hexagonal fuel (or graphite block) is discretized into triangular prism nodes having effective conductivities. Then, a meso-scale heat transfer model is applied to the unit cell geometry of a prismatic fuel block. Both unit cell geometries of multi-hole and pin-in-hole types of prismatic fuel blocks are considered in AGREE. The main objective of this work is to verify and validate the heat transfer model newly implemented for a PMR core in the AGREE code. The measured data in the HENDEL experiment were used for the validation of the heat transfer model for a pin-in-hole fuel block. However, the HENDEL tests were limited to only steady-state conditions of pin-in-hole fuel blocks. There exist no available experimental data regarding a heat transfer in multi-hole fuel blocks. Therefore, numerical benchmarks using conceptual problems are considered to verify the heat transfer model of AGREE for multi-hole fuel blocks as well as transient conditions. The CORONA and GAMMA+ codes were used to compare the numerical results. In this work, the verification and validation study were performed for the heat transfer model of the AGREE code using the HENDEL experiment and the numerical benchmarks of selected conceptual problems. The results of the present work show that the heat transfer model of AGREE is accurate and reliable for prismatic fuel blocks. Further validation of AGREE is in progress for a whole reactor problem using the HTTR safety test data such as control rod withdrawal tests and loss-of-forced convection tests.

An Eulerian transport-dispersion model of passive effluents: the Difeul code

International Nuclear Information System (INIS)

Wendum, D.

1994-11-01

R and D has decided to develop an Eulerian diffusion model easy to adapt to meteorological data coming from different sources: for instance the ARPEGE code of Meteo-France or the MERCURE code of EDF. We demand this in order to be able to apply the code in independent cases: a posteriori studies of accidental releases from nuclear power plants ar large or medium scale, simulation of urban pollution episodes within the ''Reactive Atmospheric Flows'' research project. For simplicity reasons, the numerical formulation of our code is the same as the one used in Meteo-France's MEDIA model. The numerical tests presented in this report show the good performance of those schemes. In order to illustrate the method by a concrete example a fictitious release from Saint-Laurent has been simulated at national scale: the results of this simulation agree quite well with those of the trajectory model DIFTRA. (author). 6 figs., 4 tabs

Balancing technical and regulatory concerns related to testing and control of performance assessment software

International Nuclear Information System (INIS)

Seitz, R.R.; Matthews, S.D.; Kostelnik, K.M.

1990-01-01

What activities are required to assure that a performance assessment (PA) computer code operates as it is intended? Answers to this question will vary depending on the individual's area of expertise. Different perspectives on testing and control of PA software are discussed based on interpretations of the testing and control process associated with the different involved parties. This discussion leads into the presentation of a general approach to software testing and control that address regulatory requirements. Finally, the need for balance between regulatory and scientific concerns is illustrated through lessons learned in previous implementations of software testing and control programs. Configuration control and software testing are required to provide assurance that a computer code performs as intended. Configuration control provides traceability and reproducibility of results produced with PA software and provides a system to assure that users have access to the current version of the software. Software testing is conducted to assure that the computer code has been written properly, solution techniques have been properly implemented, and the software is capable of representing the behavior of the specific system to be modeled. Comprehensive software testing includes: software analysis, verification testing, benchmark testing, and site-specific calibration/validation testing

Code Help: Can This Unique State Regulatory Intervention Improve Emergency Department Crowding?

Science.gov (United States)

Michael, Sean S; Broach, John P; Kotkowski, Kevin A; Brush, D Eric; Volturo, Gregory A; Reznek, Martin A

2018-05-01

Emergency department (ED) crowding adversely affects multiple facets of high-quality care. The Commonwealth of Massachusetts mandates specific, hospital action plans to reduce ED boarding via a mechanism termed "Code Help." Because implementation appears inconsistent even when hospital conditions should have triggered its activation, we hypothesized that compliance with the Code Help policy would be associated with reduction in ED boarding time and total ED length of stay (LOS) for admitted patients, compared to patients seen when the Code Help policy was not followed. This was a retrospective analysis of data collected from electronic, patient-care, timestamp events and from a prospective Code Help registry for consecutive adult patients admitted from the ED at a single academic center during a 15-month period. For each patient, we determined whether the concurrent hospital status complied with the Code Help policy or violated it at the time of admission decision. We then compared ED boarding time and overall ED LOS for patients cared for during periods of Code Help policy compliance and during periods of Code Help policy violation, both with reference to patients cared for during normal operations. Of 89,587 adult patients who presented to the ED during the study period, 24,017 (26.8%) were admitted to an acute care or critical care bed. Boarding time ranged from zero to 67 hours 30 minutes (median 4 hours 31 minutes). Total ED LOS for admitted patients ranged from 11 minutes to 85 hours 25 minutes (median nine hours). Patients admitted during periods of Code Help policy violation experienced significantly longer boarding times (median 20 minutes longer) and total ED LOS (median 46 minutes longer), compared to patients admitted under normal operations. However, patients admitted during Code Help policy compliance did not experience a significant increase in either metric, compared to normal operations. In this single-center experience, implementation of the

Development and Implementation of CFD-Informed Models for the Advanced Subchannel Code CTF

Science.gov (United States)

Blyth, Taylor S.

The research described in this PhD thesis contributes to the development of efficient methods for utilization of high-fidelity models and codes to inform low-fidelity models and codes in the area of nuclear reactor core thermal-hydraulics. The objective is to increase the accuracy of predictions of quantities of interests using high-fidelity CFD models while preserving the efficiency of low-fidelity subchannel core calculations. An original methodology named Physics-based Approach for High-to-Low Model Information has been further developed and tested. The overall physical phenomena and corresponding localized effects, which are introduced by the presence of spacer grids in light water reactor (LWR) cores, are dissected in corresponding four building basic processes, and corresponding models are informed using high-fidelity CFD codes. These models are a spacer grid-directed cross-flow model, a grid-enhanced turbulent mixing model, a heat transfer enhancement model, and a spacer grid pressure loss model. The localized CFD-models are developed and tested using the CFD code STAR-CCM+, and the corresponding global model development and testing in sub-channel formulation is performed in the thermal-hydraulic subchannel code CTF. The improved CTF simulations utilize data-files derived from CFD STAR-CCM+ simulation results covering the spacer grid design desired for inclusion in the CTF calculation. The current implementation of these models is examined and possibilities for improvement and further development are suggested. The validation experimental database is extended by including the OECD/NRC PSBT benchmark data. The outcome is an enhanced accuracy of CTF predictions while preserving the computational efficiency of a low-fidelity subchannel code.

Development and Implementation of CFD-Informed Models for the Advanced Subchannel Code CTF

Energy Technology Data Exchange (ETDEWEB)

Blyth, Taylor S. [Pennsylvania State Univ., University Park, PA (United States); Avramova, Maria [North Carolina State Univ., Raleigh, NC (United States)

2017-04-01

The research described in this PhD thesis contributes to the development of efficient methods for utilization of high-fidelity models and codes to inform low-fidelity models and codes in the area of nuclear reactor core thermal-hydraulics. The objective is to increase the accuracy of predictions of quantities of interests using high-fidelity CFD models while preserving the efficiency of low-fidelity subchannel core calculations. An original methodology named Physics- based Approach for High-to-Low Model Information has been further developed and tested. The overall physical phenomena and corresponding localized effects, which are introduced by the presence of spacer grids in light water reactor (LWR) cores, are dissected in corresponding four building basic processes, and corresponding models are informed using high-fidelity CFD codes. These models are a spacer grid-directed cross-flow model, a grid-enhanced turbulent mixing model, a heat transfer enhancement model, and a spacer grid pressure loss model. The localized CFD-models are developed and tested using the CFD code STAR-CCM+, and the corresponding global model development and testing in sub-channel formulation is performed in the thermal- hydraulic subchannel code CTF. The improved CTF simulations utilize data-files derived from CFD STAR-CCM+ simulation results covering the spacer grid design desired for inclusion in the CTF calculation. The current implementation of these models is examined and possibilities for improvement and further development are suggested. The validation experimental database is extended by including the OECD/NRC PSBT benchmark data. The outcome is an enhanced accuracy of CTF predictions while preserving the computational efficiency of a low-fidelity subchannel code.

The WARP Code: Modeling High Intensity Ion Beams

International Nuclear Information System (INIS)

Grote, David P.; Friedman, Alex; Vay, Jean-Luc; Haber, Irving

2005-01-01

The Warp code, developed for heavy-ion driven inertial fusion energy studies, is used to model high intensity ion (and electron) beams. Significant capability has been incorporated in Warp, allowing nearly all sections of an accelerator to be modeled, beginning with the source. Warp has as its core an explicit, three-dimensional, particle-in-cell model. Alongside this is a rich set of tools for describing the applied fields of the accelerator lattice, and embedded conducting surfaces (which are captured at sub-grid resolution). Also incorporated are models with reduced dimensionality: an axisymmetric model and a transverse ''slice'' model. The code takes advantage of modern programming techniques, including object orientation, parallelism, and scripting (via Python). It is at the forefront in the use of the computational technique of adaptive mesh refinement, which has been particularly successful in the area of diode and injector modeling, both steady-state and time-dependent. In the presentation, some of the major aspects of Warp will be overviewed, especially those that could be useful in modeling ECR sources. Warp has been benchmarked against both theory and experiment. Recent results will be presented showing good agreement of Warp with experimental results from the STS500 injector test stand

Once-through CANDU reactor models for the ORIGEN2 computer code

International Nuclear Information System (INIS)

Croff, A.G.; Bjerke, M.A.

1980-11-01

Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % 235 U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given

THE EXPERIENCE OF COMPARISON OF STATIC SECURITY CODE ANALYZERS

Directory of Open Access Journals (Sweden)

Alexey Markov

2015-09-01

Full Text Available This work presents a methodological approach to comparison of static security code analyzers. It substantiates the comparison of the static analyzers as to efficiency and functionality indicators, which are stipulated in the international regulatory documents. The test data for assessment of static analyzers efficiency is represented by synthetic sets of open-source software, which contain vulnerabilities. We substantiated certain criteria for quality assessment of the static security code analyzers subject to standards NIST SP 500-268 and SATEC. We carried out experiments that allowed us to assess a number of the Russian proprietary software tools and open-source tools. We came to the conclusion that it is of paramount importance to develop Russian regulatory framework for testing software security (firstly, for controlling undocumented features and evaluating the quality of static security code analyzers.

A review of MAAP4 code structure and core T/H model

International Nuclear Information System (INIS)

Song, Yong Mann; Park, Soo Yong

1998-03-01

The modular accident analysis program (MAAP) version 4 is a computer code that can simulate the response of LWR plants during severe accident sequences and includes models for all of the important phenomena which might occur during accident sequences. In this report, MAAP4 code structure and core thermal hydraulic (T/H) model which models the T/H behavior of the reactor core and the response of core components during all accident phases involving degraded cores are specifically reviewed and then reorganized. This reorganization is performed via getting the related models together under each topic whose contents and order are same with other two reports for MELCOR and SCDAP/RELAP5 to be simultaneously published. Major purpose of the report is to provide information about the characteristics of MAAP4 core T/H models for an integrated severe accident computer code development being performed under the one of on-going mid/long-term nuclear developing project. The basic characteristics of the new integrated severe accident code includes: 1) Flexible simulation capability of primary side, secondary side, and the containment under severe accident conditions, 2) Detailed plant simulation, 3) Convenient user-interfaces, 4) Highly modularization for easy maintenance/improvement, and 5) State-of-the-art model selection. In conclusion, MAAP4 code has appeared to be superior for 3) and 4) items but to be somewhat inferior for 1) and 2) items. For item 5), more efforts should be made in the future to compare separated models in detail with not only other codes but also recent world-wide work. (author). 17 refs., 1 tab., 12 figs

Code Generation for Protocols from CPN models Annotated with Pragmatics

DEFF Research Database (Denmark)

Simonsen, Kent Inge; Kristensen, Lars Michael; Kindler, Ekkart

software implementation satisfies the properties verified for the model. Coloured Petri Nets (CPNs) have been widely used to model and verify protocol software, but limited work exists on using CPN models of protocol software as a basis for automated code generation. In this report, we present an approach...... modelling languages, MDE further has the advantage that models are amenable to model checking which allows key behavioural properties of the software design to be verified. The combination of formally verified models and automated code generation contributes to a high degree of assurance that the resulting...... for generating protocol software from a restricted class of CPN models. The class of CPN models considered aims at being descriptive in that the models are intended to be helpful in understanding and conveying the operation of the protocol. At the same time, a descriptive model is close to a verifiable version...

International Code Assessment and Applications Program: Annual report

International Nuclear Information System (INIS)

Ting, P.; Hanson, R.; Jenks, R.

1987-03-01

This is the first annual report of the International Code Assessment and Applications Program (ICAP). The ICAP was organized by the Office of Nuclear Regulatory Research, United States Nuclear Regulatory Commission (USNRC) in 1985. The ICAP is an international cooperative reactor safety research program planned to continue over a period of approximately five years. To date, eleven European and Asian countries/organizations have joined the program through bilateral agreements with the USNRC. Seven proposed agreements are currently under negotiation. The primary mission of the ICAP is to provide independent assessment of the three major advanced computer codes (RELAP5, TRAC-PWR, and TRAC-BWR) developed by the USNRC. However, program activities can be expected to enhance the assessment process throughout member countries. The codes were developed to calculate the reactor plant response to transients and loss-of-coolant accidents. Accurate prediction of normal and abnormal plant response using the codes enhances procedures and regulations used for the safe operation of the plant and also provides technical basis for assessing the safety margin of future reactor plant designs. The ICAP is providing required assessment data that will contribute to quantification of the code uncertainty for each code. The first annual report is devoted to coverage of program activities and accomplishments during the period between April 1985 and March 1987

A Perceptual Model for Sinusoidal Audio Coding Based on Spectral Integration

Directory of Open Access Journals (Sweden)

Jensen Søren Holdt

2005-01-01

Full Text Available Psychoacoustical models have been used extensively within audio coding applications over the past decades. Recently, parametric coding techniques have been applied to general audio and this has created the need for a psychoacoustical model that is specifically suited for sinusoidal modelling of audio signals. In this paper, we present a new perceptual model that predicts masked thresholds for sinusoidal distortions. The model relies on signal detection theory and incorporates more recent insights about spectral and temporal integration in auditory masking. As a consequence, the model is able to predict the distortion detectability. In fact, the distortion detectability defines a (perceptually relevant norm on the underlying signal space which is beneficial for optimisation algorithms such as rate-distortion optimisation or linear predictive coding. We evaluate the merits of the model by combining it with a sinusoidal extraction method and compare the results with those obtained with the ISO MPEG-1 Layer I-II recommended model. Listening tests show a clear preference for the new model. More specifically, the model presented here leads to a reduction of more than 20% in terms of number of sinusoids needed to represent signals at a given quality level.

Steam generator and circulator model for the HELAP code

International Nuclear Information System (INIS)

Ludewig, H.

1975-07-01

An outline is presented of the work carried out in the 1974 fiscal year on the GCFBR safety research project consisting of the development of improved steam generator and circulator (steam turbine driven helium compressor) models which will eventually be inserted in the HELAP (1) code. Furthermore, a code was developed which will be used to generate steady state input for the primary and secondary sides of the steam generator. The following conclusions and suggestions for further work are made: (1) The steam-generator and circulator model are consistent with the volume and junction layout used in HELAP, (2) with minor changes these models, when incorporated in HELAP, could be used to simulate a direct cycle plant, (3) an explicit control valve model is still to be developed and would be very desirable to control the flow to the turbine during a transient (initially this flow will be controlled by using the existing check valve model); (4) the friction factor in the laminar flow region is computed inaccurately, this might cause significant errors in loss-of-flow accidents; and (5) it is felt that HELAP will still use a large amount of computer time and will thus be limited to design basis accidents without scram or loss of flow transients with and without scram. Finally it may also be used as a test bed for the development of prototype component models which would be incorporated in a more sophisticated system code, developed specifically for GCFBR's

PHITS code improvements by Regulatory Standard and Research Department Secretariat of Nuclear Regulation Authority

International Nuclear Information System (INIS)

Goko, Shinji

2017-01-01

As for the safety analysis to be carried out when a nuclear power company applies for installation permission of facility or equipment, business license, design approval etc., the Regulatory Standard and Research Department Secretariat of Nuclear Regulation Authority continuously conducts safety research for the introduction of various technologies and their improvement in order to evaluate the adequacy of this safety analysis. In the field of the shielding analysis of nuclear fuel transportation materials, this group improved the code to make PHITS applicable to this field, and has been promoting the improvement as a tool used for regulations since FY2013. This paper introduced the history and progress of this safety research. PHITS 2.88, which is the latest version as of November 2016, was equipped with the automatic generation function of variance reduction parameters [T-WWG] etc., and developed as the tool equipped with many effective functions in practical application to nuclear power regulations. In addition, this group conducted the verification analysis against nuclear fuel packages, which showed a good agreement with the analysis by MCNP, which is extensively used worldwide and abundant in actual results. It also shows a relatively good agreement with the measured values, when considering differences in analysis and measurement. (A.O.)

Film grain noise modeling in advanced video coding

Science.gov (United States)

Oh, Byung Tae; Kuo, C.-C. Jay; Sun, Shijun; Lei, Shawmin

2007-01-01

A new technique for film grain noise extraction, modeling and synthesis is proposed and applied to the coding of high definition video in this work. The film grain noise is viewed as a part of artistic presentation by people in the movie industry. On one hand, since the film grain noise can boost the natural appearance of pictures in high definition video, it should be preserved in high-fidelity video processing systems. On the other hand, video coding with film grain noise is expensive. It is desirable to extract film grain noise from the input video as a pre-processing step at the encoder and re-synthesize the film grain noise and add it back to the decoded video as a post-processing step at the decoder. Under this framework, the coding gain of the denoised video is higher while the quality of the final reconstructed video can still be well preserved. Following this idea, we present a method to remove film grain noise from image/video without distorting its original content. Besides, we describe a parametric model containing a small set of parameters to represent the extracted film grain noise. The proposed model generates the film grain noise that is close to the real one in terms of power spectral density and cross-channel spectral correlation. Experimental results are shown to demonstrate the efficiency of the proposed scheme.

Coding conventions and principles for a National Land-Change Modeling Framework

Science.gov (United States)

Donato, David I.

2017-07-14

This report establishes specific rules for writing computer source code for use with the National Land-Change Modeling Framework (NLCMF). These specific rules consist of conventions and principles for writing code primarily in the C and C++ programming languages. Collectively, these coding conventions and coding principles create an NLCMF programming style. In addition to detailed naming conventions, this report provides general coding conventions and principles intended to facilitate the development of high-performance software implemented with code that is extensible, flexible, and interoperable. Conventions for developing modular code are explained in general terms and also enabled and demonstrated through the appended templates for C++ base source-code and header files. The NLCMF limited-extern approach to module structure, code inclusion, and cross-module access to data is both explained in the text and then illustrated through the module templates. Advice on the use of global variables is provided.

Evaluation of Urban air quality models for regulatory use: Refinement of an approach

International Nuclear Information System (INIS)

Downton, M.W.; Dennis, R.L.

1985-01-01

Statistical measures for evaluating the performance of urban air quality models have recently been strongly recommended by several investigators. Problems that were encountered in the use of recommended performance measures in an evaluation of three versions of an urban photochemical model are described. The example demonstrates the importance of designing an evaluation to take into account the way in which the model will be used in regulatory practice, and then choosing performance measures on the basis of that design. The evaluation illustrates some limitations and possible pitfalls in the use and interpretation of statistical measures of model performance. Drawing on this experience, a procedure for evaluation of air quality models for regulatory use is suggested

Implementation of 3D models in the Monte Carlo code MCNP

International Nuclear Information System (INIS)

Lopes, Vivaldo; Millian, Felix M.; Guevara, Maria Victoria M.; Garcia, Fermin; Sena, Isaac; Menezes, Hugo

2009-01-01

On the area of numerical dosimetry Applied to medical physics, the scientific community focuses on the elaboration of new hybrids models based on 3D models. But different steps of the process of simulation with 3D models needed improvement and optimization in order to expedite the calculations and accuracy using this methodology. This project was developed with the aim of optimize the process of introduction of 3D models within the simulation code of radiation transport by Monte Carlo (MCNP). The fast implementation of these models on the simulation code allows the estimation of the dose deposited on the patient organs on a more personalized way, increasing the accuracy with this on the estimates and reducing the risks to health, caused by ionizing radiations. The introduction o these models within the MCNP was made through a input file, that was constructed through a sequence of images, bi-dimensional in the 3D model, generated using the program '3DSMAX', imported by the program 'TOMO M C' and thus, introduced as INPUT FILE of the MCNP code. (author)

Tokamak Simulation Code modeling of NSTX

International Nuclear Information System (INIS)

Jardin, S.C.; Kaye, S.; Menard, J.; Kessel, C.; Glasser, A.H.

2000-01-01

The Tokamak Simulation Code [TSC] is widely used for the design of new axisymmetric toroidal experiments. In particular, TSC was used extensively in the design of the National Spherical Torus eXperiment [NSTX]. The authors have now benchmarked TSC with initial NSTX results and find excellent agreement for plasma and vessel currents and magnetic flux loops when the experimental coil currents are used in the simulations. TSC has also been coupled with a ballooning stability code and with DCON to provide stability predictions for NSTX operation. TSC has also been used to model initial CHI experiments where a large poloidal voltage is applied to the NSTX vacuum vessel, causing a force-free current to appear in the plasma. This is a phenomenon that is similar to the plasma halo current that sometimes develops during a plasma disruption

The modelling of wall condensation with noncondensable gases for the containment codes

Energy Technology Data Exchange (ETDEWEB)

Leduc, C.; Coste, P.; Barthel, V.; Deslandes, H. [Commissariat a l`Energi Atomique, Grenoble (France)

1995-09-01

This paper presents several approaches in the modelling of wall condensation in the presence of noncondensable gases for containment codes. The lumped-parameter modelling and the local modelling by 3-D codes are discussed. Containment analysis codes should be able to predict the spatial distributions of steam, air, and hydrogen as well as the efficiency of cooling by wall condensation in both natural convection and forced convection situations. 3-D calculations with a turbulent diffusion modelling are necessary since the diffusion controls the local condensation whereas the wall condensation may redistribute the air and hydrogen mass in the containment. A fine mesh modelling of film condensation in forced convection has been in the developed taking into account the influence of the suction velocity at the liquid-gas interface. It is associated with the 3-D model of the TRIO code for the gas mixture where a k-{xi} turbulence model is used. The predictions are compared to the Huhtiniemi`s experimental data. The modelling of condensation in natural convection or mixed convection is more complex. As no universal velocity and temperature profile exist for such boundary layers, a very fine nodalization is necessary. More simple models integrate equations over the boundary layer thickness, using the heat and mass transfer analogy. The model predictions are compared with a MIT experiment. For the containment compartments a two node model is proposed using the lumped parameter approach. Heat and mass transfer coefficients are tested on separate effect tests and containment experiments. The CATHARE code has been adapted to perform such calculations and shows a reasonable agreement with data.

Sodium pool fire model for CONACS code

International Nuclear Information System (INIS)

Yung, S.C.

1982-01-01

The modeling of sodium pool fires constitutes an important ingredient in conducting LMFBR accident analysis. Such modeling capability has recently come under scrutiny at Westinghouse Hanford Company (WHC) within the context of developing CONACS, the Containment Analysis Code System. One of the efforts in the CONACS program is to model various combustion processes anticipated to occur during postulated accident paths. This effort includes the selection or modification of an existing model and development of a new model if it clearly contributes to the program purpose. As part of this effort, a new sodium pool fire model has been developed that is directed at removing some of the deficiencies in the existing models, such as SOFIRE-II and FEUNA

A predictive coding account of bistable perception - a model-based fMRI study.

Science.gov (United States)

Weilnhammer, Veith; Stuke, Heiner; Hesselmann, Guido; Sterzer, Philipp; Schmack, Katharina

2017-05-01

In bistable vision, subjective perception wavers between two interpretations of a constant ambiguous stimulus. This dissociation between conscious perception and sensory stimulation has motivated various empirical studies on the neural correlates of bistable perception, but the neurocomputational mechanism behind endogenous perceptual transitions has remained elusive. Here, we recurred to a generic Bayesian framework of predictive coding and devised a model that casts endogenous perceptual transitions as a consequence of prediction errors emerging from residual evidence for the suppressed percept. Data simulations revealed close similarities between the model's predictions and key temporal characteristics of perceptual bistability, indicating that the model was able to reproduce bistable perception. Fitting the predictive coding model to behavioural data from an fMRI-experiment on bistable perception, we found a correlation across participants between the model parameter encoding perceptual stabilization and the behaviourally measured frequency of perceptual transitions, corroborating that the model successfully accounted for participants' perception. Formal model comparison with established models of bistable perception based on mutual inhibition and adaptation, noise or a combination of adaptation and noise was used for the validation of the predictive coding model against the established models. Most importantly, model-based analyses of the fMRI data revealed that prediction error time-courses derived from the predictive coding model correlated with neural signal time-courses in bilateral inferior frontal gyri and anterior insulae. Voxel-wise model selection indicated a superiority of the predictive coding model over conventional analysis approaches in explaining neural activity in these frontal areas, suggesting that frontal cortex encodes prediction errors that mediate endogenous perceptual transitions in bistable perception. Taken together, our current work

A predictive coding account of bistable perception - a model-based fMRI study.

Directory of Open Access Journals (Sweden)

Veith Weilnhammer

2017-05-01

Full Text Available In bistable vision, subjective perception wavers between two interpretations of a constant ambiguous stimulus. This dissociation between conscious perception and sensory stimulation has motivated various empirical studies on the neural correlates of bistable perception, but the neurocomputational mechanism behind endogenous perceptual transitions has remained elusive. Here, we recurred to a generic Bayesian framework of predictive coding and devised a model that casts endogenous perceptual transitions as a consequence of prediction errors emerging from residual evidence for the suppressed percept. Data simulations revealed close similarities between the model's predictions and key temporal characteristics of perceptual bistability, indicating that the model was able to reproduce bistable perception. Fitting the predictive coding model to behavioural data from an fMRI-experiment on bistable perception, we found a correlation across participants between the model parameter encoding perceptual stabilization and the behaviourally measured frequency of perceptual transitions, corroborating that the model successfully accounted for participants' perception. Formal model comparison with established models of bistable perception based on mutual inhibition and adaptation, noise or a combination of adaptation and noise was used for the validation of the predictive coding model against the established models. Most importantly, model-based analyses of the fMRI data revealed that prediction error time-courses derived from the predictive coding model correlated with neural signal time-courses in bilateral inferior frontal gyri and anterior insulae. Voxel-wise model selection indicated a superiority of the predictive coding model over conventional analysis approaches in explaining neural activity in these frontal areas, suggesting that frontal cortex encodes prediction errors that mediate endogenous perceptual transitions in bistable perception. Taken together

Effects of Secondary Circuit Modeling on Results of Pressurized Water Reactor Main Steam Line Break Benchmark Calculations with New Coupled Code TRAB-3D/SMABRE

International Nuclear Information System (INIS)

Daavittila, Antti; Haemaelaeinen, Anitta; Kyrki-Rajamaeki, Riitta

2003-01-01

All of the three exercises of the Organization for Economic Cooperation and Development/Nuclear Regulatory Commission pressurized water reactor main steam line break (PWR MSLB) benchmark were calculated at VTT, the Technical Research Centre of Finland. For the first exercise, the plant simulation with point-kinetic neutronics, the thermal-hydraulics code SMABRE was used. The second exercise was calculated with the three-dimensional reactor dynamics code TRAB-3D, and the third exercise with the combination TRAB-3D/SMABRE. VTT has over ten years' experience of coupling neutronic and thermal-hydraulic codes, but this benchmark was the first time these two codes, both developed at VTT, were coupled together. The coupled code system is fast and efficient; the total computation time of the 100-s transient in the third exercise was 16 min on a modern UNIX workstation. The results of all the exercises are similar to those of the other participants. In order to demonstrate the effect of secondary circuit modeling on the results, three different cases were calculated. In case 1 there is no phase separation in the steam lines and no flow reversal in the aspirator. In case 2 the flow reversal in the aspirator is allowed, but there is no phase separation in the steam lines. Finally, in case 3 the drift-flux model is used for the phase separation in the steam lines, but the aspirator flow reversal is not allowed. With these two modeling variations, it is possible to cover a remarkably broad range of results. The maximum power level reached after the reactor trip varies from 534 to 904 MW, the range of the time of the power maximum being close to 30 s. Compared to the total calculated transient time of 100 s, the effect of the secondary side modeling is extremely important

Modeling of the YALINA booster facility by the Monte Carlo code MONK

International Nuclear Information System (INIS)

Talamo, A.; Gohar, Y.; Kondev, F.; Kiyavitskaya, H.; Serafimovich, I.; Bournos, V.; Fokov, Y.; Routkovskaya, C.

2007-01-01

The YALINA-Booster facility has been modeled according to the benchmark specifications defined for the IAEA activity without any geometrical homogenization using the Monte Carlo codes MONK and MCNP/MCNPX/MCB. The MONK model perfectly matches the MCNP one. The computational analyses have been extended through the MCB code, which is an extension of the MCNP code with burnup capability because of its additional feature for analyzing source driven multiplying assemblies. The main neutronics arameters of the YALINA-Booster facility were calculated using these computer codes with different nuclear data libraries based on ENDF/B-VI-0, -6, JEF-2.2, and JEF-3.1.

Code on the safety of nuclear power plants: Governmental organization

International Nuclear Information System (INIS)

1988-01-01

This Code recommends requirements for a regulatory body responsible for regulating the siting, design, construction, commissioning, operation and decommissioning of nuclear power plants for safety. It forms part of the Agency's programme for establishing Codes and Safety Guides relating to land based stationary thermal neutron power plants

Installation of aerosol behavior model into multi-dimensional thermal hydraulic analysis code AQUA

International Nuclear Information System (INIS)

Kisohara, Naoyuki; Yamaguchi, Akira

1997-12-01

The safety analysis of FBR plant system for sodium leak phenomena needs to evaluate the deposition of the aerosol particle to the components in the plant, the chemical reaction of aerosol to humidity in the air and the effect of the combustion heat through aerosol to the structural component. For this purpose, ABC-INTG (Aerosol Behavior in Containment-INTeGrated Version) code has been developed and used until now. This code calculates aerosol behavior in the gas area of uniform temperature and pressure by 1 cell-model. Later, however, more detailed calculation of aerosol behavior requires the installation of aerosol model into multi-cell thermal hydraulic analysis code AQUA. AQUA can calculate the carrier gas flow, temperature and the distribution of the aerosol spatial concentration. On the other hand, ABC-INTG can calculate the generation, deposition to the wall and flower, agglomeration of aerosol particle and figure out the distribution of the aerosol particle size. Thus, the combination of these two codes enables to deal with aerosol model coupling the distribution of the aerosol spatial concentration and that of the aerosol particle size. This report describes aerosol behavior model, how to install the aerosol model to AQUA and new subroutine equipped to the code. Furthermore, the test calculations of the simple structural model were executed by this code, appropriate results were obtained. Thus, this code has prospect to predict aerosol behavior by the introduction of coupling analysis with multi-dimensional gas thermo-dynamics for sodium combustion evaluation. (J.P.N.)

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.

The ELOCA fuel modelling code: past, present and future

International Nuclear Information System (INIS)

Williams, A.F.

2005-01-01

ELOCA is the Industry Standard Toolset (IST) computer code for modelling CANDU fuel under the transient coolant conditions typical of an accident scenario. Since its original inception in the early 1970's, the code has undergone continual development and improvement. The code now embodies much of the knowledge and experience of fuel behaviour gained by the Canadian nuclear industry over this period. ELOCA has proven to be a valuable tool for the safety analyst, and continues to be used extensively to support the licensing cases of CANDU reactors. This paper provides a brief and much simplified view of this development history, its current status, and plans for future development. (author)

An improved thermal model for the computer code NAIAD

International Nuclear Information System (INIS)

Rainbow, M.T.

1982-12-01

An improved thermal model, based on the concept of heat slabs, has been incorporated as an option into the thermal hydraulic computer code NAIAD. The heat slabs are one-dimensional thermal conduction models with temperature independent thermal properties which may be internal and/or external to the fluid. Thermal energy may be added to or removed from the fluid via heat slabs and passed across the external boundary of external heat slabs at a rate which is a linear function of the external surface temperatures. The code input for the new option has been restructured to simplify data preparation. A full description of current input requirements is presented

Advanced codes and methods supporting improved fuel cycle economics - 5493

International Nuclear Information System (INIS)

Curca-Tivig, F.; Maupin, K.; Thareau, S.

2015-01-01

AREVA's code development program was practically completed in 2014. The basic codes supporting a new generation of advanced methods are the followings. GALILEO is a state-of-the-art fuel rod performance code for PWR and BWR applications. Development is completed, implementation started in France and the U.S.A. ARCADIA-1 is a state-of-the-art neutronics/ thermal-hydraulics/ thermal-mechanics code system for PWR applications. Development is completed, implementation started in Europe and in the U.S.A. The system thermal-hydraulic codes S-RELAP5 and CATHARE-2 are not really new but still state-of-the-art in the domain. S-RELAP5 was completely restructured and re-coded such that its life cycle increases by further decades. CATHARE-2 will be replaced in the future by the new CATHARE-3. The new AREVA codes and methods are largely based on first principles modeling with an extremely broad international verification and validation data base. This enables AREVA and its customers to access more predictable licensing processes in a fast evolving regulatory environment (new safety criteria, requests for enlarged qualification databases, statistical applications, uncertainty propagation...). In this context, the advanced codes and methods and the associated verification and validation represent the key to avoiding penalties on products, on operational limits, or on methodologies themselves

FAVOR: A new fracture mechanics code for reactor pressure vessels subjected to pressurized thermal shock

International Nuclear Information System (INIS)

Dickson, T.L.

1993-01-01

Probabilistic fracture mechanics (PFM) analysis is a major element of the comprehensive probabilistic methodology endorsed by the Nuclear Regulatory Commission (NRC) for evaluation of the integrity of pressurized water reactor pressure vessels subjected to pressurized-thermal-shock (PTS) transients. OCA-P and VISA-II are PTS PFM computer codes that are currently referenced in Regulatory Guide 1.154 as acceptable codes for performing plant-specific analyses. These codes perform PFM analyses to estimate the increase in vessel failure probability as the vessel accumulates radiation damage over the operating life of the vessel. Experience with the application of these codes in the last few years has provided insights into areas where they could be improved. As more plants approach the PTS screening criteria and are required to perform plant-specific analyses, there will be an increasing need for an improved and validated PTS PFM code that is accepted by the NRC and utilities. The NRC funded Heavy Section Steel Technology Program (HSST) at the Oak Ridge National Laboratory is currently developing the FAVOR (Fracture Analysis of Vessels: Oak Ridge) code, which is expected to meet this need. The FAVOR code incorporates the most important features of both OCA-P and VISA-II and contains some new capabilities such as (1) a PFM global modeling methodology; (2) the calculation of the axial stress component associated with coolant streaming beneath an inlet nozzle; (3) a library of stress intensity factor influence coefficients, generated by the NQA-1 certified ABAQUS computer code, for an appropriate range of two and three dimensional inner-surface flaws; (4) the flexibility to generate a variety of output reports; and (5) enhanced user friendliness

A statistical methodology for quantification of uncertainty in best estimate code physical models

International Nuclear Information System (INIS)

Vinai, Paolo; Macian-Juan, Rafael; Chawla, Rakesh

2007-01-01

A novel uncertainty assessment methodology, based on a statistical non-parametric approach, is presented in this paper. It achieves quantification of code physical model uncertainty by making use of model performance information obtained from studies of appropriate separate-effect tests. Uncertainties are quantified in the form of estimated probability density functions (pdf's), calculated with a newly developed non-parametric estimator. The new estimator objectively predicts the probability distribution of the model's 'error' (its uncertainty) from databases reflecting the model's accuracy on the basis of available experiments. The methodology is completed by applying a novel multi-dimensional clustering technique based on the comparison of model error samples with the Kruskall-Wallis test. This takes into account the fact that a model's uncertainty depends on system conditions, since a best estimate code can give predictions for which the accuracy is affected by the regions of the physical space in which the experiments occur. The final result is an objective, rigorous and accurate manner of assigning uncertainty to coded models, i.e. the input information needed by code uncertainty propagation methodologies used for assessing the accuracy of best estimate codes in nuclear systems analysis. The new methodology has been applied to the quantification of the uncertainty in the RETRAN-3D void model and then used in the analysis of an independent separate-effect experiment. This has clearly demonstrated the basic feasibility of the approach, as well as its advantages in yielding narrower uncertainty bands in quantifying the code's accuracy for void fraction predictions

MINI-TRAC code: a driver program for assessment of constitutive equations of two-fluid model

International Nuclear Information System (INIS)

Akimoto, Hajime; Abe, Yutaka; Ohnuki, Akira; Murao, Yoshio

1991-05-01

MINI-TRAC code, a driver program for assessment of constitutive equations of two-fluid model, has been developed to perform assessment and improvement of constitutive equations of two-fluid model widely and efficiently. The MINI-TRAC code uses one-dimensional conservation equations for mass, momentum and energy based on the two-fluid model. The code can work on a personal computer because it can be operated with a core memory size less than 640 KB. The MINI-TRAC code includes constitutive equations of TRAC-PF1/MOD1 code, TRAC-BF1 code and RELAP5/MOD2 code. The code is modulated so that one can easily change constitutive equations to perform a test calculation. This report is a manual of the MINI-TRAC code. The basic equations, numerics, constitutive, equations included in the MINI-TRAC code will be described. The user's manual such as input description will be presented. The program structure and contents of main variables will also be mentioned in this report. (author)

The implementation of a toroidal limiter model into the gyrokinetic code ELMFIRE

Energy Technology Data Exchange (ETDEWEB)

Leerink, S.; Janhunen, S.J.; Kiviniemi, T.P.; Nora, M. [Euratom-Tekes Association, Helsinki University of Technology (Finland); Heikkinen, J.A. [Euratom-Tekes Association, VTT, P.O. Box 1000, FI-02044 VTT (Finland); Ogando, F. [Universidad Nacional de Educacion a Distancia, Madrid (Spain)

2008-03-15

The ELMFIRE full nonlinear gyrokinetic simulation code has been developed for calculations of plasma evolution and dynamics of turbulence in tokamak geometry. The code is applicable for calculations of strong perturbations in particle distribution function, rapid transients and steep gradients in plasma. Benchmarking against experimental reflectometry data from the FT2 tokamak is being discussed and in this paper a model for comparison and studying poloidal velocity is presented. To make the ELMFIRE code suitable for scrape-off layer simulations a simplified toroidal limiter model has been implemented. The model is be discussed and first results are presented. (copyright 2008 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

MIGFRAC - a code for modelling of radionuclide transport in fracture media

International Nuclear Information System (INIS)

Satyanarayana, S.V.M.; Mohankumar, N.; Sasidhar, P.

2002-05-01

Radionuclides migrate through diffusion process from radioactive waste disposal facilities into fractures present in the host rock. The transport phenomenon is aided by the circulating ground waters. To model the transport of radionuclides in the charnockite rock formations present at Kalpakkam, a numerical code - MIGFRAC has been developed at SHINE Group, IGCAR. The code has been subjected to rigorous tests and the results of the build up of radionuclide concentrations are validated with a test case up to a distance of 100 meter along the fracture. The report discusses the model, code features and the results obtained up to a distance of 400 meter are presented. (author)

MEMOPS: data modelling and automatic code generation.

Science.gov (United States)

Fogh, Rasmus H; Boucher, Wayne; Ionides, John M C; Vranken, Wim F; Stevens, Tim J; Laue, Ernest D

2010-03-25

In recent years the amount of biological data has exploded to the point where much useful information can only be extracted by complex computational analyses. Such analyses are greatly facilitated by metadata standards, both in terms of the ability to compare data originating from different sources, and in terms of exchanging data in standard forms, e.g. when running processes on a distributed computing infrastructure. However, standards thrive on stability whereas science tends to constantly move, with new methods being developed and old ones modified. Therefore maintaining both metadata standards, and all the code that is required to make them useful, is a non-trivial problem. Memops is a framework that uses an abstract definition of the metadata (described in UML) to generate internal data structures and subroutine libraries for data access (application programming interfaces--APIs--currently in Python, C and Java) and data storage (in XML files or databases). For the individual project these libraries obviate the need for writing code for input parsing, validity checking or output. Memops also ensures that the code is always internally consistent, massively reducing the need for code reorganisation. Across a scientific domain a Memops-supported data model makes it easier to support complex standards that can capture all the data produced in a scientific area, share them among all programs in a complex software pipeline, and carry them forward to deposition in an archive. The principles behind the Memops generation code will be presented, along with example applications in Nuclear Magnetic Resonance (NMR) spectroscopy and structural biology.

Establishing a regulatory value chain model: An innovative approach to strengthening medicines regulatory systems in resource-constrained settings.

Science.gov (United States)

Chahal, Harinder Singh; Kashfipour, Farrah; Susko, Matt; Feachem, Neelam Sekhri; Boyle, Colin

2016-05-01

Medicines Regulatory Authorities (MRAs) are an essential part of national health systems and are charged with protecting and promoting public health through regulation of medicines. However, MRAs in resource-constrained settings often struggle to provide effective oversight of market entry and use of health commodities. This paper proposes a regulatory value chain model (RVCM) that policymakers and regulators can use as a conceptual framework to guide investments aimed at strengthening regulatory systems. The RVCM incorporates nine core functions of MRAs into five modules: (i) clear guidelines and requirements; (ii) control of clinical trials; (iii) market authorization of medical products; (iv) pre-market quality control; and (v) post-market activities. Application of the RVCM allows national stakeholders to identify and prioritize investments according to where they can add the most value to the regulatory process. Depending on the economy, capacity, and needs of a country, some functions can be elevated to a regional or supranational level, while others can be maintained at the national level. In contrast to a "one size fits all" approach to regulation in which each country manages the full regulatory process at the national level, the RVCM encourages leveraging the expertise and capabilities of other MRAs where shared processes strengthen regulation. This value chain approach provides a framework for policymakers to maximize investment impact while striving to reach the goal of safe, affordable, and rapidly accessible medicines for all.

Three-dimensional all-speed CFD code for safety analysis of nuclear reactor containment: Status of GASFLOW parallelization, model development, validation and application

Energy Technology Data Exchange (ETDEWEB)

Xiao, Jianjun, E-mail: jianjun.xiao@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Travis, John R., E-mail: jack_travis@comcast.com [Engineering and Scientific Software Inc., 3010 Old Pecos Trail, Santa Fe, NM 87505 (United States); Royl, Peter, E-mail: peter.royl@partner.kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Necker, Gottfried, E-mail: gottfried.necker@partner.kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Svishchev, Anatoly, E-mail: anatoly.svishchev@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Jordan, Thomas, E-mail: thomas.jordan@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany)

2016-05-15

Highlights: • 3-D scalable semi-implicit pressure-based CFD code for containment safety analysis. • Robust solution algorithm valid for all-speed flows. • Well validated and widely used CFD code for hydrogen safety analysis. • Code applied in various types of nuclear reactor containments. • Parallelization enables high-fidelity models in large scale containment simulations. - Abstract: GASFLOW is a three dimensional semi-implicit all-speed CFD code which can be used to predict fluid dynamics, chemical kinetics, heat and mass transfer, aerosol transportation and other related phenomena involved in postulated accidents in nuclear reactor containments. The main purpose of the paper is to give a brief review on recent GASFLOW code development, validations and applications in the field of nuclear safety. GASFLOW code has been well validated by international experimental benchmarks, and has been widely applied to hydrogen safety analysis in various types of nuclear power plants in European and Asian countries, which have been summarized in this paper. Furthermore, four benchmark tests of a lid-driven cavity flow, low Mach number jet flow, 1-D shock tube and supersonic flow over a forward-facing step are presented in order to demonstrate the accuracy and wide-ranging capability of ICE’d ALE solution algorithm for all-speed flows. GASFLOW has been successfully parallelized using the paradigms of Message Passing Interface (MPI) and domain decomposition. The parallel version, GASFLOW-MPI, adds great value to large scale containment simulations by enabling high-fidelity models, including more geometric details and more complex physics. It will be helpful for the nuclear safety engineers to better understand the hydrogen safety related physical phenomena during the severe accident, to optimize the design of the hydrogen risk mitigation systems and to fulfill the licensing requirements by the nuclear regulatory authorities. GASFLOW-MPI is targeting a high

Realistic edge field model code REFC for designing and study of isochronous cyclotron

International Nuclear Information System (INIS)

Ismail, M.

1989-01-01

The focussing properties and the requirements for isochronism in cyclotron magnet configuration are well-known in hard edge field model. The fact that they quite often change considerably in realistic field can be attributed mainly to the influence of the edge field. A solution to this problem requires a field model which allows a simple construction of equilibrium orbit and yield simple formulae. This can be achieved by using a fitted realistic edge field (Hudson et al 1975) in the region of the pole edge and such a field model is therefore called a realistic edge field model. A code REFC based on realistic edge field model has been developed to design the cyclotron sectors and the code FIELDER has been used to study the beam properties. In this report REFC code has been described along with some relevant explaination of the FIELDER code. (author). 11 refs., 6 figs

The WARP Code: Modeling High Intensity Ion Beams

International Nuclear Information System (INIS)

Grote, D P; Friedman, A; Vay, J L; Haber, I

2004-01-01

The Warp code, developed for heavy-ion driven inertial fusion energy studies, is used to model high intensity ion (and electron) beams. Significant capability has been incorporated in Warp, allowing nearly all sections of an accelerator to be modeled, beginning with the source. Warp has as its core an explicit, three-dimensional, particle-in-cell model. Alongside this is a rich set of tools for describing the applied fields of the accelerator lattice, and embedded conducting surfaces (which are captured at sub-grid resolution). Also incorporated are models with reduced dimensionality: an axisymmetric model and a transverse ''slice'' model. The code takes advantage of modern programming techniques, including object orientation, parallelism, and scripting (via Python). It is at the forefront in the use of the computational technique of adaptive mesh refinement, which has been particularly successful in the area of diode and injector modeling, both steady-state and time-dependent. In the presentation, some of the major aspects of Warp will be overviewed, especially those that could be useful in modeling ECR sources. Warp has been benchmarked against both theory and experiment. Recent results will be presented showing good agreement of Warp with experimental results from the STS500 injector test stand. Additional information can be found on the web page http://hif.lbl.gov/theory/WARP( ) summary.html

Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC): gap analysis for high fidelity and performance assessment code development

International Nuclear Information System (INIS)

Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe Jr.; Webb, Stephen Walter; Dewers, Thomas A.; Mariner, Paul E.; Edwards, Harold Carter; Fuller, Timothy J.; Freeze, Geoffrey A.; Jove-Colon, Carlos F.; Wang, Yifeng

2011-01-01

This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are

Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC) : gap analysis for high fidelity and performance assessment code development.

Energy Technology Data Exchange (ETDEWEB)

Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe, Jr.; Webb, Stephen Walter; Dewers, Thomas A.; Mariner, Paul E.; Edwards, Harold Carter; Fuller, Timothy J.; Freeze, Geoffrey A.; Jove-Colon, Carlos F.; Wang, Yifeng

2011-03-01

This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are

Groundwater flow code verification ''benchmarking'' activity (COVE-2A): Analysis of participants' work

International Nuclear Information System (INIS)

Dykhuizen, R.C.; Barnard, R.W.

1992-02-01

The Nuclear Waste Repository Technology Department at Sandia National Laboratories (SNL) is investigating the suitability of Yucca Mountain as a potential site for underground burial of nuclear wastes. One element of the investigations is to assess the potential long-term effects of groundwater flow on the integrity of a potential repository. A number of computer codes are being used to model groundwater flow through geologic media in which the potential repository would be located. These codes compute numerical solutions for problems that are usually analytically intractable. Consequently, independent confirmation of the correctness of the solution is often not possible. Code verification is a process that permits the determination of the numerical accuracy of codes by comparing the results of several numerical solutions for the same problem. The international nuclear waste research community uses benchmarking for intercomparisons that partially satisfy the Nuclear Regulatory Commission (NRC) definition of code verification. This report presents the results from the COVE-2A (Code Verification) project, which is a subset of the COVE project

Computational modeling identifies key gene regulatory interactions underlying phenobarbital-mediated tumor promotion

Science.gov (United States)

Luisier, Raphaëlle; Unterberger, Elif B.; Goodman, Jay I.; Schwarz, Michael; Moggs, Jonathan; Terranova, Rémi; van Nimwegen, Erik

2014-01-01

Gene regulatory interactions underlying the early stages of non-genotoxic carcinogenesis are poorly understood. Here, we have identified key candidate regulators of phenobarbital (PB)-mediated mouse liver tumorigenesis, a well-characterized model of non-genotoxic carcinogenesis, by applying a new computational modeling approach to a comprehensive collection of in vivo gene expression studies. We have combined our previously developed motif activity response analysis (MARA), which models gene expression patterns in terms of computationally predicted transcription factor binding sites with singular value decomposition (SVD) of the inferred motif activities, to disentangle the roles that different transcriptional regulators play in specific biological pathways of tumor promotion. Furthermore, transgenic mouse models enabled us to identify which of these regulatory activities was downstream of constitutive androstane receptor and β-catenin signaling, both crucial components of PB-mediated liver tumorigenesis. We propose novel roles for E2F and ZFP161 in PB-mediated hepatocyte proliferation and suggest that PB-mediated suppression of ESR1 activity contributes to the development of a tumor-prone environment. Our study shows that combining MARA with SVD allows for automated identification of independent transcription regulatory programs within a complex in vivo tissue environment and provides novel mechanistic insights into PB-mediated hepatocarcinogenesis. PMID:24464994

Current and anticipated uses of the thermal hydraulics codes at the NRC

Energy Technology Data Exchange (ETDEWEB)

Caruso, R.

1997-07-01

The focus of Thermal-Hydraulic computer code usage in nuclear regulatory organizations has undergone a considerable shift since the codes were originally conceived. Less work is being done in the area of {open_quotes}Design Basis Accidents,{close_quotes}, and much more emphasis is being placed on analysis of operational events, probabalistic risk/safety assessment, and maintenance practices. All of these areas need support from Thermal-Hydraulic computer codes to model the behavior of plant fluid systems, and they all need the ability to perform large numbers of analyses quickly. It is therefore important for the T/H codes of the future to be able to support these needs, by providing robust, easy-to-use, tools that produce easy-to understand results for a wider community of nuclear professionals. These tools need to take advantage of the great advances that have occurred recently in computer software, by providing users with graphical user interfaces for both input and output. In addition, reduced costs of computer memory and other hardware have removed the need for excessively complex data structures and numerical schemes, which make the codes more difficult and expensive to modify, maintain, and debug, and which increase problem run-times. Future versions of the T/H codes should also be structured in a modular fashion, to allow for the easy incorporation of new correlations, models, or features, and to simplify maintenance and testing. Finally, it is important that future T/H code developers work closely with the code user community, to ensure that the code meet the needs of those users.

Current and anticipated uses of the thermal hydraulics codes at the NRC

International Nuclear Information System (INIS)

Caruso, R.

1997-01-01

The focus of Thermal-Hydraulic computer code usage in nuclear regulatory organizations has undergone a considerable shift since the codes were originally conceived. Less work is being done in the area of Design Basis Accidents, , and much more emphasis is being placed on analysis of operational events, probabalistic risk/safety assessment, and maintenance practices. All of these areas need support from Thermal-Hydraulic computer codes to model the behavior of plant fluid systems, and they all need the ability to perform large numbers of analyses quickly. It is therefore important for the T/H codes of the future to be able to support these needs, by providing robust, easy-to-use, tools that produce easy-to understand results for a wider community of nuclear professionals. These tools need to take advantage of the great advances that have occurred recently in computer software, by providing users with graphical user interfaces for both input and output. In addition, reduced costs of computer memory and other hardware have removed the need for excessively complex data structures and numerical schemes, which make the codes more difficult and expensive to modify, maintain, and debug, and which increase problem run-times. Future versions of the T/H codes should also be structured in a modular fashion, to allow for the easy incorporation of new correlations, models, or features, and to simplify maintenance and testing. Finally, it is important that future T/H code developers work closely with the code user community, to ensure that the code meet the needs of those users

Long non-coding RNAs: Mechanism of action and functional utility

OpenAIRE

Bhat, Shakil Ahmad; Ahmad, Syed Mudasir; Mumtaz, Peerzada Tajamul; Malik, Abrar Ahad; Dar, Mashooq Ahmad; Urwat, Uneeb; Shah, Riaz Ahmad; Ganai, Nazir Ahmad

2016-01-01

Recent RNA sequencing studies have revealed that most of the human genome is transcribed, but very little of the total transcriptomes has the ability to encode proteins. Long non-coding RNAs (lncRNAs) are non-coding transcripts longer than 200 nucleotides. Members of the non-coding genome include microRNA (miRNA), small regulatory RNAs and other short RNAs. Most of long non-coding RNA (lncRNAs) are poorly annotated. Recent recognition about lncRNAs highlights their effects in many biological ...

An evaluation of the effectiveness of the EPA comply code to demonstrate compliance with radionuclide emission standards at three manufacturing facilities