Kaizen implementation: A best case analysis

NARCIS (Netherlands)

Yokozawa, Kodo; Steenhuis, H.J.

2011-01-01

This study add insights to the concept on kaizen and its implementation process based upon an in-depth case study at OMN, a Japanese manufacturer in the Netherlands which has successfully adopted kaizen concepts. It was found that employee discipline and personal-initiatives are the two critical

Fast neutron analysis code SAD1

International Nuclear Information System (INIS)

Jung, M.; Ott, C.

1985-01-01

A listing and an example of outputs of the M.C. code SAD1 are given here. This code has been used many times to predict responses of fast neutrons in hydrogenic materials (in our case emulsions or plastics) towards the elastic n, p scattering. It can be easily extended to other kinds of such materials and to any kind of incident fast neutron spectrum

Generalized rank weights of reducible codes, optimal cases and related properties

DEFF Research Database (Denmark)

Martinez Peñas, Umberto

2018-01-01

in network coding. In this paper, we study their security behavior against information leakage on networks when applied as coset coding schemes, giving the following main results: 1) we give lower and upper bounds on their generalized rank weights (GRWs), which measure worst case information leakage...... to the wire tapper; 2) we find new parameters for which these codes are MRD (meaning that their first GRW is optimal) and use the previous bounds to estimate their higher GRWs; 3) we show that all linear (over the extension field) codes, whose GRWs are all optimal for fixed packet and code sizes but varying...... length are reducible codes up to rank equivalence; and 4) we show that the information leaked to a wire tapper when using reducible codes is often much less than the worst case given by their (optimal in some cases) GRWs. We conclude with some secondary related properties: conditions to be rank...

HYDROCOIN [HYDROlogic COde INtercomparison] Level 1: Benchmarking and verification test results with CFEST [Coupled Fluid, Energy, and Solute Transport] code: Draft report

International Nuclear Information System (INIS)

Yabusaki, S.; Cole, C.; Monti, A.M.; Gupta, S.K.

1987-04-01

Part of the safety analysis is evaluating groundwater flow through the repository and the host rock to the accessible environment by developing mathematical or analytical models and numerical computer codes describing the flow mechanisms. This need led to the establishment of an international project called HYDROCOIN (HYDROlogic COde INtercomparison) organized by the Swedish Nuclear Power Inspectorate, a forum for discussing techniques and strategies in subsurface hydrologic modeling. The major objective of the present effort, HYDROCOIN Level 1, is determining the numerical accuracy of the computer codes. The definition of each case includes the input parameters, the governing equations, the output specifications, and the format. The Coupled Fluid, Energy, and Solute Transport (CFEST) code was applied to solve cases 1, 2, 4, 5, and 7; the Finite Element Three-Dimensional Groundwater (FE3DGW) Flow Model was used to solve case 6. Case 3 has been ignored because unsaturated flow is not pertinent to SRP. This report presents the Level 1 results furnished by the project teams. The numerical accuracy of the codes is determined by (1) comparing the computational results with analytical solutions for cases that have analytical solutions (namely cases 1 and 4), and (2) intercomparing results from codes for cases which do not have analytical solutions (cases 2, 5, 6, and 7). Cases 1, 2, 6, and 7 relate to flow analyses, whereas cases 4 and 5 require nonlinear solutions. 7 refs., 71 figs., 9 tabs

Survey of 1 1/2D transport codes

International Nuclear Information System (INIS)

Grad, H.

1978-10-01

A survey is given of a family of classical transport codes, recently termed ''1 1/2D'', which efficiently and accurately follow the evolution of plasma configurations on a long time scale, following coupled changes in plasma shape and topology with transport (but not wave motion). Codes have been constructed and operated (since 1974) which include various combinations of finite beta, general plasma cross-section and aspect, various topologies (Doublet, tearing, reversed-field mirror) including time dependent transitions in topology resulting from external coil variation and plasma transport, with models including (classical) tensor resistivity and heat flow as well as the adiabatic limiting case

Benchmarking NNWSI flow and transport codes: COVE 1 results

International Nuclear Information System (INIS)

Hayden, N.K.

1985-06-01

The code verification (COVE) activity of the Nevada Nuclear Waste Storage Investigations (NNWSI) Project is the first step in certification of flow and transport codes used for NNWSI performance assessments of a geologic repository for disposing of high-level radioactive wastes. The goals of the COVE activity are (1) to demonstrate and compare the numerical accuracy and sensitivity of certain codes, (2) to identify and resolve problems in running typical NNWSI performance assessment calculations, and (3) to evaluate computer requirements for running the codes. This report describes the work done for COVE 1, the first step in benchmarking some of the codes. Isothermal calculations for the COVE 1 benchmarking have been completed using the hydrologic flow codes SAGUARO, TRUST, and GWVIP; the radionuclide transport codes FEMTRAN and TRUMP; and the coupled flow and transport code TRACR3D. This report presents the results of three cases of the benchmarking problem solved for COVE 1, a comparison of the results, questions raised regarding sensitivities to modeling techniques, and conclusions drawn regarding the status and numerical sensitivities of the codes. 30 refs

DISP1 code

International Nuclear Information System (INIS)

Vokac, P.

1999-12-01

DISP1 code is a simple tool for assessment of the dispersion of the fission product cloud escaping from a nuclear power plant after an accident. The code makes it possible to tentatively check the feasibility of calculations by more complex PSA3 codes and/or codes for real-time dispersion calculations. The number of input parameters is reasonably low and the user interface is simple enough to allow a rapid processing of sensitivity analyses. All input data entered through the user interface are stored in the text format. Implementation of dispersion model corrections taken from the ARCON96 code enables the DISP1 code to be employed for assessment of the radiation hazard within the NPP area, in the control room for instance. (P.A.)

[Seasonal distribution of clinical case codes (DOC study)].

Science.gov (United States)

von Dercks, N; Melz, R; Hepp, P; Theopold, J; Marquass, B; Josten, C

2017-02-01

The German diagnosis-related groups remuneration system (G-DRG) was implemented in 2004 and patient-related diagnoses and procedures lead to allocation to specific DRGs. This system includes several codes, such as case mix (CM), case mix index (CMI) and number of cases. Seasonal distribution of these codes as well as distribution of diagnoses and DRGs may lead to logistical consequences for clinical management. From 2004 to 2013 all the main diagnoses and DRGs for inpatients were recorded. Monthly and seasonal distributions were analyzed using ANOVA. The average monthly number of cases was 265 ± 25 cases, the average CM was 388.50 ± 51.75 and the average CMI was 1.46 ± 0.15 with no significant seasonal differences (p > 0.1). Concussion was the most frequently occurring main diagnosis (3739 cases) followed by fractures of the humeral head (699). Significant distribution differences could be shown for humeral head fractures in monthly (p = 0.018) and seasonal comparisons (p = 0.006) with a maximum in winter. Radius (p = 0.01) and ankle fractures (p ≤ 0.001) also occurred most frequently in winter. Non-bony lesions of the shoulder were significantly less in spring (p = 0.04). The DRGs showed no evidence of a monthly or seasonal clustering (p > 0.1). The significant clustering of injuries in specific months and seasons should lead to logistic consequences (e.g. operating room slots, availability of nursing and anesthesia staff). For a needs assessment the analysis of main diagnoses is more appropriate than DRGs.

Integration of QR codes into an anesthesia information management system for resident case log management.

Science.gov (United States)

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

2015-04-01

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

Evolution of ASTEC V1.2 rev.1 code for WWER-1000 reactors/SBO sequence

International Nuclear Information System (INIS)

Georgieva, J.; Stefanova, A.; Groudev, P.; Tusheva, P.; Kalchev, B.; Passalacqua, R.

2006-01-01

In this paper a comparison between calculations of severe accidents occurred from WWER-1000 with ASTEC code specified for an event of full unloading with relief valves stuck opened with no hydroaccumulators intervention is presented. The purpose of the analyses provided is to present the relationship between the improvements of the actual version (ASTEC Vl.2 rev. 1) and ASTEC V1.1 p2 like: code modifications, incoming data improvements. Such discrepancies are to be examined. Case by case suggestions for ASTEC improvements are to be provided

The SEDA computer code and its utilization for Angra 1

International Nuclear Information System (INIS)

Fernandes Filho, T.L.

1988-11-01

The implementation of SEDA 2.0 computer code, developed at Ezeiza Atomic Center, Argentine for Angra 1 reactor is described. The SEDA code gives an estimate for radiological consequences of nuclear accidents with release of radiactive materials for the environment. This code is now available for an IBM PC-XT. The computer environment, the files used, data, the programining structure and the models used are presented. The input data and results for two sample case are described. (author) [pt

Interpretation, with respect to ASME code Case N-318, of limit moment and fatigue tests of lugs welded to pipe

International Nuclear Information System (INIS)

Foster, D.C.; Van Duyne, D.A.; Budlong, L.A.; Muffett, J.W.; Wais, E.A.; Streck, G.; Rodabaugh, E.C.

1990-01-01

Two nonmandatory ASME code cases have been used often in the evaluation of lugs on nuclear-power- plant piping systems. ASME Code Case N-318 provides guidance for evaluation of the design of rectangular cross-section attachments on Class 2 or 3 piping, and ASME Code Case N-122 provides guidance for evaluation of lugs on Class 1 piping. These code cases have been reviewed and evaluated based on available test data. The results indicate that the Code cases are overly conservative. Recommendations for revisions to the cases are presented which, if adopted, will reduce the overconservatism

Case Study of the NENE Code Project

National Research Council Canada - National Science Library

Kendall, Richard; Post, Douglass; Mark, Andrew

2007-01-01

...) Program is sponsoring a series of case studies to identify the life cycles, workflows, and technical challenges of computational science and engineering code development that are representative...

Sample test cases using the environmental computer code NECTAR

International Nuclear Information System (INIS)

Ponting, A.C.

1984-06-01

This note demonstrates a few of the many different ways in which the environmental computer code NECTAR may be used. Four sample test cases are presented and described to show how NECTAR input data are structured. Edited output is also presented to illustrate the format of the results. Two test cases demonstrate how NECTAR may be used to study radio-isotopes not explicitly included in the code. (U.K.)

Comprehensive Report For Proposed Elevated Temperature Elastic Perfectly Plastic (EPP) Code Cases Representative Example Problems

Energy Technology Data Exchange (ETDEWEB)

Hollinger, Greg L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-06-01

Background: The current rules in the nuclear section of the ASME Boiler and Pressure Vessel (B&PV) Code , Section III, Subsection NH for the evaluation of strain limits and creep-fatigue damage using simplified methods based on elastic analysis have been deemed inappropriate for Alloy 617 at temperatures above 1200F (650C)1. To address this issue, proposed code rules have been developed which are based on the use of elastic-perfectly plastic (E-PP) analysis methods and which are expected to be applicable to very high temperatures. The proposed rules for strain limits and creep-fatigue evaluation were initially documented in the technical literature 2, 3, and have been recently revised to incorporate comments and simplify their application. The revised code cases have been developed. Task Objectives: The goal of the Sample Problem task is to exercise these code cases through example problems to demonstrate their feasibility and, also, to identify potential corrections and improvements should problems be encountered. This will provide input to the development of technical background documents for consideration by the applicable B&PV committees considering these code cases for approval. This task has been performed by Hollinger and Pease of Becht Engineering Co., Inc., Nuclear Services Division and a report detailing the results of the E-PP analyses conducted on example problems per the procedures of the E-PP strain limits and creep-fatigue draft code cases is enclosed as Enclosure 1. Conclusions: The feasibility of the application of the E-PP code cases has been demonstrated through example problems that consist of realistic geometry (a nozzle attached to a semi-hemispheric shell with a circumferential weld) and load (pressure; pipe reaction load applied at the end of the nozzle, including axial and shear forces, bending and torsional moments; through-wall transient temperature gradient) and design and operating conditions (Levels A, B and C).

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

The PWR spectral code GELS. Pt. 1

International Nuclear Information System (INIS)

Penndorf, K.; Schult, F.; Schulz, G.

1976-01-01

The code procedures group constant libraries for the static PWR design of whatever fuel cycle - Uranium, Thorium, or Plutonium. The whole reach of temperatures is covered and the treatment of strong lumped absorbers as control or burnable poison pins is included. The main features are: 1) Good accuracy in spite of not fitting the material data to critical experiments; 2) speed and relatively low computer equipment; 3) restriction to PWR's only. In case of demands for higher accuracy there is a further restriction concerning the library data of the epithermal resonance absorbers: They are strictly valid only for several special lattice geometrics. Three samples are given each representing a typical application of the code. Two of them likewise are demonstrations of recalculated experiments. (orig.) [de

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

Coding response to a case-mix measurement system based on multiple diagnoses.

Science.gov (United States)

Preyra, Colin

2004-08-01

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

Coding Response to a Case-Mix Measurement System Based on Multiple Diagnoses

Science.gov (United States)

Preyra, Colin

2004-01-01

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

Preliminary application of the draft code case for alloy 617 for a high temperature component

International Nuclear Information System (INIS)

Lee, Hyeong Yeon; Kim, Yong Wan; Song, Kee Nam

2008-01-01

The ASME draft Code Case for Alloy 617 was developed in the late 1980s for the design of very-high-temperature gas cooled reactors. The draft Code Case was patterned after the ASME Code Section III Subsection NH and was intended to cover Ni-Cr-Co-Mo Alloy 617 to 982 .deg. C (1800 .deg. F). But the draft Code Case is still in an incomplete status, lacking necessary material properties and design data. In this study, a preliminary evaluation on the creep-fatigue damage for a high temperature hot duct pipe structure has been carried out according to the draft Code Case. The evaluation procedures and results according to the draft Code Case for Alloy 617 material were compared with those of the ASME Subsection NH and RCC-MR for Alloy 800H material. It was shown that many data including material properties, fatigue and creep data should be supplemented for the draft Code Case. However, when the evaluation results on the creep-fatigue damage according to the draft Code Case, ASME-NH and RCC-MR were compared based on the preliminary evaluation, it was shown that the Alloy 617 results from the draft Code Case tended to be more resistant to the creep damage while less resistant to the fatigue damage than those from the ASME-NH and RCC-MR

Evaluating Coding Accuracy in General Surgery Residents' Accreditation Council for Graduate Medical Education Procedural Case Logs.

Science.gov (United States)

Balla, Fadi; Garwe, Tabitha; Motghare, Prasenjeet; Stamile, Tessa; Kim, Jennifer; Mahnken, Heidi; Lees, Jason

The Accreditation Council for Graduate Medical Education (ACGME) case log captures resident operative experience based on Current Procedural Terminology (CPT) codes and is used to track operative experience during residency. With increasing emphasis on resident operative experiences, coding is more important than ever. It has been shown in other surgical specialties at similar institutions that the residents' ACGME case log may not accurately reflect their operative experience. What barriers may influence this remains unclear. As the only objective measure of resident operative experience, an accurate case log is paramount in representing one's operative experience. This study aims to determine the accuracy of procedural coding by general surgical residents at a single institution. Data were collected from 2 consecutive graduating classes of surgical residents' ACGME case logs from 2008 to 2014. A total of 5799 entries from 7 residents were collected. The CPT codes entered by residents were compared to departmental billing records submitted by the attending surgeon for each procedure. Assigned CPT codes by institutional American Academy of Professional Coders certified abstract coders were considered the "gold standard." A total of 4356 (75.12%) of 5799 entries were identified in billing records. Excel 2010 and SAS 9.3 were used for analysis. In the event of multiple codes for the same patient, any match between resident codes and billing record codes was considered a "correct" entry. A 4-question survey was distributed to all current general surgical residents at our institution for feedback on coding habits, limitations to accurate coding, and opinions on ACGME case log representation of their operative experience. All 7 residents had a low percentage of correctly entered CPT codes. The overall accuracy proportion for all residents was 52.82% (range: 43.32%-60.07%). Only 1 resident showed significant improvement in accuracy during his/her training (p = 0

Independent validation testing of the FLAME computer code, Version 1.0

International Nuclear Information System (INIS)

Martian, P.; Chung, J.N.

1992-07-01

Independent testing of the FLAME computer code, Version 1.0, was conducted to determine if the code is ready for use in hydrological and environmental studies at Department of Energy sites. This report describes the technical basis, approach, and results of this testing. Validation tests, (i.e., tests which compare field data to the computer generated solutions) were used to determine the operational status of the FLAME computer code and were done on a qualitative basis through graphical comparisons of the experimental and numerical data. These tests were specifically designed to check: (1) correctness of the FORTRAN coding, (2) computational accuracy, and (3) suitability to simulating actual hydrologic conditions. This testing was performed using a structured evaluation protocol which consisted of: (1) independent applications, and (2) graduated difficulty of test cases. Three tests ranging in complexity from simple one-dimensional steady-state flow field problems under near-saturated conditions to two-dimensional transient flow problems with very dry initial conditions

Contributions to the validation of the ASTEC V1 code

International Nuclear Information System (INIS)

Constantin, Marin; Rizoiu, Andrei; Turcu, Ilie

2004-01-01

In the frame of PHEBEN2 project (Validation of the severe accidents codes for applications to nuclear power plants, based on the PHEBUS FP experiments), a project developed within the EU research Frame Program 5 (FP5), the INR-Pitesti's team has received the task of determining the ASTEC code sensitivity. The PHEBEN2 project has been initiated in 1998 and gathered 13 partners from 6 EU member states. To the project 4 partners from 3 candidate states (Hungary, Bulgaria and Romania) joined later. The works were contracted with the European Commission (under FIKS-CT1999-00009 contract) that supports financially the research effort up to about 50%. According to the contract provisions, INR's team participated in developing the Working Package 1 (WP1) which refers to validation of the integral computation codes that use the PHOEBUS experimental data and the Working Package 3 (WP3) referring to the evaluation of the codes to be applied in nuclear power plants for risk evaluation, nuclear safety margin evaluation and determination/evaluation of the measures to be adopted in case of severe accident. The present work continues the efforts to validate preliminarily the ASTEC code. Focused are the the stand-alone sensitivity analyses applied to two most important modules of the code, namely DIVA and SOPHAEROS

48 CFR 1.105-1 - Publication and code arrangement.

Science.gov (United States)

2010-10-01

... 48 Federal Acquisition Regulations System 1 2010-10-01 2010-10-01 false Publication and code arrangement. 1.105-1 Section 1.105-1 Federal Acquisition Regulations System FEDERAL ACQUISITION REGULATION GENERAL FEDERAL ACQUISITION REGULATIONS SYSTEM Purpose, Authority, Issuance 1.105-1 Publication and code...

Verification of the AZNHEX code v.1.4 with MCNP6 for different reference cases; Verificacion del codigo AZNHEX v.1.4 con MCNP6 para diferentes casos de referencia

Energy Technology Data Exchange (ETDEWEB)

Galicia A, J.; Francois L, J. L.; Bastida O, G. E. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, Circuito Exterior s/n, 04510 Ciudad de Mexico (Mexico); Del Valle G, E., E-mail: jgaliciaa87@gmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, 07738 Ciudad de Mexico (Mexico)

2017-09-15

The codes that make up the AZTLAN platform (AZTHECA, AZTRAN, AZKIND and AZNHEX) are currently in the testing phase simulating a variety of nuclear reactor assemblies and cores to compare and validate the results obtained for a particular case, with codes globally used in the nuclear area such as CASMO, Serpent and MCNP. The objective of this work is to continue improving the future versions of the codes of the AZTLAN platform so that accurate and reliable results can be obtained for the user. To test the current version of the AZNHEX code, 3 cases were taken into account, the first being the simulation of a VVER-440 reactor assembly; for the second case, the assembly of a fast reactor cooled with helium was simulated and for the third case it was decided to take up the case of the core of a fast reactor cooled with sodium, this because the previous versions of AZNHEX did not show adequate results and, in addition, they presented a considerable amount of limitations. The comparison and validation of the results (neutron multiplication factor, radial power, radial flow, axial power) for these three cases were made using the code MCNP6. The results obtained show that this version of AZNHEX produces values of the neutron multiplication factor and the neutron and power flow distributions very close to those of MCNP6. (Author)

ANGRA-1 neutron kinetics model at BOL using WIMSD-5B and PARCS V2.7 codes

International Nuclear Information System (INIS)

Hamers, Adolfo R.; Reis, Patricia A.L.; Rodrigues, Thiago D.A.; Pereira, Claubia; Costa, Antonella L.

2015-01-01

A steady-state neutron kinetics model of the Angra-1 NPP at BOL (Beginning Of Life) has been developed with the PARCS V2.7 neutron diffusion code. The information of the burnable poison rods, fuel enrichments and control rod banks distributions within the core have been taken from the Angra-1 FSAR (Final Safety Analysis Report) and implemented in the model. The macroscopic cross sections for the fast and thermal neutron groups have been calculated with the WIMSD-5B lattice cell code. The cross sections were obtained for the rodded and unrodded cases for each composition in the core. In order to establish the initial steady-state, an eigenvalue was made with the PARCS V2.7 code for three steady-state scenario cases reported at the FSAR; a K eff of 1.0733 was obtained for the unrodded case, K eff of 1.0718 for a 24% of bank D inserted case and K eff of 0.8512 for the full rodded case. The normalized core power density distributions were obtained and compared with the corresponding FSAR case. (author)

Differentially Encoded LDPC Codes—Part I: Special Case of Product Accumulate Codes

Directory of Open Access Journals (Sweden)

(Tiffany JingLi

2008-01-01

Full Text Available Part I of a two-part series investigates product accumulate codes, a special class of differentially-encoded low density parity check (DE-LDPC codes with high performance and low complexity, on flat Rayleigh fading channels. In the coherent detection case, Divsalar's simple bounds and iterative thresholds using density evolution are computed to quantify the code performance at finite and infinite lengths, respectively. In the noncoherent detection case, a simple iterative differential detection and decoding (IDDD receiver is proposed and shown to be robust for different Doppler shifts. Extrinsic information transfer (EXIT charts reveal that, with pilot symbol assisted differential detection, the widespread practice of inserting pilot symbols to terminate the trellis actually incurs a loss in capacity, and a more efficient way is to separate pilots from the trellis. Through analysis and simulations, it is shown that PA codes perform very well with both coherent and noncoherent detections. The more general case of DE-LDPC codes, where the LDPC part may take arbitrary degree profiles, is studied in Part II Li 2008.

MARG1D: One dimensional outer region matching data code

International Nuclear Information System (INIS)

Tokuda, Shinji; Watanabe, Tomoko.

1995-08-01

A code MARG1D has been developed which computes outer region matching data of the one dimensional Newcomb equation. Matching data play an important role in the resistive (and non ideal) Magneto-hydrodynamic (MHD) stability analysis in a tokamak plasma. The MARG1D code computes matching data by using the boundary value method or by the eigenvalue method. Variational principles are derived for the problems to be solved and a finite element method is applied. Except for the case of marginal stability, the eigenvalue method is equivalent to the boundary value method. However, the eigenvalue method has the several advantages: it is a new method of ideal MHD stability analysis for which the marginally stable state can be identified, and it guarantees numerical stability in computing matching data close to marginal stability. We perform detailed numerical experiments for a model equation with analytical solutions and for the Newcomb equation in the m=1 mode theory. Numerical experiments show that MARG1D code gives the matching data with numerical stability and high accuracy. (author)

Code-Switching: L1-Coded Mediation in a Kindergarten Foreign Language Classroom

Science.gov (United States)

Lin, Zheng

2012-01-01

This paper is based on a qualitative inquiry that investigated the role of teachers' mediation in three different modes of coding in a kindergarten foreign language classroom in China (i.e. L2-coded intralinguistic mediation, L1-coded cross-lingual mediation, and L2-and-L1-mixed mediation). Through an exploratory examination of the varying effects…

Incommensurate phases in the improper ferroelastic MgGeF sub 6 centre dot 6H sub 2 O:Mn sup 2 sup + studied by means of EPR

CERN Document Server

Skrylnik, P G

2002-01-01

The results of an EPR study of the inhomogeneous phases existing in the temperature interval T sub C = 311.0 +- 0.3 K < T < T sub i sub 1 = 403 +- 0.3 K in improper ferroelastic crystals of MgGeF sub 6 centre dot 6H sub 2 O:Mn sup 2 sup + are presented. On the basis of the analysis of the temperature and angle dependences of the experimental parameters and numerical calculations, the conclusion has been drawn that at T sub i sub 1 the crystals considered undergo a transition to a structurally modulated phase and the order parameter of this transition may be the angle of the Mg[H sub 2 O] sub 6 sup 2 sup + octahedra rotation around the crystal C sub 3 -axis. From T sub i sub 1 to T sub C the modes of the modulated phase follow according to a completely classical scenario for incommensurate crystals: the origin of the incommensurate structure with plane-wave modulation at T sub i sub 1 , the appearance of structural phase solitons below T sub i sub 2 = 380 +- 0.3 K and decrease of the soliton density to v...

Hermitian self-dual quasi-abelian codes

Directory of Open Access Journals (Sweden)

Herbert S. Palines

2017-12-01

Full Text Available Quasi-abelian codes constitute an important class of linear codes containing theoretically and practically interesting codes such as quasi-cyclic codes, abelian codes, and cyclic codes. In particular, the sub-class consisting of 1-generator quasi-abelian codes contains large families of good codes. Based on the well-known decomposition of quasi-abelian codes, the characterization and enumeration of Hermitian self-dual quasi-abelian codes are given. In the case of 1-generator quasi-abelian codes, we offer necessary and sufficient conditions for such codes to be Hermitian self-dual and give a formula for the number of these codes. In the case where the underlying groups are some $p$-groups, the actual number of resulting Hermitian self-dual quasi-abelian codes are determined.

Verification of the geostatistical inference code INFERENS, Version 1.1, and demonstration using data from Finnsjoen

International Nuclear Information System (INIS)

Geier, J.

1993-06-01

This report describes preliminary verification and demonstration of the geostatistical inference code, INFERENS Version 1.1. This code performs regularization of packer test conductivities, and iterative generalized least-squares estimation (IGLSE) of nested covariance models and spatial trends for the regularized data. Cross-validation is used to assess the quality of the estimated models in terms of statistics for the kriging errors. The code includes a capability to generate synthetic datasets for a given configuration of packer tests; this capability can be used for verification exercises and numerical experiments to aid in the design of packer testing programs. The report presents the results of a set of verification test cases. The test cases were designed to test the ability of INFERENS 1.1 to estimate the parameters of a variety of covariance models, with or without trends. This was done using synthetic datasets. This report also describes an application of INFERENS 1.1 to the dataset from the Finnsjoen site. The results are roughly similar to those obtained previously by Norman (1992a) using INFERENS 1.0, for the comparable cases. The actual numerical results are different, which may be due to changes in the fitting algorithms, and differences in how the lag pairs are divided into lag classes. The demonstrations confirm the result previously obtained by Norman, that the fitted horizontally isotropic models are less good, in terms of their cross-validation statistics, than the corresponding isotropic models. The use of nested covariance models is demonstrated to give visually improved fits to the sample semivariograms, at both short and long lag distances. However, despite the good match to the semivariograms, the nested models obtained are not better than the simple models, in terms of cross-validation statistics

The arbitrary order design code Tlie 1.0

International Nuclear Information System (INIS)

Zeijts, J. van; Neri, Filippo

1993-01-01

We describe the arbitrary order charged particle transfer map code TLIE. This code is a general 6D relativistic design code with a MAD compatible input language and among others implements user defined functions and subroutines and nested fitting and optimization. First we describe the mathematics and physics in the code. Aside from generating maps for all the standard accelerator elements we describe an efficient method for generating nonlinear transfer maps for realistic magnet models. We have implemented the method to arbitrary order in our accelerator design code for cylindrical current sheet magnets. We also have implemented a self-consistent space-charge approach as in CHARLIE. Subsequently we give a description of the input language and finally, we give several examples from productions run, such as cases with stacked multipoles with overlapping fringe fields. (Author)

BRT-1 code for IBM 370/135

International Nuclear Information System (INIS)

Preda, I.

1976-01-01

BRT-1 is a transport code to obtain the thermal neutrons spectrum, point dependent, in one reactor cell. The code BRT-1 described in this paper, is the code BRT-1, written in FORTRAN IV language for the computer UNIVAC 1108 with CSCX operating system, converted for the computer IBM 370/135 disk operating system. (author)

A Case for Dynamic Reverse-code Generation to Debug Non-deterministic Programs

Directory of Open Access Journals (Sweden)

Jooyong Yi

2013-09-01

Full Text Available Backtracking (i.e., reverse execution helps the user of a debugger to naturally think backwards along the execution path of a program, and thinking backwards makes it easy to locate the origin of a bug. So far backtracking has been implemented mostly by state saving or by checkpointing. These implementations, however, inherently do not scale. Meanwhile, a more recent backtracking method based on reverse-code generation seems promising because executing reverse code can restore the previous states of a program without state saving. In the literature, there can be found two methods that generate reverse code: (a static reverse-code generation that pre-generates reverse code through static analysis before starting a debugging session, and (b dynamic reverse-code generation that generates reverse code by applying dynamic analysis on the fly during a debugging session. In particular, we espoused the latter one in our previous work to accommodate non-determinism of a program caused by e.g., multi-threading. To demonstrate the usefulness of our dynamic reverse-code generation, this article presents a case study of various backtracking methods including ours. We compare the memory usage of various backtracking methods in a simple but nontrivial example, a bounded-buffer program. In the case of non-deterministic programs such as this bounded-buffer program, our dynamic reverse-code generation outperforms the existing backtracking methods in terms of memory efficiency.

Structural evaluation method for class 1 vessels by using elastic-plastic finite element analysis in code case of JSME rules on design and construction

International Nuclear Information System (INIS)

Asada, Seiji; Hirano, Takashi; Nagata, Tetsuya; Kasahara, Naoto

2008-01-01

A structural evaluation method by using elastic-plastic finite element analysis has been developed and published as a code case of Rules on Design and Construction for Nuclear Power Plants (The First Part: Light Water Reactor Structural Design Standard) in the JSME Codes for Nuclear Power Generation Facilities. Its title is 'Alternative Structural Evaluation Criteria for Class 1 Vessels Based on Elastic-Plastic Finite Element Analysis' (NC-CC-005). This code case applies elastic-plastic analysis to evaluation of such failure modes as plastic collapse, thermal ratchet, fatigue and so on. Advantage of this evaluation method is free from stress classification, consistently use of Mises stress and applicability to complex 3-dimensional structures which are hard to be treated by the conventional stress classification method. The evaluation method for plastic collapse has such variation as the Lower Bound Approach Method, Twice-Elastic-Slope Method and Elastic Compensation Method. Cyclic Yield Area (CYA) based on elastic analysis is applied to screening evaluation of thermal ratchet instead of secondary stress evaluation, and elastic-plastic analysis is performed when the CYA screening criteria is not satisfied. Strain concentration factors can be directly calculated based on elastic-plastic analysis. (author)

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

Directory of Open Access Journals (Sweden)

Bih-Chyun Yeh

2016-01-01

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

A controlled evaluation of case clinical effect coding by poison center specialists for detection of WMD scenarios.

Science.gov (United States)

Beuhler, Michael C; Wittler, Mary A; Ford, Marsha; Dulaney, Anna R

2011-08-01

Many public health entities employ computer-based syndromic surveillance to monitor for aberrations including possible exposures to weapons of mass destruction (WMD). Often, this is done by screening signs and symptoms reported for cases against syndromic definitions. Poison centers (PCs) may offer significant contributions to public health surveillance because of their detailed clinical effect data field coding and real-time data entry. Because improper clinical effect coding may impede syndromic surveillance, it is important to assess this accuracy for PCs. An AAPCC-certified regional PC assessed the accuracy of clinical effect coding by specialists in poison information (SPIs) listening to audio recordings of standard cases. Eighteen different standardized cases were used, consisting of six cyanide, six botulism, and six control cases. Cases were scripted to simulate clinically relevant telephone conversations and converted to audio recordings. Ten SPIs were randomly selected from the center's staff to listen to and code case information from the recorded cases. Kappa scores and the percentage of correctly coding a present clinical effect were calculated for individual clinical effects summed over all test cases along with corresponding 95% confidence intervals. The rate of the case coding by the SPIs triggering the PC's automated botulism and cyanide alerts was also determined. The kappa scores and the percentage of correctly coding a present clinical effect varied depending on the specific clinical effect, with greater accuracy observed for the clinical effects of vomiting and agitation/irritability, and poor accuracy observed for the clinical effects of visual defect and anion gap increase. Lack of correct coding resulted in only 60 and 86% of the cases that met the botulism and cyanide surveillance definitions, respectively, triggering the corresponding alert. There was no difference observed in the percentage of coding a present clinical effect between

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)

Relaxation of inservice test frequency requirement for Kori 1 ASME code pumps

International Nuclear Information System (INIS)

Sohn, Gap Heon; Choi, Hae Yoon; Min, Kyung Sung; Rim, Nam Jin

1994-08-01

The objective of this investigation is to evaluate the technical and regulational requirements to justify the relaxation of the test frequency of Kori 1 pumps through reviewing the related rules and codes and standards, technical specifications of Kori 1 and other similar plants, standard technical specifications, research results for tech. spec. improvements and site test records. It is concluded that the relaxation of test frequency to quarterly be justified based on the conformance with rules and codes and standard, quarterly test cases in similar plants and standard tech. spec., recommendations of research result and stable site test records. (Author) 16 refs., 26 figs., 13 tabs

Code for ethical international recruitment practices: the CGFNS alliance case study.

Science.gov (United States)

Shaffer, Franklin A; Bakhshi, Mukul; Dutka, Julia To; Phillips, Janice

2016-06-30

Projections indicate a global workforce shortage of approximately 4.3 million across the health professions. The need to ensure an adequate supply of health workers worldwide has created a context for the increased global migration of these professionals. The global trend in the migration of health professionals has given rise to the international recruitment industry to facilitate the passage of health workers from source to destination countries. This is particularly the case in the United States, where the majority of immigrant health professionals have come by way of the recruiting industry. This industry is largely unregulated in the United States as well as in many other countries, for which voluntary codes have been used as a means to increase transparency of the recruitment process, shape professional conduct, and mitigate harm to foreign-educated health workers. The CGFNS Alliance case study presented herein describes a multi-stakeholder effort in the United States to promote ethical recruitment practices. Such codes not only complement the WHO Global Code of Practice but are necessary to maximize the impact of these global standards on local settings. This case study offers both a historical perspective and a conceptual framework for examining the multiplicity of factors affecting the migration of human resources for health. The lessons learned provide critical insights into the factors pertaining to the relevancy and effectiveness of the WHO Code from the perspectives of both source and destination countries. This study provides a conceptual model for examining the usefulness of the WHO Code as well as how best to ensure its viability, sustainability, relevancy, and effectiveness in the global environment. This case study concludes with recommendations for evolving business models that need to be in place to strengthen the effectiveness of the WHO Code in the marketplace and to ensure its impact on the international recruitment industry in advancing

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

International Nuclear Information System (INIS)

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

1990-02-01

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

FARO base case post-test analysis by COMETA code

Energy Technology Data Exchange (ETDEWEB)

Annunziato, A.; Addabbo, C. [Joint Research Centre, Ispra (Italy)

1995-09-01

The paper analyzes the COMETA (Core Melt Thermal-Hydraulic Analysis) post test calculations of FARO Test L-11, the so-called Base Case Test. The FARO Facility, located at JRC Ispra, is used to simulate the consequences of Severe Accidents in Nuclear Power Plants under a variety of conditions. The COMETA Code has a 6 equations two phase flow field and a 3 phases corium field: the jet, the droplets and the fused-debris bed. The analysis shown that the code is able to pick-up all the major phenomena occurring during the fuel-coolant interaction pre-mixing phase.

The use of case tools in OPG safety analysis code qualification

International Nuclear Information System (INIS)

Pascoe, J.; Cheung, A.; Westbye, C.

2001-01-01

Ontario Power Generation (OPG) is currently qualifying its critical safety analysis software. The software quality assurance (SQA) framework is described. Given the legacy nature of much of the safety analysis software the reverse engineering methodology has been adopted. The safety analysis suite of codes was developed over a period of many years to differing standards of quality and had sparse or incomplete documentation. Key elements of the reverse engineering process require recovery of design information from existing coding. This recovery, if performed manually, could represent an enormous effort. Driven by a need to maximize productivity and enhance the repeatability and objectivity of software qualification activities the decision was made to acquire or develop and implement Computer Aided Software Engineering (CASE) tools. This paper presents relevant background information on CASE tools and discusses how the OPG SQA requirements were used to assess the suitability of available CASE tools. Key findings from the application of CASE tools to the qualification of the OPG safety analysis software are discussed. (author)

PM1 steganographic algorithm using ternary Hamming Code

Directory of Open Access Journals (Sweden)

Kamil Kaczyński

2015-12-01

Full Text Available PM1 algorithm is a modification of well-known LSB steganographic algorithm. It has increased resistance to selected steganalytic attacks and increased embedding efficiency. Due to its uniqueness, PM1 algorithm allows us to use of larger alphabet of symbols, making it possible to further increase steganographic capacity. In this paper, we present the modified PM1 algorithm which utilizies so-called syndrome coding and ternary Hamming code. The modified algorithm has increased embedding efficiency, which means fewer changes introduced to carrier and increased capacity.[b]Keywords[/b]: steganography, linear codes, PM1, LSB, ternary Hamming code

BERMUDA-1DG: a one-dimensional photon transport code

International Nuclear Information System (INIS)

Suzuki, Tomoo; Hasegawa, Akira; Nakashima, Hiroshi; Kaneko, Kunio.

1984-10-01

A one-dimensional photon transport code BERMUDA-1DG has been developed for spherical and infinite slab geometries. The purpose of development is to equip the function of gamma rays calculation for the BERMUDA code system, which was developed by 1983 only for neutron transport calculation as a preliminary version. A group constants library has been prepared for 30 nuclides, and it now consists of the 36-group total cross sections and secondary gamma ray yields by the 120-group neutron flux. For the Compton scattering, group-angle transfer matrices are accurately obtained by integrating the Klein-Nishina formula taking into account the energy and scattering angle correlation. The pair production cross sections are now calculated in the code from atomic number and midenergy of each group. To obtain angular flux distribution, the transport equation is solved in the same way as in case of neutron, using the direct integration method in a multigroup model. Both of an independent gamma ray source problem and a neutron-gamma source problem are possible to be solved. This report is written as a user's manual with a brief description of the calculational method. (author)

Validation of ASTEC v1.0 computer code against FPT2 test

International Nuclear Information System (INIS)

Mladenov, I.; Tusheva, P.; Kalchev, B.; Dimov, D.; Ivanov, I.

2005-01-01

The aim of the work is by various nodalization schemes of the model to investigate the ASTEC v1.0 computer code sensitivity and to validate the code against PHEBUS - FPT2 experiment. This code is used for severe accident analysis. The aim corresponds to the main technical objective of the experiment which is to contribute to the validation of models and computer codes to be used for the calculation of the source term in case of a severe accident in a Light Water Reactor. The objective's scope of the FPT2 is large - separately for the bundle, the experimental circuit and the containment. Additional objectives are to characterize aerosol sizing and deposition processes, and also potential FP poisoning effects on hydrogen recombiner coupons exposed to containment atmospheric conditions representative of a LWR severe accident. The analyses of the results of the performed calculations show a good accordance with the reference case calculations, and then with the experimental data. Some differences in the calculations for the thermal behavior appear locally during the oxidation phase and the heat-up phase. There is very good confirmation regarding the volatile and semi-volatile fission products release from the fuel pellets. Important for analysis of the process is the final axial distribution of the mass of fuel relocation obtained at the end of the calculation

Evaluation of Geometric Progression (GP Buildup Factors using MCNP Codes (MCNP6.1 and MCNP5-1.60

Directory of Open Access Journals (Sweden)

Kim Kyung-O

2016-01-01

Full Text Available The gamma-ray buildup factors of three-dimensional point kernel code (QAD-CGGP are re-evaluated by using MCNP codes (MCNP6.1 and MCNPX5-1.60 and ENDF/B-VI.8 photoatomic data, which cover an energy range of 0.015–15 MeV and an iron thickness of 0.5–40 Mean Free Path (MFP. These new data are fitted to the Geometric Progression (GP fitting function and are then compared with ANS standard data equipped with QAD-CGGP. In addition, a simple benchmark calculation was performed to compare the QAD-CGGP results applied with new and existing buildup factors based on the MCNP codes. In the case of the buildup factors of low-energy gamma-rays, new data are evaluated to be about 5% higher than the existing data. In other cases, these new data present a similar trend based on the specific penetration depth, while existing data continuously increase beyond that depth. In a simple benchmark, the calculations using the existing data were slightly underestimated compared to the reference data at a deep penetration depth. On the other hand, the calculations with new data were stabilized with an increasing penetration depth, despite a slight overestimation at a shallow penetration depth.

Validation of the ATHLET-code 2.1A by calculation of the ECTHOR experiment; Validierung des ATHLET-Codes 2.1A anhand des Einzeleffekt-Tests ECTHOR

Energy Technology Data Exchange (ETDEWEB)

Jung, Andreas; Sarkadi, Peter; Schaffrath, Andreas [TUEV NORD SysTec GmbH und Co. KG, Hamburg (Germany)

2010-05-15

Before a numerical code (e.g. ATHLET) is used for simulation of physical phenomena being new or unknown for the code and/or the user, the user ensures the applicability of the code and his own experience of handling with it by means of a so-called validation. Parametric studies with the code are executed for that matter and the results have to be compared with verified experimental data. Corresponding reference values are available in terms of so-called single-effect-tests (e.g. ECTHOR). In this work the system-code ATHLET Mod. 2.1 Cycle A is validated by post test calculation of the ECTHOR experiment due to the above named aspects. With the ECTHOR-tests the clearing of a water-filled model of a loop seal by means of an air-stream was investigated including momentum exchange at the phase interface under adiabatic and atmospheric conditions. The post test calculations show that the analytical results meet the experimental data within the reproducibility of the experiments. Further findings of the parametric studies are: - The experimental results obtained with the system water-air (ECTHOR) can be assigned to a water-steam-system, if the densities of the phases are equal in both cases. - The initial water level in the loop seal has no influence on the results as long as the gas mass flow is increased moderately. - The loop seal is appropriately nodalized if the mean length of the control volumes accords approx. 1.5 tim es the hydraulic pipe diameter. (orig.)

Validation of the ATHLET-code 2.1A by calculation of the ECTHOR experiment; Validierung des ATHLET-Codes 2.1A anhand des Einzeleffekt-Tests ECTHOR

Energy Technology Data Exchange (ETDEWEB)

Jung, Andreas; Sarkadi, Peter; Schaffrath, Andreas [TUEV NORD SysTec GmbH und Co. KG, Hamburg (Germany)

2010-06-15

Before a numerical code (e.g. ATHLET) is used for simulation of physical phenomena being new or unknown for the code and/or the user, the user ensures the applicability of the code and his own experience of handling with it by means of a so-called validation. Parametric studies with the code are executed for that matter und the results have to be compared with verified experimental data. Corresponding reference values are available in terms of so-called single-effect-tests (e.g. ECTHOR). In this work the system-code ATHLET Mod. 2.1 Cycle A is validated by post test calculation of the ECTHOR experiment due to the above named aspects. With the ECTHOR-tests the clearing of a water-filled model of a loop seal by means of an air-stream was investigated including momentum exchange at the phase interface under adiabatic and atmospheric conditions. The post test calculations show that the analytical results meet the experimental data within the reproducibility of the experiments. Further findings of the parametric studies are: - The experimental results obtained with the system water-air (ECTHOR) can be assigned to a water-steam-system, if the densities of the phases are equal in both cases. - The initial water level in the loop seal has no influence on the results as long as the gas mass flow is increased moderately. - The loop seal is appropriately nodalized if the mean length of the control volumes accords approx. 1.5 times the hydraulic pipe diameter. (orig.)

B{sup ¯}{sub s}→K semileptonic decay from an Omnès improved constituent quark model

Energy Technology Data Exchange (ETDEWEB)

Albertus, C. [Departamento de Física Atómica, Nuclear y Molecular e Instituto Carlos I de Física Teórica y Computacional, Universidad de Granada, Avenida de Fuentenueva s/n, E-18071 Granada (Spain); Hernández, E. [Departamento de Física Fundamental e IUFFyM, Universidad de Salamanca, Plaza de la Merced s/n, E-37008 Salamanca (Spain); Hidalgo-Duque, C.; Nieves, J. [Instituto de Física Corpuscular (IFIC), Centro Mixto CSIC-Universidad de Valencia, Institutos de Investigación de Paterna, Apartado 22085, E-46071 Valencia (Spain)

2014-11-10

We study the f{sup +} form factor for the semileptonic B{sup ¯}{sub s}→K{sup +}ℓ{sup −}ν{sup ¯}{sub ℓ} decay in a constituent quark model. The valence quark estimate is supplemented with the contribution from the B{sup ¯⁎} pole that dominates the high q{sup 2} region. We use a multiply-subtracted Omnès dispersion relation to extend the quark model predictions from its region of applicability near q{sub max}{sup 2}=(M{sub B{sub s}}−M{sub K}){sup 2}∼23.75 GeV{sup 2} to all q{sup 2} values accessible in the physical decay. To better constrain the dependence of f{sup +} on q{sup 2}, we fit the subtraction constants to a combined input from previous light cone sum rule by Duplancic and Melic (2008) [11] and the present quark model results. From this analysis, we obtain Γ(B{sup ¯}{sub s}→K{sup +}ℓ{sup −}ν{sup ¯}{sub ℓ})=(5.47{sub −0.46}{sup +0.54})|V{sub ub}|{sup 2}×10{sup −9} MeV, which is about 10% and 20% higher than the predictions based on Lattice QCD and QCD light cone sum rules respectively. The former predictions, for both the form factor f{sup +}(q{sup 2}) and the differential decay width, lie within the 1σ band of our estimated uncertainties for all q{sup 2} values accessible in the physical decay, except for a quite small region very close to q{sub max}{sup 2}. Differences with the light cone sum results for the form factor f{sup +} are larger than 20% in the region above q{sup 2}=15 GeV{sup 2}.

Assessment of the computer code COBRA/CFTL

International Nuclear Information System (INIS)

Baxi, C.B.; Burhop, C.J.

1981-07-01

The COBRA/CFTL code has been developed by Oak Ridge National Laboratory (ORNL) for thermal-hydraulic analysis of simulated gas-cooled fast breeder reactor (GCFR) core assemblies to be tested in the core flow test loop (CFTL). The COBRA/CFTL code was obtained by modifying the General Atomic code COBRA*GCFR. This report discusses these modifications, compares the two code results for three cases which represent conditions from fully rough turbulent flow to laminar flow. Case 1 represented fully rough turbulent flow in the bundle. Cases 2 and 3 represented laminar and transition flow regimes. The required input for the COBRA/CFTL code, a sample problem input/output and the code listing are included in the Appendices

Transient cases analyses of the TRIGA IPR-R1 using thermal hydraulic and neutron kinetic coupled codes

Energy Technology Data Exchange (ETDEWEB)

Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Scari, Maria E., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: melizabethscari@yahoo.com [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Belo Horizonte (Brazil); Miro, Rafael; Verdu, Gumersindo, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear

2015-07-01

Simulations and analyses of nuclear reactors have been improved by utilization of coupled thermal-hydraulic (TH) and neutron kinetics (NK) system codes especially to simulate transients that involve strong feedback effects between NK and TH. The TH-NK coupling technique was initially developed and used to simulate the behavior of power reactors; however, several coupling methodologies are now being applied for research reactors. This work presents the coupling methodology application between RELAP5 and PARCS codes using as a model the TRIGA IPR-R1 research reactor. Analyses of steady state and transient conditions and comparisons with results from simulations using only the RELAP5 code are being presented in this paper. (author)

Technical basis for the extension of ASME Code Case N-494 for assessment of austenitic piping

International Nuclear Information System (INIS)

Bloom, J.M.

1995-01-01

In 1990, the ASME Boiler and Pressure Vessel Code for Nuclear Components approved Code Case N-494 as an alternative procedure for evaluating laws in Light Water Reactor alterative procedure for evaluating flaws in Light Water Reactor (LWR) ferritic piping. The approach is an alternative to Appendix H of the ASME Code and alloys the user to remove some unnecessary conservatism in the existing procedure by allowing the use of pipe specific material properties. The Code Case is an implementation of the methodology of the Deformation Plasticity Failure Assessment diagram (DPFAD). The key ingredient in the application of DPFAD is that the material stress-strain curve must be in the format of a simple power law hardening stress-strain curve such as the Ramberg-Osgood (R-O) model. Ferritic materials can be accurately fit by the R-O model and, therefore, it was natural to use the DPFAD methodology for the assessment of LWR ferritic piping. An extension of Code Case N-494 to austenitic piping required a modification of the existing DPFAD methodology. The Code Case N-494 approach was revised using the PWFAD procedure in the same manner as in the development of the original N-494 approach for ferritic materials. A lower bound stress-strain curve was used to generate a PWFAD curve for the geometry of a part-through wall circumferential flaw in a cylinder under tension. Earlier work demonstrated that a cylinder under axial tension with a 50% flaw depth, 90 degrees in circumference, and radius to thickness of 10, produced a lower bound FAD curve. Validation of the new proposed Code Case procedure for austenitic piping was performed using actual pipe test data. Using the lower bound PWFAD curve, pipe test results were conservatively predicted. The resultant development of ht PWFAD curve for austenitic piping led to a revision of Code Case N-494 to include a procedure for assessment of flaws in austenitic piping

Decoding of concatenated codes with interleaved outer codes

DEFF Research Database (Denmark)

Justesen, Jørn; Høholdt, Tom; Thommesen, Christian

2004-01-01

Recently Bleichenbacher et al. proposed a decoding algorithm for interleaved (N, K) Reed-Solomon codes, which allows close to N-K errors to be corrected in many cases. We discuss the application of this decoding algorithm to concatenated codes.......Recently Bleichenbacher et al. proposed a decoding algorithm for interleaved (N, K) Reed-Solomon codes, which allows close to N-K errors to be corrected in many cases. We discuss the application of this decoding algorithm to concatenated codes....

26 CFR 1.167(i)-1 - Depreciation of improvements in the case of mines, etc.

Science.gov (United States)

2010-04-01

... 26 Internal Revenue 2 2010-04-01 2010-04-01 false Depreciation of improvements in the case of... and Corporations § 1.167(i)-1 Depreciation of improvements in the case of mines, etc. Property used in... depreciation provided in section 611 shall be treated for all purposes of the Code as if it were property...

Radiant Energy Measurements from a Scaled Jet Engine Axisymmetric Exhaust Nozzle for a Baseline Code Validation Case

Science.gov (United States)

Baumeister, Joseph F.

1994-01-01

A non-flowing, electrically heated test rig was developed to verify computer codes that calculate radiant energy propagation from nozzle geometries that represent aircraft propulsion nozzle systems. Since there are a variety of analysis tools used to evaluate thermal radiation propagation from partially enclosed nozzle surfaces, an experimental benchmark test case was developed for code comparison. This paper briefly describes the nozzle test rig and the developed analytical nozzle geometry used to compare the experimental and predicted thermal radiation results. A major objective of this effort was to make available the experimental results and the analytical model in a format to facilitate conversion to existing computer code formats. For code validation purposes this nozzle geometry represents one validation case for one set of analysis conditions. Since each computer code has advantages and disadvantages based on scope, requirements, and desired accuracy, the usefulness of this single nozzle baseline validation case can be limited for some code comparisons.

Development of olmesartan medoxomil optimized nanosuspension using the Box-Behnken design to improve oral bioavailability.

Science.gov (United States)

Nagaraj, K; Narendar, D; Kishan, V

2017-07-01

The aim of the present investigation was to enhance the oral bioavailability of olmesartan medoxomil by improving its solubility and dissolution rate by preparing nanosuspension (OM-NS), using the Box-Behnken design. In this, four factors were evaluated at three levels. Independent variables include: concentration of drug (X 1 ), concentration of surfactant (X 2 ), concentration of polymer (X 3 ) and number of homogenization cycles (X 4 ). Based on preliminary studies, the size (Y 1 ), zeta potential (ZP) (Y 2 ) and % drug release at 5 min (Y 3 ) were chosen as dependent responses. OM-NS was prepared by high pressure homogenization method. The size, PDI, ZP, assay, in vitro release and morphology of OM-NS were characterized. Further, the pharmacokinetic (PK) behavior of OM-NS was evaluated in male wistar rats. Statistically optimized OM-NS formulation exhibited mean particle size of 492 nm, ZP of -27.9 mV and 99.29% release in 5 min. OM-NS showed more than four times increase in its solubility than pure OM. DSC and XRD analyses indicated that the drug incorporated into OM-NS was in amorphous form. The morphology of OM-NS was found to be nearly spherical with high dispersity by scanning electron microscopic studies. The PK results showed that OM lyophilized nanosuspension (NS) exhibited improved PK properties compared to coarse powder suspension and marketed tablet powder suspension (TS). Oral bioavailability of lyophilized NS was increased by 2.45 and 2.25 folds when compared to marketed TS and coarse powder suspension, respectively. Results of this study lead to conclusion that NS approach was effective in preparing OM formulations with enhanced dissolution and improved oral bioavailability.

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

International Nuclear Information System (INIS)

Koo, Gyeong Hoi; Yoo, B.

2001-02-01

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

Governance codes: facts or fictions? a study of governance codes in colombia1,2

Directory of Open Access Journals (Sweden)

Julián Benavides Franco

2010-10-01

Full Text Available This article studies the effects on accounting performance and financing decisions of Colombian firms after issuing a corporate governance code. We assemble a database of Colombian issuers and test the hypotheses of improved performance and higher leverage after issuing a code. The results show that the firms’ return on assets after the code introduction improves in excess of 1%; the effect is amplified by the code quality. Additionally, the firms leverage increased, in excess of 5%, when the code quality was factored into the analysis. These results suggest that controlling parties commitment to self restrain, by reducing their private benefits and/or the expropriation of non controlling parties, through the code introduction, is indeed an effective measure and that the financial markets agree, increasing the supply of funds to the firms.

Technical Review on Fitness-for-Service for Buried Pipe by ASME Code Case N-806

International Nuclear Information System (INIS)

Park, Sang Kyu; Lee, Yo Seop; So, Il Su; Lim, Bu Taek

2012-01-01

Fitness-for-Service is a useful technology to determine replacement timing, next inspection timing or in-service when nuclear power plant's buried pipes are damaged. If is possible for buried pipes to be aged by material loss, cracks and occlusion as operating time goes by. Therefore Fitness-for-Service technology for buried pipe is useful for plant industry to perform replacement and repair. Fitness-for-Service for buried pipe is studied in terms of existing code and standard for Fitness-for-Service and a current developing code case. Fitness-for-Service for buried pipe was performed according to Code Case N-806 developed by ASME (American Society of Mechanical Engineers)

The FOCON96 1.0 computer code

International Nuclear Information System (INIS)

Merle-Szeremeta, A.; Thomassin, A.

1999-01-01

The Institute of Protection and Nuclear Safety (I.P.S.N.) has developed a computer code, FOCON96 1.0 to calculate the dosimetric consequences of atmospheric radioactive releases from nuclear installations after several years of usual operation. This communication describes the principal characteristics of FOCON96 1.0 and its functionalities. The principal elements of a comparison between FOCON96 1.0 and PC-CREAM ( European computer code developed by the N.R.P.B. and answering the same criteria) are given here. (N.C.)

User's manual for the G.T.M.-1 computer code

International Nuclear Information System (INIS)

Prado-Herrero, P.

1992-01-01

This document describes the GTM-1 ( Geosphere Transport Model, release-1) computer code and is intended to provide the reader with enough detailed information in order to use the code. GTM-1 was developed for the assessment of radionuclide migration by the ground water through geologic deposits whose properties can change along the pathway.GTM-1 solves the transport equation by the finite differences method ( Crank-Nicolson scheme ). It was developped for specific use within Probabilistic System Assessment (PSA) Monte Carlo Method codes; in this context the first application of GTM-1 was within the LISA (Long Term Isolation System Assessment) code. GTM-1 is also available as an independent model, which includes various submodels simulating a multi-barrier disposal system. The code has been tested with the PSACOIN ( Probabilistic System Assessment Codes intercomparison) benchmarks exercises from PSAC User Group (OECD/NEA). 10 refs., 6 Annex., 2 tabs

Coding Partitions

Directory of Open Access Journals (Sweden)

Fabio Burderi

2007-05-01

Full Text Available Motivated by the study of decipherability conditions for codes weaker than Unique Decipherability (UD, we introduce the notion of coding partition. Such a notion generalizes that of UD code and, for codes that are not UD, allows to recover the ``unique decipherability" at the level of the classes of the partition. By tacking into account the natural order between the partitions, we define the characteristic partition of a code X as the finest coding partition of X. This leads to introduce the canonical decomposition of a code in at most one unambiguouscomponent and other (if any totally ambiguouscomponents. In the case the code is finite, we give an algorithm for computing its canonical partition. This, in particular, allows to decide whether a given partition of a finite code X is a coding partition. This last problem is then approached in the case the code is a rational set. We prove its decidability under the hypothesis that the partition contains a finite number of classes and each class is a rational set. Moreover we conjecture that the canonical partition satisfies such a hypothesis. Finally we consider also some relationships between coding partitions and varieties of codes.

The priority cases of the FUMEX-III exercises simulated with the TRANSURANUS code

International Nuclear Information System (INIS)

Boneva, S.

2011-01-01

The FUMEX-III project provides a good basis for testing common code priorities and the needs for further developments. The GAIN experiment contains results on four Gd 2 O 3 doped UO 2 rods and offers good opportunities for testing of the fuel performance codes in the case of Gd-doped fuel. A good agreement between the TRANSURANUS calculations and the measurements is achieved for the fuel and the cladding deformation. The FUMEX-III priority cases cover two rods from the GINNA reactor experiment: rod2 with fuel solid pellets, and rod4 with annular pellets and standard Zircaloy-4 cladding. Both rods were irradiated 5 cycles up to 52MWd/kgU. The simulations of the GINNA and US PWR experiments are part of the ongoing validation of the TRANSURANUS code - for different pellet design. The simulations of irradiation transients reveal the need for improving the fission gas release model, including burst release and release from the high burn-up structure

Materials and design bases issues in ASME Code Case N-47

International Nuclear Information System (INIS)

Huddleston, R.L.; Swindeman, R.W.

1993-04-01

A preliminary evaluation of the design bases (principally ASME Code Case N-47) was conducted for design and operation of reactors at elevated temperatures where the time-dependent effects of creep, creep-fatigue, and creep ratcheting are significant. Areas where Code rules or regulatory guides may be lacking or inadequate to ensure the operation over the expected life cycles for the next-generation advanced high-temperature reactor systems, with designs to be certified by the US Nuclear Regulatory Commission, have been identified as unresolved issues. Twenty-two unresolved issues were identified and brief scoping plans developed for resolving these issues

48 CFR 3001.105-1 - Publication and code arrangement.

Science.gov (United States)

2010-10-01

... 48 Federal Acquisition Regulations System 7 2010-10-01 2010-10-01 false Publication and code arrangement. 3001.105-1 Section 3001.105-1 Federal Acquisition Regulations System DEPARTMENT OF HOMELAND... Purpose, Authority, Issuance 3001.105-1 Publication and code arrangement. (a) The HSAR is published in: (1...

Study of no-man's land physics in the total-f gyrokinetic code XGC1

Science.gov (United States)

Ku, Seung Hoe; Chang, C. S.; Lang, J.

2014-10-01

While the ``transport shortfall'' in the ``no-man's land'' has been observed often in delta-f codes, it has not yet been observed in the global total-f gyrokinetic particle code XGC1. Since understanding the interaction between the edge and core transport appears to be a critical element in the prediction for ITER performance, understanding the no-man's land issue is an important physics research topic. Simulation results using the Holland case will be presented and the physics causing the shortfall phenomenon will be discussed. Nonlinear nonlocal interaction of turbulence, secondary flows, and transport appears to be the key.

International Observe the Moon Night: Providing Opportunities for the Public to Engage in Lunar Observation

Science.gov (United States)

Hsu, B. C.; Bleacher, L.; Day, B. H.; Daou, D.; Jones, A. P.; Mitchell, B.; Shaner, A. J.; Shipp, S. S.

2010-12-01

International Observe the Moon Night (InOMN) is designed to engage lunar science and education communities, our partner networks, amateur astronomers, space enthusiasts, and the general public in annual lunar observation campaigns that share the excitement of lunar science and exploration. InOMN enables the public to maintain its curiosity about the Moon and gain a better understanding of the Moon's formation, its evolution, and its place in the sky. For 2010, members of the public were encouraged to host their own InOMN events. InOMN hosts such as astronomy clubs, museums, schools, or other groups could find helpful resources and share information about InOMN events they organized on the InOMN website (http://observethemoonnight.org). Images, feedback, and lessons learned from the 2010 InOMN event will be shared in order to encourage increased planning and hosting of InOMN events in 2011. From various interpretations of the lunar “face,” early pictograms of the Moon’s phases, or to the use of the lunar cycle for festivals or harvests, the Moon has an undeniable influence on human civilization. We have chosen the 2011 InOMN theme to provide an opportunity for individuals to share their personal or cultural connections to the Moon. For 2011, the InOMN website will include a ‘lunar bulletin board’ where InOMN participants can post pictures and share stories of what the Moon means to them. The 2011 InOMN contest will encourage people to submit their works of art, poems, short stories, or music about the Moon all centered around the theme “What does the Moon mean to you?” As with the winners of previous contests, winning entries will be incorporated into the following year’s InOMN advertisements and events.

48 CFR 501.105-1 - Publication and code arrangement.

Science.gov (United States)

2010-10-01

... 48 Federal Acquisition Regulations System 4 2010-10-01 2010-10-01 false Publication and code arrangement. 501.105-1 Section 501.105-1 Federal Acquisition Regulations System GENERAL SERVICES... 501.105-1 Publication and code arrangement. The GSAR is published in the following sources: (a) Daily...

Assessment of TRAC-PF1/MOD1 code for large break LOCA in PWR

International Nuclear Information System (INIS)

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

1993-03-01

As the first step of the REFLA/TRAC code development, the TRAC/PF1/MOD1 code has been assessed for various experiments that simulate postulated large-break loss-of-coolant accident (LBLOCA) in PWR to understand the predictive capability and to identify the problem areas of the code. The assessment calculations were performed for separate effect tests for critical flow, counter current flow, condensation at cold leg and reflood as well as integral tests to understand predictability for individual phenomena. This report summarizes results from the assessment calculations of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The assessment calculations made clear the predictive capability and problem areas of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The areas, listed below, should be improved for more realistic and effective simulation of LBLOCA in PWR: (1) core heat transfer model during blowdown, (2) ECC bypass model at downcomer during refill, (3) condensation model during accumulator injection, and (4) core thermal hydraulic model during reflood. (author) 57 refs

FINAL SIMULATION RESULTS FOR DEMONSTRATION CASE 1 AND 2

Energy Technology Data Exchange (ETDEWEB)

David Sloan; Woodrow Fiveland

2003-10-15

The goal of this DOE Vision-21 project work scope was to develop an integrated suite of software tools that could be used to simulate and visualize advanced plant concepts. Existing process simulation software did not meet the DOE's objective of ''virtual simulation'' which was needed to evaluate complex cycles. The overall intent of the DOE was to improve predictive tools for cycle analysis, and to improve the component models that are used in turn to simulate equipment in the cycle. Advanced component models are available; however, a generic coupling capability that would link the advanced component models to the cycle simulation software remained to be developed. In the current project, the coupling of the cycle analysis and cycle component simulation software was based on an existing suite of programs. The challenge was to develop a general-purpose software and communications link between the cycle analysis software Aspen Plus{reg_sign} (marketed by Aspen Technology, Inc.), and specialized component modeling packages, as exemplified by industrial proprietary codes (utilized by ALSTOM Power Inc.) and the FLUENT{reg_sign} computational fluid dynamics (CFD) code (provided by Fluent Inc). A software interface and controller, based on an open CAPE-OPEN standard, has been developed and extensively tested. Various test runs and demonstration cases have been utilized to confirm the viability and reliability of the software. ALSTOM Power was tasked with the responsibility to select and run two demonstration cases to test the software--(1) a conventional steam cycle (designated as Demonstration Case 1), and (2) a combined cycle test case (designated as Demonstration Case 2). Demonstration Case 1 is a 30 MWe coal-fired power plant for municipal electricity generation, while Demonstration Case 2 is a 270 MWe, natural gas-fired, combined cycle power plant. Sufficient data was available from the operation of both power plants to complete the cycle

NeuCode Proteomics Reveals Bap1 Regulation of Metabolism

Directory of Open Access Journals (Sweden)

Joshua M. Baughman

2016-07-01

Full Text Available We introduce neutron-encoded (NeuCode amino acid labeling of mice as a strategy for multiplexed proteomic analysis in vivo. Using NeuCode, we characterize an inducible knockout mouse model of Bap1, a tumor suppressor and deubiquitinase whose in vivo roles outside of cancer are not well established. NeuCode proteomics revealed altered metabolic pathways following Bap1 deletion, including profound elevation of cholesterol biosynthetic machinery coincident with reduced expression of gluconeogenic and lipid homeostasis proteins in liver. Bap1 loss increased pancreatitis biomarkers and reduced expression of mitochondrial proteins. These alterations accompany a metabolic remodeling with hypoglycemia, hypercholesterolemia, hepatic lipid loss, and acinar cell degeneration. Liver-specific Bap1 null mice present with fully penetrant perinatal lethality, severe hypoglycemia, and hepatic lipid deficiency. This work reveals Bap1 as a metabolic regulator in liver and pancreas, and it establishes NeuCode as a reliable proteomic method for deciphering in vivo biology.

A restructuring of RN1 package for MIDAS computer code

International Nuclear Information System (INIS)

Park, S. H.; Kim, D. H.; Kim, K. R.

2003-01-01

RN1 package, which is one of two fission product-related packages in MELCOR, has been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and modernized data structure. To do this, data transferring methods of current MELCOR code are modified and adopted into the RN1 package. The data structure of the current MELCOR code using FORTRAN77 causes a difficult grasping of meaning of the variables as well as waste of memory. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type, which lead to an efficient memory treatment and an easy understanding of the code. Restructuring of the RN1 package addressed in this paper includes module development, subroutine modification, and treats MELGEN, which generates data file, as well as MELCOR, which is processing a calculation. The verification has been done by comparing the results of the modified code with those from the existing code. As the trends are similar to each other, it hints that the same approach could be extended to the entire code package. It is expected that code restructuring will accelerate the code domestication thanks to direct understanding of each variable and easy implementation of modified or newly developed models

Recent results in the decoding of Algebraic geometry codes

DEFF Research Database (Denmark)

Høholdt, Tom; Jensen, Helge Elbrønd; Nielsen, Rasmus Refslund

1998-01-01

We analyse the known decoding algorithms for algebraic geometry codes in the case where the number of errors is [(dFR-1)/2]+1, where dFR is the Feng-Rao distance......We analyse the known decoding algorithms for algebraic geometry codes in the case where the number of errors is [(dFR-1)/2]+1, where dFR is the Feng-Rao distance...

Verification of RESRAD-build computer code, version 3.1

International Nuclear Information System (INIS)

2003-01-01

for the review and any actions that were taken when these items were missing are documented in Section 5 of this report. The availability and use of user experience were limited to extensive experience in performing RESRAD-BUILD calculations by the verification project manager and by participation in the RESRAD-BUILD workshop offered by the code developers on May 11, 2001. The level of a posteriori verification that was implemented is defined in Sections 2 through 4 of this report. In general, a rigorous verification review plan addresses program requirements, design, coding, documentation, test coverage, and evaluation of test results. The scope of the RESRAD-BUILD verification is to focus primarily on program requirements, documentation, testing and evaluation. Detailed program design and source code review would be warranted only in those cases when the evaluation of test results and user experience revealed possible problems in these areas. The verification tasks were conducted in three parts and were applied to version 3.1 of the RESRAD-BUILD code and the final version of the user.s manual, issued in November 2001 (Yu (and others) 2001). These parts include the verification of the deterministic models used in RESRAD-BUILD (Section 2), the verification of the uncertainty analysis model included in RESRAD-BUILD (Section 3), and recommendations for improvement of the RESRAD-BUILD user interface, including evaluations of the user's manual, code design, and calculation methodology (Section 4). Any verification issues that were identified were promptly communicated to the RESRAD-BUILD development team, in particular those that arose from the database and parameter verification tasks. This allowed the developers to start implementing necessary database or coding changes well before this final report was issued

Case file coding of child maltreatment: Methods, challenges, and innovations in a longitudinal project of youth in foster care☆

Science.gov (United States)

Huffhines, Lindsay; Tunno, Angela M.; Cho, Bridget; Hambrick, Erin P.; Campos, Ilse; Lichty, Brittany; Jackson, Yo

2016-01-01

State social service agency case files are a common mechanism for obtaining information about a child’s maltreatment history, yet these documents are often challenging for researchers to access, and then to process in a manner consistent with the requirements of social science research designs. Specifically, accessing and navigating case files is an extensive undertaking, and a task that many researchers have had to maneuver with little guidance. Even after the files are in hand and the research questions and relevant variables have been clarified, case file information about a child’s maltreatment exposure can be idiosyncratic, vague, inconsistent, and incomplete, making coding such information into useful variables for statistical analyses difficult. The Modified Maltreatment Classification System (MMCS) is a popular tool used to guide the process, and though comprehensive, this coding system cannot cover all idiosyncrasies found in case files. It is not clear from the literature how researchers implement this system while accounting for issues outside of the purview of the MMCS or that arise during MMCS use. Finally, a large yet reliable file coding team is essential to the process, however, the literature lacks training guidelines and methods for establishing reliability between coders. In an effort to move the field toward a common approach, the purpose of the present discussion is to detail the process used by one large-scale study of child maltreatment, the Studying Pathways to Adjustment and Resilience in Kids (SPARK) project, a longitudinal study of resilience in youth in foster care. The article addresses each phase of case file coding, from accessing case files, to identifying how to measure constructs of interest, to dealing with exceptions to the coding system, to coding variables reliably, to training large teams of coders and monitoring for fidelity. Implications for a comprehensive and efficient approach to case file coding are discussed. PMID

[Differentiation of coding quality in orthopaedics by special, illustration-oriented case group analysis in the G-DRG System 2005].

Science.gov (United States)

Schütz, U; Reichel, H; Dreinhöfer, K

2007-01-01

We introduce a grouping system for clinical practice which allows the separation of DRG coding in specific orthopaedic groups based on anatomic regions, operative procedures, therapeutic interventions and morbidity equivalent diagnosis groups. With this, a differentiated aim-oriented analysis of illustrated internal DRG data becomes possible. The group-specific difference of the coding quality between the DRG groups following primary coding by the orthopaedic surgeon and final coding by the medical controlling is analysed. In a consecutive series of 1600 patients parallel documentation and group-specific comparison of the relevant DRG parameters were carried out in every case after primary and final coding. Analysing the group-specific share in the additional CaseMix coding, the group "spine surgery" dominated, closely followed by the groups "arthroplasty" and "surgery due to infection, tumours, diabetes". Altogether, additional cost-weight-relevant coding was necessary most frequently in the latter group (84%), followed by group "spine surgery" (65%). In DRGs representing conservative orthopaedic treatment documented procedures had nearly no influence on the cost weight. The introduced system of case group analysis in internal DRG documentation can lead to the detection of specific problems in primary coding and cost-weight relevant changes of the case mix. As an instrument for internal process control in the orthopaedic field, it can serve as a communicative interface between an economically oriented classification of the hospital performance and a specific problem solution of the medical staff involved in the department management.

Introduction and immigration of TRAC-PF1 code

International Nuclear Information System (INIS)

Yan Yuhua; Gao Zuying; Gao Cheng; Li Jincai

1997-01-01

TRAC-PF1 code performs best-estimate predictions of postulated accidents for pressurized light water reactors. It is one of the few system analyse codes which use two fluid model to treat two phase problems in nuclear system. In order to use this advanced software in China and make it possible to be run in different compute systems, IBM version of TRAC-PF1 code, imported from USA National Energy Software Center, is immigrated to CDC NOS/VE system and SUN workstation. The differences in computer languages from IBM 370 to CDC NOS/VE and to SUN workstation are modified properly. All the benchmark problems are calculated, and the results show that the immigration is successful

LDGM Codes for Channel Coding and Joint Source-Channel Coding of Correlated Sources

Directory of Open Access Journals (Sweden)

Javier Garcia-Frias

2005-05-01

Full Text Available We propose a coding scheme based on the use of systematic linear codes with low-density generator matrix (LDGM codes for channel coding and joint source-channel coding of multiterminal correlated binary sources. In both cases, the structures of the LDGM encoder and decoder are shown, and a concatenated scheme aimed at reducing the error floor is proposed. Several decoding possibilities are investigated, compared, and evaluated. For different types of noisy channels and correlation models, the resulting performance is very close to the theoretical limits.

Development of 1D Liner Compression Code for IDL

Science.gov (United States)

Shimazu, Akihisa; Slough, John; Pancotti, Anthony

2015-11-01

A 1D liner compression code is developed to model liner implosion dynamics in the Inductively Driven Liner Experiment (IDL) where FRC plasmoid is compressed via inductively-driven metal liners. The driver circuit, magnetic field, joule heating, and liner dynamics calculations are performed at each time step in sequence to couple these effects in the code. To obtain more realistic magnetic field results for a given drive coil geometry, 2D and 3D effects are incorporated into the 1D field calculation through use of correction factor table lookup approach. Commercial low-frequency electromagnetic fields solver, ANSYS Maxwell 3D, is used to solve the magnetic field profile for static liner condition at various liner radius in order to derive correction factors for the 1D field calculation in the code. The liner dynamics results from the code is verified to be in good agreement with the results from commercial explicit dynamics solver, ANSYS Explicit Dynamics, and previous liner experiment. The developed code is used to optimize the capacitor bank and driver coil design for better energy transfer and coupling. FRC gain calculations are also performed using the liner compression data from the code for the conceptual design of the reactor sized system for fusion energy gains.

Variation in National ACGME Case Log Data for Pediatric Orthopaedic Fellowships: Are Fellow Coding Practices Responsible?

Science.gov (United States)

McClure, Philip K; Woiczik, Marcella; Karol, Lori; Sankar, Wudbhav N

The introduction of the 80-hour work week for Accreditation Council for Graduate Medical Education (ACGME) accredited fellowship programs initiated many efforts to optimize surgical training. One particular area of interest is on recording and tracking surgical experiences. The current standard is logging cases based on Current Procedural Terminology codes, which are primarily designed for billing. Proposed guidelines from the ACGME regarding logging exist, but their implementation is unknown, as is the variation in case volume across fellowship programs. The purpose of this study was to investigate variability in the national case log data, and explore potential sources of variation using fellow surveys. National ACGME case log data for pediatric orthopaedic fellowships from 2012 to 2015 were reviewed, with particular attention to the domains of spine, pelvis/hip, arthroscopy, trauma, and other (which includes clubfoot casting). To explore potential sources of case log variability, a survey on case logging behavior was distributed to all pediatric orthopaedic fellows for the academic year 2015 to 2016. Reported experiences based on ACGME case logs varied widely between fellows with percentage difference of up to 100% in all areas. Similarly, wide variability is present in coding practices of pediatric orthopaedic fellows, who often lack formal education on the topic of appropriate coding/logging. In the survey, hypothetical case scenarios had an absolute difference in recorded codes of up to 13 and a percentage difference of up to 100%. ACGME case log data for pediatric orthopaedic fellowships demonstrates wide variability in reported surgical experiences. This variability may be due, in part, to differences in logging practices by individual fellows. This observation makes meaningful interpretation of national data on surgical volume challenging. Proposed surgical experience minimums should be interpreted in light of these data, and may not be advisable unless

The impact of three discharge coding methods on the accuracy of diagnostic coding and hospital reimbursement for inpatient medical care.

Science.gov (United States)

Tsopra, Rosy; Peckham, Daniel; Beirne, Paul; Rodger, Kirsty; Callister, Matthew; White, Helen; Jais, Jean-Philippe; Ghosh, Dipansu; Whitaker, Paul; Clifton, Ian J; Wyatt, Jeremy C

2018-07-01

Coding of diagnoses is important for patient care, hospital management and research. However coding accuracy is often poor and may reflect methods of coding. This study investigates the impact of three alternative coding methods on the inaccuracy of diagnosis codes and hospital reimbursement. Comparisons of coding inaccuracy were made between a list of coded diagnoses obtained by a coder using (i)the discharge summary alone, (ii)case notes and discharge summary, and (iii)discharge summary with the addition of medical input. For each method, inaccuracy was determined for the primary, secondary diagnoses, Healthcare Resource Group (HRG) and estimated hospital reimbursement. These data were then compared with a gold standard derived by a consultant and coder. 107 consecutive patient discharges were analysed. Inaccuracy of diagnosis codes was highest when a coder used the discharge summary alone, and decreased significantly when the coder used the case notes (70% vs 58% respectively, p coded from the discharge summary with medical support (70% vs 60% respectively, p coding with case notes, and 35% for coding with medical support. The three coding methods resulted in an annual estimated loss of hospital remuneration of between £1.8 M and £16.5 M. The accuracy of diagnosis codes and percentage of correct HRGs improved when coders used either case notes or medical support in addition to the discharge summary. Further emphasis needs to be placed on improving the standard of information recorded in discharge summaries. Copyright © 2018 Elsevier B.V. All rights reserved.

Sandia reactor kinetics codes: SAK and PK1D

International Nuclear Information System (INIS)

Pickard, P.S.; Odom, J.P.

1978-01-01

The Sandia Kinetics code (SAK) is a one-dimensional coupled thermal-neutronics transient analysis code for use in simulation of reactor transients. The time-dependent cross section routines allow arbitrary time-dependent changes in material properties. The one-dimensional heat transfer routines are for cylindrical geometry and allow arbitrary mesh structure, temperature-dependent thermal properties, radiation treatment, and coolant flow and heat-transfer properties at the surface of a fuel element. The Point Kinetics 1 Dimensional Heat Transfer Code (PK1D) solves the point kinetics equations and has essentially the same heat-transfer treatment as SAK. PK1D can address extended reactor transients with minimal computer execution time

Systematic review of validated case definitions for diabetes in ICD-9-coded and ICD-10-coded data in adult populations.

Science.gov (United States)

Khokhar, Bushra; Jette, Nathalie; Metcalfe, Amy; Cunningham, Ceara Tess; Quan, Hude; Kaplan, Gilaad G; Butalia, Sonia; Rabi, Doreen

2016-08-05

With steady increases in 'big data' and data analytics over the past two decades, administrative health databases have become more accessible and are now used regularly for diabetes surveillance. The objective of this study is to systematically review validated International Classification of Diseases (ICD)-based case definitions for diabetes in the adult population. Electronic databases, MEDLINE and Embase, were searched for validation studies where an administrative case definition (using ICD codes) for diabetes in adults was validated against a reference and statistical measures of the performance reported. The search yielded 2895 abstracts, and of the 193 potentially relevant studies, 16 met criteria. Diabetes definition for adults varied by data source, including physician claims (sensitivity ranged from 26.9% to 97%, specificity ranged from 94.3% to 99.4%, positive predictive value (PPV) ranged from 71.4% to 96.2%, negative predictive value (NPV) ranged from 95% to 99.6% and κ ranged from 0.8 to 0.9), hospital discharge data (sensitivity ranged from 59.1% to 92.6%, specificity ranged from 95.5% to 99%, PPV ranged from 62.5% to 96%, NPV ranged from 90.8% to 99% and κ ranged from 0.6 to 0.9) and a combination of both (sensitivity ranged from 57% to 95.6%, specificity ranged from 88% to 98.5%, PPV ranged from 54% to 80%, NPV ranged from 98% to 99.6% and κ ranged from 0.7 to 0.8). Overall, administrative health databases are useful for undertaking diabetes surveillance, but an awareness of the variation in performance being affected by case definition is essential. The performance characteristics of these case definitions depend on the variations in the definition of primary diagnosis in ICD-coded discharge data and/or the methodology adopted by the healthcare facility to extract information from patient records. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://www.bmj.com/company/products-services/rights-and-licensing/

CRA Control Logic Realization for MARS 1-D/MASTER coupled Code System

International Nuclear Information System (INIS)

Han, Soonkyoo; Jeong, Sungsu; Lee, Suyong

2013-01-01

Both Multi-dimensional Analysis Reactor Safety (MARS) code and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) code, developed by Korea Atomic Energy Research Institute (KAERI), can be coupled for various simulations of nuclear reactor system. In the MARS 1-D/MASTER coupled code system, MARS is used for the thermal hydraulic calculations and MASTER is used for reactor core calculations. In case of using this coupled code system, the movements of control rod assembly (CRA) are controlled by MASTER. MASTER, however, has a CRA control function which is inputted by user as a form of time dependent table. When simulations related to sequential CRA insertion or withdrawal which are not ejection or drop are performed, this CRA control function is not sufficient to demonstrate the process of CRA movements. Therefore an alternative way is proposed for realization of CRA control logic in MASTER. In this study, the manually realized CRA control logic was applied by inputting the time dependent CRA positions into MASTER. And the points of CRA movements were decided by iterations. At the end of CRA movement, the reactor power difference and the average coolant temperature difference were not out of the range of their dead bands. Therefore it means that this manually realized CRA control logic works appropriately in the dead bands of the logic. Therefore the proper CRA movement points could be decided by using this manually realized CRA control logic. Based on these results, it is verified that the proper CRA movement points can be chosen by using the proposed CRA control logic in this article. In conclusion, it is expected that this proposed CRA control logic in MASTER can be used to properly demonstrate the process related to CRA sequential movements in the MARS 1-D/MASTER coupled code system

17 CFR 275.204A-1 - Investment adviser codes of ethics.

Science.gov (United States)

2010-04-01

... ethics. 275.204A-1 Section 275.204A-1 Commodity and Securities Exchanges SECURITIES AND EXCHANGE... codes of ethics. (a) Adoption of code of ethics. If you are an investment adviser registered or required... enforce a written code of ethics that, at a minimum, includes: (1) A standard (or standards) of business...

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

Science.gov (United States)

2010-01-01

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

The development of the code package PERMAK--3D//SC--1

International Nuclear Information System (INIS)

Bolobov, P. A.; Oleksuk, D. A.

2011-01-01

Code package PERMAK-3D//SC-1 was developed for performing pin-by-pin coupled neutronic and thermal hydraulic calculation of the core fragment of seven fuel assemblies and was designed on the basis of 3D multigroup pin-by-pin code PERMAK-3D and 3D (subchannel) thermal hydraulic code SC-1 The code package predicts axial and radial pin-by-pin power distribution and coolant parameters in stimulated region (enthalpies,, velocities,, void fractions,, boiling and DNBR margins).. The report describes some new steps in code package development. Some PERMAK-3D//SC-1 outcomes of WWER calculations are presented in the report. (Authors)

Code on nuclear air and gas treatment ASME/ANSI AG-1

International Nuclear Information System (INIS)

Miller, W.H. Jr.

1993-01-01

The focus of this panel is on equipment code section work over the past two years. Major topics include changes in Filter Code Sections, revamping of the Ductwork Code Section, and emergence of an improved I ampersand C Code Section. Actual applications of AG-1 are to be discussed by CONAGT members. Remaining time will be devoted to fielding questions concerning ASMA/ANSI AG-1

The LIONS code (version 1.0)

International Nuclear Information System (INIS)

Bertrand, P.

1993-01-01

The new LIONS code (Lancement d'IONS or Ion Launching), a dynamical code implemented in the SPIRaL project for the CIME cyclotron studies, is presented. The various software involves a 3D magnetostatic code, 2D or 3D electrostatic codes for generation of realistic field maps, and several dynamical codes for studying the behaviour of the reference particle from the cyclotron center up to the ejection and for launching particles packets complying with given correlations. Its interactions with the other codes are described. The LIONS code, written in Fortran 90 is already used in studying the CIME cyclotron, from the center to the ejection. It is designed to be used, with minor modifications, in other contexts such as for the simulation of mass spectrometer facilities

"Hour of Code": A Case Study

Science.gov (United States)

Du, Jie; Wimmer, Hayden; Rada, Roy

2018-01-01

This study investigates the delivery of the "Hour of Code" tutorials to college students. The college students who participated in this study were surveyed about their opinion of the Hour of Code. First, the students' comments were discussed. Next, a content analysis of the offered tutorials highlights their reliance on visual…

International Observe the Moon Night: Eight Years of Engaging Scientists, Educators, and Citizen Enthusiasts in NASA Science

Science.gov (United States)

Buxner, Sanlyn; Jones, Andrea; Bleacher, Lora; Wasser, Molly; Day, Brian; Bakerman, Maya; Shaner, Andrew; Joseph, Emily; International Observe the Moon Night Coordinating Committee

2018-01-01

International Observe the Moon Night (InOMN) is an annual worldwide event, held in the fall, that celebrates lunar and planetary science and exploration. InOMN is sponsored by NASA’s Lunar Reconnaissance Orbiter (LRO) in collaboration with NASA’s Solar System Exploration Research Virtual Institute (SSERVI), the NASA’s Heliophysics Education Consortium, CosmoQuest, Night Sky Network, and Science Festival Alliance. Other key partners include the NASA Museum Alliance, Night Sky Network, and NASA Solar System Ambassadors.In 2017 InOMN will be held on October 28th, and will engage thousands of people across the globe to observe and learn about the Moon and its connection to planetary science. This year, we have partnered with the NASA Science Mission Directorate total solar eclipse team to highlight InOMN as an opportunity to harness and sustain the interest and momentum in space science and observation following the August 21st eclipse. Since 2010, over 3,800 InOMN events have been registered engaging over 550,000 visitors worldwide. Most InOMN events are held in the United States, with strong representation from many other countries. We will present current results from the 2017 InOMN evaluation.Through InOMN, we annually provide resources such as event-specific Moon maps, presentations, advertising materials, and certificates of participation. Additionally, InOMN highlights partner resources such as online interfaces including Moon Trek (https://moontrek.jpl.nasa.gov) and CosmoQuest (https://cosmoquest.org/x/) to provide further opportunities to engage with NASA science.Learn more about InOMN at http://observethemoonnight.org.

Acceleration of a Monte Carlo radiation transport code

International Nuclear Information System (INIS)

Hochstedler, R.D.; Smith, L.M.

1996-01-01

Execution time for the Integrated TIGER Series (ITS) Monte Carlo radiation transport code has been reduced by careful re-coding of computationally intensive subroutines. Three test cases for the TIGER (1-D slab geometry), CYLTRAN (2-D cylindrical geometry), and ACCEPT (3-D arbitrary geometry) codes were identified and used to benchmark and profile program execution. Based upon these results, sixteen top time-consuming subroutines were examined and nine of them modified to accelerate computations with equivalent numerical output to the original. The results obtained via this study indicate that speedup factors of 1.90 for the TIGER code, 1.67 for the CYLTRAN code, and 1.11 for the ACCEPT code are achievable. copyright 1996 American Institute of Physics

Updating of ASME Nuclear Code Case N-201 to Accommodate the Needs of Metallic Core Support Structures for High Temperature Gas Cooled Reactors Currently in Development

International Nuclear Information System (INIS)

Basol, Mit; Kielb, John F.; MuHooly, John F.; Smit, Kobus

2007-01-01

On September 29, 2005, ASME Standards Technology, LLC (ASME ST-LLC) executed a multi-year, cooperative agreement with the United States DOE for the Generation IV Reactor Materials project. The project's objective is to update and expand appropriate materials, construction, and design codes for application in future Generation IV nuclear reactor systems that operate at elevated temperatures. Task 4 was embarked upon in recognition of the large quantity of ongoing reactor designs utilizing high temperature technology. Since Code Case N-201 had not seen a significant revision (except for a minor revision in September, 2006 to change the SA-336 forging reference for 304SS and 316SS to SA-965 in Tables 1.2(a) and 1.2(b), and some minor editorial changes) since December 1994, identifying recommended updates to support the current high temperature Core Support Structure (CSS) designs and potential new designs was important. As anticipated, the Task 4 effort identified a number of Code Case N-201 issues. Items requiring further consideration range from addressing apparent inconsistencies in definitions and certain material properties between CC-N-201 and Subsection NH, to inclusion of additional materials to provide the designer more flexibility of design. Task 4 developed a design parameter survey that requested input from the CSS designers of ongoing high temperature gas cooled reactor metallic core support designs. The responses to the survey provided Task 4 valuable input to identify the design operating parameters and future needs of the CSS designers. Types of materials, metal temperature, time of exposure, design pressure, design life, and fluence levels were included in the Task 4 survey responses. The results of the survey are included in this report. This research proves that additional work must be done to update Code Case N-201. Task 4 activities provide the framework for the Code Case N-201 update and future work to provide input on materials. Candidate

The oxidation state of manganese in ferromanganese nodules and deep-sea sediments from the Central Indian Ocean

Digital Repository Service at National Institute of Oceanography (India)

Pattan, J.N.; Mudholkar, A.V.

nodule field (lO°S). Sotid Phase Mn (wf.%) O/Mn ratio Solid Phase Mn (wt?/o) O/Mn ratio S i6 1.7 1.8 1.9 2.,0 I~ 1.,S I~6 Ii7 I.,8 I.,9 I I I ._m..m I~ I = I 3o ~\\[ 36 I SK/20/226 5S/10/657 Fig. 4. Distribution of solid-phase Mn (wt.%) and O...

Code portability and data management considerations in the SAS3D LMFBR accident-analysis code

International Nuclear Information System (INIS)

Dunn, F.E.

1981-01-01

The SAS3D code was produced from a predecessor in order to reduce or eliminate interrelated problems in the areas of code portability, the large size of the code, inflexibility in the use of memory and the size of cases that can be run, code maintenance, and running speed. Many conventional solutions, such as variable dimensioning, disk storage, virtual memory, and existing code-maintenance utilities were not feasible or did not help in this case. A new data management scheme was developed, coding standards and procedures were adopted, special machine-dependent routines were written, and a portable source code processing code was written. The resulting code is quite portable, quite flexible in the use of memory and the size of cases that can be run, much easier to maintain, and faster running. SAS3D is still a large, long running code that only runs well if sufficient main memory is available

Transport calculations with the BALDUR code. Pt. 1

International Nuclear Information System (INIS)

Lackner, K.; Wunderlich, R.

1979-12-01

1-d transport calculations with the BALDUR-code are described for predicting the performance of ZEPHYR under D-T operation. Results presented in this report refer to the impurity-free case, and ion and electron heat conduction losses described by CHIsub(i) = neoclassical and CHIsub(e) = 6.25 x 10 17 /nsub(e) (cgs-units). A simple refuelling scenario taking account of the density limit for the ohmic heating phase, the contribution of neutral injection to the refuelling rate and the need for an approximately balanced D-T mixture at the instance of ignition is adopted. The heating scenario assumes a neutral injection beam with 160 keV particle energy in the main component, with a duration of 1.1 sec. Major radius compression by a factor of 1.5 starts 1 sec after the onset of neutral injection and lasts 100 msec. For this standard scenario the performance is studied in different density regimes and for different neutral injection powers. Under the above assumption ignition is predicted for total neutral injection powers < approx. 16 MW (9.6 MW in the main energy component) and average total β-values < 2.8%. Results including impurities, alternative scaling laws, and deviations from the standard scenario will be presented in another report. (orig.) 891 GG/orig. 892 HIS

Microdosimetry computation code of internal sources - MICRODOSE 1

International Nuclear Information System (INIS)

Li Weibo; Zheng Wenzhong; Ye Changqing

1995-01-01

This paper describes a microdosimetry computation code, MICRODOSE 1, on the basis of the following described methods: (1) the method of calculating f 1 (z) for charged particle in the unit density tissues; (2) the method of calculating f(z) for a point source; (3) the method of applying the Fourier transform theory to the calculation of the compound Poisson process; (4) the method of using fast Fourier transform technique to determine f(z) and, giving some computed examples based on the code, MICRODOSE 1, including alpha particles emitted from 239 Pu in the alveolar lung tissues and from radon progeny RaA and RAC in the human respiratory tract. (author). 13 refs., 6 figs

MACCS version 1.5.11.1: A maintenance release of the code

International Nuclear Information System (INIS)

Chanin, D.; Foster, J.; Rollstin, J.; Miller, L.

1993-10-01

A new version of the MACCS code (version 1.5.11.1) has been developed by Sandia National Laboratories under sponsorship of the US Nuclear Regulatory Commission. MACCS was developed to support evaluations of the off-site consequences from hypothetical severe accidents at commercial power plants. MACCS is the only current public domain code in the US that embodies all of the following modeling capabilities: (1) weather sampling using a year of recorded weather data; (2) mitigative actions such as evacuation, sheltering, relocation, decontamination, and interdiction; (3) economic costs of mitigative actions; (4) cloudshine, groundshine, and inhalation pathways as well as food and water ingestion; (5) calculation of both individual and societal doses to various organs; and (6) calculation of both acute (nonstochastic) and latent (stochastic) health effects and risks of health effects. All of the consequence measures may be fun generated in the form of a complementary cumulative distribution function (CCDF). The current version implements a revised cancer model consistent with recent reports such as BEIR V and ICRP 60. In addition, a number of error corrections and portability enhancements have been implemented. This report describes only the changes made in creating the new version. Users of the code will need to obtain the code's original documentation, NUREG/CR-4691

Some Families of Asymmetric Quantum MDS Codes Constructed from Constacyclic Codes

Science.gov (United States)

Huang, Yuanyuan; Chen, Jianzhang; Feng, Chunhui; Chen, Riqing

2018-02-01

Quantum maximal-distance-separable (MDS) codes that satisfy quantum Singleton bound with different lengths have been constructed by some researchers. In this paper, seven families of asymmetric quantum MDS codes are constructed by using constacyclic codes. We weaken the case of Hermitian-dual containing codes that can be applied to construct asymmetric quantum MDS codes with parameters [[n,k,dz/dx

Rateless feedback codes

DEFF Research Database (Denmark)

Sørensen, Jesper Hemming; Koike-Akino, Toshiaki; Orlik, Philip

2012-01-01

This paper proposes a concept called rateless feedback coding. We redesign the existing LT and Raptor codes, by introducing new degree distributions for the case when a few feedback opportunities are available. We show that incorporating feedback to LT codes can significantly decrease both...... the coding overhead and the encoding/decoding complexity. Moreover, we show that, at the price of a slight increase in the coding overhead, linear complexity is achieved with Raptor feedback coding....

Vectorization of nuclear codes 90-1

International Nuclear Information System (INIS)

Nonomiya, Iwao; Nemoto, Toshiyuki; Ishiguro, Misako; Harada, Hiroo; Hori, Takeo.

1990-09-01

The vectorization has been made for four codes: SONATINA-2V HTTR version, TRIDOSE, VIENUS, and SCRYU. SONATINA-2V HTTR version is a code for analyzing the dynamic behavior of fuel blocks in the vertical slice of the HTGR (High Temperature Gas-cooled Reactor) core under seismic perturbation, TRIDOSE is a code for calculating environmental tritium concentration and dose, VIENUS is a code for analyzing visco elastic stress of the fuel block of HTTR (High Temperature gas-cooled Test Reactor), and SCRYU is a thermal-hydraulics code with boundary fitted coordinate system. The total speedup ratio of the vectorized versions to the original scalar ones is 5.2 for SONATINA-2V HTTR version. 5.9 ∼ 6.9 for TRIDOSE, 6.7 for VIENUS, 7.6 for SCRYU, respectively. In this report, we describe outline of codes, techniques used for the vectorization, verification of computed results, and speedup effect on the vectorized codes. (author)

Performance Analysis of a Decoding Algorithm for Algebraic Geometry Codes

DEFF Research Database (Denmark)

Jensen, Helge Elbrønd; Nielsen, Rasmus Refslund; Høholdt, Tom

1998-01-01

We analyse the known decoding algorithms for algebraic geometry codes in the case where the number of errors is greater than or equal to [(dFR-1)/2]+1, where dFR is the Feng-Rao distance......We analyse the known decoding algorithms for algebraic geometry codes in the case where the number of errors is greater than or equal to [(dFR-1)/2]+1, where dFR is the Feng-Rao distance...

International benchmark study of advanced thermal hydraulic safety analysis codes against measurements on IEA-R1 research reactor

Energy Technology Data Exchange (ETDEWEB)

Hainoun, A., E-mail: pscientific2@aec.org.sy [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Doval, A. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400 S.C de Bariloche, Rio Negro (Argentina); Umbehaun, P. [Centro de Engenharia Nuclear – CEN, IPEN-CNEN/SP, Av. Lineu Prestes 2242-Cidade Universitaria, CEP-05508-000 São Paulo, SP (Brazil); Chatzidakis, S. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Ghazi, N. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Park, S. [Research Reactor Design and Engineering Division, Basic Science Project Operation Dept., Korea Atomic Energy Research Institute (Korea, Republic of); Mladin, M. [Institute for Nuclear Research, Campului Street No. 1, P.O. Box 78, 115400 Mioveni, Arges (Romania); Shokr, A. [Division of Nuclear Installation Safety, Research Reactor Safety Section, International Atomic Energy Agency, A-1400 Vienna (Austria)

2014-12-15

Highlights: • A set of advanced system thermal hydraulic codes are benchmarked against IFA of IEA-R1. • Comparative safety analysis of IEA-R1 reactor during LOFA by 7 working teams. • This work covers both experimental and calculation effort and presents new out findings on TH of RR that have not been reported before. • LOFA results discrepancies from 7% to 20% for coolant and peak clad temperatures are predicted conservatively. - Abstract: In the framework of the IAEA Coordination Research Project on “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal hydraulic computational methods and tools for operation and safety analysis of research reactors” the Brazilian research reactor IEA-R1 has been selected as reference facility to perform benchmark calculations for a set of thermal hydraulic codes being widely used by international teams in the field of research reactor (RR) deterministic safety analysis. The goal of the conducted benchmark is to demonstrate the application of innovative reactor analysis tools in the research reactor community, validation of the applied codes and application of the validated codes to perform comprehensive safety analysis of RR. The IEA-R1 is equipped with an Instrumented Fuel Assembly (IFA) which provided measurements for normal operation and loss of flow transient. The measurements comprised coolant and cladding temperatures, reactor power and flow rate. Temperatures are measured at three different radial and axial positions of IFA summing up to 12 measuring points in addition to the coolant inlet and outlet temperatures. The considered benchmark deals with the loss of reactor flow and the subsequent flow reversal from downward forced to upward natural circulation and presents therefore relevant phenomena for the RR safety analysis. The benchmark calculations were performed independently by the participating teams using different thermal hydraulic and safety

International Observe the Moon Night: Using Public Outreach Events to Tell Your Story to the Public

Science.gov (United States)

Hsu, B. C.; International Observe the Moon Night Coordinating Committee

2011-12-01

From various interpretations of the lunar "face," early pictograms of the Moon's phases, or to the use of the lunar cycle for festivals or harvests, the Moon has an undeniable influence on human civilization. International Observe the Moon Night (InOMN) capitalizes on the human connection to the Moon by engaging the public in annual lunar observation campaigns that share the excitement of lunar science and exploration. In 2010 (InOMN's inaugural year), over 500,000 people attended events in 53 countries around the world. About 68% of InOMN hosts - astronomy clubs, museums, schools, or other groups - used the resources on the InOMN website (http://observethemoonnight.org). The InOMN website provided supporting materials for InOMN event hosts in the form of downloadable advertising materials, Moon maps, suggestions for hands-on educational activities, and links to lunar science content. InOMN event participants shared their experiences with the world using the Web and social media, event hosts shared their experiences with evaluation data, and amateur astronomers and photographers shared their images of the Moon through the lunar photography contest. The overwhelming response from InOMN in 2010 represents an untapped potential for infusing cutting edge lunar science and exploration into a large-scale public outreach event.

THYDE-B1/MOD1: a computer code for analysis of small-break loss-of-coolant accident of boiling water reactors

International Nuclear Information System (INIS)

Muramatsu, Ken; Akimoto, Masayuki

1982-08-01

THYDE-B1/MOD1 is a computer code to analyze thermo-hydraulic transients of the reactor cooling system of a BWR, mainly during a small-break loss-of-coolant accidnet (SB-LOCA) with a special emphasis on the behavior of pressure and mixture level in the pressure vessel. The coolant behavior is simulated with a volume-and-junction method based on assumptions of thermal equilibrium and homogeneous conditions for two-phase flow. A characteristic feature of this code is a three-region representation of the state of the coolant in a control volume, in which three regions, i.e., subcooled liquid, saturated mixture and saturated steam regions are allowed to exist. The regions are separated by moving boundaries, tracked by mass and energy balances for each region. The interior of the pressure vessel is represented by two volumes with three regions: one for inside of the shroud and the other for outside, while other portions of the system are treated with homogeneous model. This method, although it seems to be very simple, has been verified to be adequate for cases of BWR SB-LOCAs in which the hydraulic transient is relatively slow and the cooling of the core strongly depends on the mixture level behavior in the vessel. In order to simulate the system behavior, THYDE-B1 is provided with analytical models for reactor kinetics, heat generation and conduction in fuel rods and structures, heat transfer between coolant and solid surfaces, coolant injection systems, breaks and discharge systems, jet pumps, recirculation pumps, and so on. The verification of the code has been conducted. A good predictability of the code has been indicated through the comparison of calculated results with experimental data provided by ROSA-III small-break tests. This report presents the analytical models, solution method, and input data requirements of the THYDE-B1/MOD1 code. (author)

The Genomic Code: Genome Evolution and Potential Applications

KAUST Repository

Bernardi, Giorgio

2016-01-25

The genome of metazoans is organized according to a genomic code which comprises three laws: 1) Compositional correlations hold between contiguous coding and non-coding sequences, as well as among the three codon positions of protein-coding genes; these correlations are the consequence of the fact that the genomes under consideration consist of fairly homogeneous, long (≥200Kb) sequences, the isochores; 2) Although isochores are defined on the basis of purely compositional properties, GC levels of isochores are correlated with all tested structural and functional properties of the genome; 3) GC levels of isochores are correlated with chromosome architecture from interphase to metaphase; in the case of interphase the correlation concerns isochores and the three-dimensional “topological associated domains” (TADs); in the case of mitotic chromosomes, the correlation concerns isochores and chromosomal bands. Finally, the genomic code is the fourth and last pillar of molecular biology, the first three pillars being 1) the double helix structure of DNA; 2) the regulation of gene expression in prokaryotes; and 3) the genetic code.

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

International Nuclear Information System (INIS)

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

2009-05-01

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

Procedure and code for calculating black control rods taking into account epithermal absorption, code CAS-1

International Nuclear Information System (INIS)

Martinc, R.; Trivunac, N.; Zivkovic, Z.

1964-12-01

This report describes the computer code CAS-1, calculation method and procedure applied for calculating the black control rods taking into account the epithermal neutron absorption. Results obtained for supercell method applied for regular lattice reflected in the multiplication medium is part of this report in addition to the computer code manual

Converter of a continuous code into the Grey code

International Nuclear Information System (INIS)

Gonchar, A.I.; TrUbnikov, V.R.

1979-01-01

Described is a converter of a continuous code into the Grey code used in a 12-charged precision amplitude-to-digital converter to decrease the digital component of spectrometer differential nonlinearity to +0.7% in the 98% range of the measured band. To construct the converter of a continuous code corresponding to the input signal amplitude into the Grey code used is the regularity in recycling of units and zeroes in each discharge of the Grey code in the case of a continuous change of the number of pulses of a continuous code. The converter is constructed on the elements of 155 series, the frequency of continuous code pulse passing at the converter input is 25 MHz

TRAC-PF1/MOD1 computer code

International Nuclear Information System (INIS)

Liles, D.R.; Mahaffy, J.H.

1984-01-01

The TRAC-P1 program was designed primarily for the analysis of large-break loss-of-coolant accidents (LOCAs) in pressurized water reactors (PWRs). Because of its versatility, however, it can be applied directly to many analyses ranging from blowdowns in simple pipes to integral LOCA tests in multiloop facilities. A refined version, called TRAC-P1A, was released to the National Energy Software Center (NESC) in March 1979. Although it still treats the same class of problems, TRAC-P1A is more efficient than TRAC-P1 and incorporates improved hydrodynamic and heat-transfer models. It also is easier to implement on various computers. TRAC-PD2 contains improved reflood and heat-transfer models and improvements in the numerical solution methods. Although a large LOCA code, it has been applied successfully to small-break problems and to the Three Mile Island incident. Distinguishing characteristics of the TRAC-PF1/MOD1 are summarized

International Observe the Moon Night: A Worldwide Public Observing Event that Annually Engages Scientists, Educators, and Citizen Enthusiasts in NASA Science

Science.gov (United States)

Buxner, S.; Jones, A. P.; Bleacher, L.; Wasser, M. L.; Day, B. H.; Shaner, A. J.; Bakerman, M. N.; Joseph, E.

2017-12-01

International Observe the Moon Night (InOMN) is an annual worldwide event, held in the fall, that celebrates lunar and planetary science and exploration. InOMN is sponsored by NASA's Lunar Reconnaissance Orbiter (LRO) in collaboration with NASA's Solar System Exploration Research Virtual Institute (SSERVI), the NASA's Heliophysics Education Consortium, CosmoQuest, Night Sky Network, and Science Festival Alliance. Other key partners include the NASA Museum Alliance, Night Sky Network, and NASA Solar System Ambassadors. In 2017, InOMN will bring together thousands of people across the globe to observe and learn about the Moon and its connection to planetary science. We are partnering with the NASA Science Mission Directorate total solar eclipse team to highlight InOMN as an opportunity to harness and sustain the interest and momentum in space science and observation following the August 21st eclipse. This is part of a new partnership with the Sun-Earth Day team, through the Heliophysics Education Consortium, to better connect the two largest NASA-sponsored public engagement events, increase participation in both events, and share best practices in implementation and evaluation between the teams. Over 3,800 InOMN events have been registered between 2010 and 2016, engaging over 550,000 visitors worldwide. Most InOMN events are held in the United States, with strong representation from many other countries. InOMN events are evaluated to determine the value of the events and to allow us to improve the experience for event hosts and visitors. Our results show that InOMN events are hosted by scientists, educators, and citizen enthusiasts around the world who leverage InOMN to bring communities together, get visitors excited and learn about the Moon - and beyond, and share resources to extend engagement in lunar and planetary science and observation. Through InOMN, we annually provide resources such as event-specific Moon maps, presentations, advertising materials, and

Guidelines for selecting codes for ground-water transport modeling of low-level waste burial sites. Volume 2. Special test cases

International Nuclear Information System (INIS)

Simmons, C.S.; Cole, C.R.

1985-08-01

This document was written for the National Low-Level Waste Management Program to provide guidance for managers and site operators who need to select ground-water transport codes for assessing shallow-land burial site performance. The guidance given in this report also serves the needs of applications-oriented users who work under the direction of a manager or site operator. The guidelines are published in two volumes designed to support the needs of users having different technical backgrounds. An executive summary, published separately, gives managers and site operators an overview of the main guideline report. Volume 1, titled ''Guideline Approach,'' consists of Chapters 1 through 5 and a glossary. Chapters 2 through 5 provide the more detailed discussions about the code selection approach. This volume, Volume 2, consists of four appendices reporting on the technical evaluation test cases designed to help verify the accuracy of ground-water transport codes. 20 refs

Developments on ASME Code Cases to Risk-Informed Repair/Replacement Activities in Support of Risk-Informed Regulation Initiatives

International Nuclear Information System (INIS)

Balkey, Kenneth R.; Holston, William C.

2002-01-01

ASME Code Case N-658, 'Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities' and Code Case N-660, 'Alternative Repair/Replacement Requirements For Items Classified In Accordance With Risk-Informed Processes' are being completed to expand the breadth of risk-informed requirements for pressure-retaining items. This initiative, which is built from prior ASME Section XI risk-informed inservice inspection developments over the past decade, has been undertaken in conjunction with U.S. risk-informed regulation efforts. The U.S. Nuclear Regulatory Commission (NRC) is working with the industry on risk informing Title 10 Code of Federal Regulations Part 50 (10CFR50). The Nuclear Regulatory Commission's basic proposal is to allow modification of some of the special treatment requirements of 10CFR50. Their effort is proceeding via an Advanced Notice of Public Rulemaking, March 3, 2000, and an announcement of Availability of Draft Rule Wording, November 29, 2001, to add 10 CFR 50.69, 'Risk-Informed Treatment of Structures, Systems and Components'. A parallel task by the Nuclear Energy Institute (NEI) to develop a guideline on how to implement the results of the rulemaking is also well underway via NEI 00-04 (Draft Revision B), 'Option 2 Implementation Guideline', May 2001. This paper summarizes the content and status of approval of the proposed ASME Code Cases, including how they relate to the above NRC and NEI efforts. Some initial results from trial application of the Code Cases will also be cited. (authors)

Information sets as permutation cycles for quadratic residue codes

Directory of Open Access Journals (Sweden)

Richard A. Jenson

1982-01-01

Full Text Available The two cases p=7 and p=23 are the only known cases where the automorphism group of the [p+1,   (p+1/2] extended binary quadratic residue code, O(p, properly contains PSL(2,p. These codes have some of their information sets represented as permutation cycles from Aut(Q(p. Analysis proves that all information sets of Q(7 are so represented but those of Q(23 are not.

SSC-K code users manual (rev.1)

International Nuclear Information System (INIS)

Kwon, Y. M.; Lee, Y. B.; Chang, W. P.; Hahn, D.

2002-01-01

The Supper System Code of KAERI (SSC-K) is a best-estimate system code for analyzing a variety of off-normal or accidents in the heat transport system of a pool type LMR design. It is being developed at Korea Atomic Energy Research Institution (KAERI) on the basis of SSC-L, originally developed at BNL to analyze loop-type LMR transients. SSC-K can handle both designs of loop and pool type LMRs. SSC-K contains detailed mechanistic models of transient thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, coolant, fuel elements, and structures to accident conditions. This report provides a revised User's Manual (rev.1) of the SSC-K computer code, focusing on phenomenological model descriptions for new thermal, hydraulic, neutronic, and mechanical modules. A comprehensive description of the models for pool-type reactor is given in Chapters 2 and 3; the steady-state plant characterization, prior to the initiation of transient is described in Chapter 2 and their transient counterparts are discussed in Chapter 3. Discussions on the intermediate heat exchanger (IHX) and the electromagnetic (EM) pump are described in Chapter 4 and 5, respectively. A model of passive safety decay heat removal system (PSDRS) is discussed in Chapter 6, and models for various reactivity feedback effects are discussed in Chapter 7. In Chapter 8, constitutive laws and correlations required to execute the SSC-K are described. New models developed for SSC-K rev.1 are two dimensional hot pool model in Chapter 9, and long term cooling model in Chapter 10. Finally, a brief description of MINET code adopted to simulate BOP is presented in Chapter 11. Based on test runs for typical LMFBR accident analyses, it was found that the present version of SSC-K would be used for the safety analysis of KALIMER. However, the further validation of SSC-K is required for real applications. It is noted that the user's manual of SSC-K will be revised later with the

Benchmarking (Code2Code) of the 1Hs 3-Bladed Onshore VAWT

DEFF Research Database (Denmark)

Galinos, Christos; Schmidt Paulsen, Uwe

This study is part of the Inflow project. In this report the Nenuphar’s onshore 3-bladed Vertical Axis Wind Turbine (VAWT) prototype (1HS) is modelled in HAWC2 aeroelastic code. In the first part the model properties are summarized. Then the analysis is focused on the rotor performance and various...

Benchmark calculation of subchannel analysis codes

International Nuclear Information System (INIS)

1996-02-01

In order to evaluate the analysis capabilities of various subchannel codes used in thermal-hydraulic design of light water reactors, benchmark calculations were performed. The selected benchmark problems and major findings obtained by the calculations were as follows: (1)As for single-phase flow mixing experiments between two channels, the calculated results of water temperature distribution along the flow direction were agreed with experimental results by tuning turbulent mixing coefficients properly. However, the effect of gap width observed in the experiments could not be predicted by the subchannel codes. (2)As for two-phase flow mixing experiments between two channels, in high water flow rate cases, the calculated distributions of air and water flows in each channel were well agreed with the experimental results. In low water flow cases, on the other hand, the air mixing rates were underestimated. (3)As for two-phase flow mixing experiments among multi-channels, the calculated mass velocities at channel exit under steady-state condition were agreed with experimental values within about 10%. However, the predictive errors of exit qualities were as high as 30%. (4)As for critical heat flux(CHF) experiments, two different results were obtained. A code indicated that the calculated CHF's using KfK or EPRI correlations were well agreed with the experimental results, while another code suggested that the CHF's were well predicted by using WSC-2 correlation or Weisman-Pei mechanistic model. (5)As for droplets entrainment and deposition experiments, it was indicated that the predictive capability was significantly increased by improving correlations. On the other hand, a remarkable discrepancy between codes was observed. That is, a code underestimated the droplet flow rate and overestimated the liquid film flow rate in high quality cases, while another code overestimated the droplet flow rate and underestimated the liquid film flow rate in low quality cases. (J.P.N.)

[Quality assurance in coding expertise of hospital cases in the German DRG system. Evaluation of inter-rater reliability in MDK expertise].

Science.gov (United States)

Huber, H; Brambrink, M; Funk, R; Rieger, M

2012-10-01

The purpose of this study was to evaluate differences in the D-DRG results of a hospital case by 2 independently coding MKD raters. Calculation of the 2-inter-rater reliability was performed by examination of the coding of individual hospital cases. The reasons for the non-agreement of the expert evaluations and suggestions to improve the process are discussed. From the expert evaluation pool of the MDK-WL a random sample of 0.7% of the 57,375 expertises was taken. Distribution equality with the basic total was tested by the χ² test or, respectively, Fisher's exact test. For the total of 402 individual hospital cases, the G-DRG case sums of 2 experts of the MDK were determined independently and the results checked for each individual case for agreement or non-agreement. The corresponding confidence intervals with standard errors were analysed to test if certain major diagnosis categories (MDC) were statistically significantly more affected by differing expertise results than others. In 280 of the total 402 tested hospital cases, the 2 MDK raters independently reached the same G-DRG results; in 122 cases the G-DRG case sums determined by the 2 raters differed (agreement 70%; CI 65.2-74.1). Different DRG results between the 2 experts occurred regularly in the entire MDC spectrum. No MDC chapter in which significant differences between the 2 raters arose could be identified. The results of our study demonstrate an almost 70% agreement in the evaluation of hospital cost accounts by 2 independently operating MDK. This result leaves room for improvement. Optimisation potentials can be recognised on the basis of the results. Potential for improvement was established in combination with regular further training and the expansion of binding internal code recommendations as well as exchange of code-relevant information among experts in internal forums. The presented model is in principle suitable for cross-border examinations within the MDK system with the advantage that

Generalized concatenated quantum codes

International Nuclear Information System (INIS)

Grassl, Markus; Shor, Peter; Smith, Graeme; Smolin, John; Zeng Bei

2009-01-01

We discuss the concept of generalized concatenated quantum codes. This generalized concatenation method provides a systematical way for constructing good quantum codes, both stabilizer codes and nonadditive codes. Using this method, we construct families of single-error-correcting nonadditive quantum codes, in both binary and nonbinary cases, which not only outperform any stabilizer codes for finite block length but also asymptotically meet the quantum Hamming bound for large block length.

Validation and optimisation of an ICD-10-coded case definition for sepsis using administrative health data

Science.gov (United States)

Jolley, Rachel J; Jetté, Nathalie; Sawka, Keri Jo; Diep, Lucy; Goliath, Jade; Roberts, Derek J; Yipp, Bryan G; Doig, Christopher J

2015-01-01

Objective Administrative health data are important for health services and outcomes research. We optimised and validated in intensive care unit (ICU) patients an International Classification of Disease (ICD)-coded case definition for sepsis, and compared this with an existing definition. We also assessed the definition's performance in non-ICU (ward) patients. Setting and participants All adults (aged ≥18 years) admitted to a multisystem ICU with general medicosurgical ICU care from one of three tertiary care centres in the Calgary region in Alberta, Canada, between 1 January 2009 and 31 December 2012 were included. Research design Patient medical records were randomly selected and linked to the discharge abstract database. In ICU patients, we validated the Canadian Institute for Health Information (CIHI) ICD-10-CA (Canadian Revision)-coded definition for sepsis and severe sepsis against a reference standard medical chart review, and optimised this algorithm through examination of other conditions apparent in sepsis. Measures Sensitivity (Sn), specificity (Sp), positive predictive value (PPV) and negative predictive value (NPV) were calculated. Results Sepsis was present in 604 of 1001 ICU patients (60.4%). The CIHI ICD-10-CA-coded definition for sepsis had Sn (46.4%), Sp (98.7%), PPV (98.2%) and NPV (54.7%); and for severe sepsis had Sn (47.2%), Sp (97.5%), PPV (95.3%) and NPV (63.2%). The optimised ICD-coded algorithm for sepsis increased Sn by 25.5% and NPV by 11.9% with slightly lowered Sp (85.4%) and PPV (88.2%). For severe sepsis both Sn (65.1%) and NPV (70.1%) increased, while Sp (88.2%) and PPV (85.6%) decreased slightly. Conclusions This study demonstrates that sepsis is highly undercoded in administrative data, thus under-ascertaining the true incidence of sepsis. The optimised ICD-coded definition has a higher validity with higher Sn and should be preferentially considered if used for surveillance purposes. PMID:26700284

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

International Nuclear Information System (INIS)

Tai Asayama; Masaki Morishita; Masanori Tashimo

2006-01-01

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

Improvement of multi-dimensional realistic thermal-hydraulic system analysis code, MARS 1.3

International Nuclear Information System (INIS)

Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok

1998-09-01

The MARS (Multi-dimensional Analysis of Reactor Safety) code is a multi-dimensional, best-estimate thermal-hydraulic system analysis code. This report describes the new features that have been improved in the MARS 1.3 code since the release of MARS 1.3 in July 1998. The new features include: - implementation of point kinetics model into the 3D module - unification of the heat structure model - extension of the control function to the 3D module variables - improvement of the 3D module input check function. Each of the items has been implemented in the developmental version of the MARS 1.3.1 code and, then, independently verified and assessed. The effectiveness of the new features is well verified and it is shown that these improvements greatly extend the code capability and enhance the user friendliness. Relevant input data changes are also described. In addition to the improvements, this report briefly summarizes the future code developmental activities that are being carried out or planned, such as coupling of MARS 1.3 with the containment code CONTEMPT and the three-dimensional reactor kinetics code MASTER 2.0. (author). 8 refs

Improvement of multi-dimensional realistic thermal-hydraulic system analysis code, MARS 1.3

Energy Technology Data Exchange (ETDEWEB)

Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok

1998-09-01

The MARS (Multi-dimensional Analysis of Reactor Safety) code is a multi-dimensional, best-estimate thermal-hydraulic system analysis code. This report describes the new features that have been improved in the MARS 1.3 code since the release of MARS 1.3 in July 1998. The new features include: - implementation of point kinetics model into the 3D module - unification of the heat structure model - extension of the control function to the 3D module variables - improvement of the 3D module input check function. Each of the items has been implemented in the developmental version of the MARS 1.3.1 code and, then, independently verified and assessed. The effectiveness of the new features is well verified and it is shown that these improvements greatly extend the code capability and enhance the user friendliness. Relevant input data changes are also described. In addition to the improvements, this report briefly summarizes the future code developmental activities that are being carried out or planned, such as coupling of MARS 1.3 with the containment code CONTEMPT and the three-dimensional reactor kinetics code MASTER 2.0. (author). 8 refs.

A toroidal plasma MHD equilibrium code 'EQUCIR version 1'

International Nuclear Information System (INIS)

Ninomiya, Hiromasa; Shinya, Kichiro; Kameari, Akihisa.

1980-10-01

A new free-boundary toroidal MHD equilibrium code ''EQUCIR version 1'' has been developed. The central problems approached by this code is as follows: 1) The magnetic flux distribution of a plasma at equilibrium is determined in the given external field. 2) A set of circuit equations between the plasma and the external conductors are constructed. These circuit equations and the Grad-Shafranov equation are solved self-consistently and the time evolutions of plasma equilibria and currents in external conductors are determined at the same time. 3) The currents in the external conductors are determined so that the plasma cross-section and plasma parameters are to be maintained with desired ones. It is shown that this code is very useful for study of the tokamak plasma equilibria, for design of the poloidal coil system and for investigation of experimental results. (author)

Assessment of ICARE/CATHARE V1 Severe Accident Code

International Nuclear Information System (INIS)

Chatelard, Patrick; Fleurot, Joelle; Marchand, Olivier; Drai, Patrick

2006-01-01

The ICARE/CATHARE code system has been developed by the French 'Institut de Radioprotection et de Surete Nucleaire' (IRSN) in the last decade for the detailed evaluation of Severe Accident (SA) consequences in a primary system. It is composed of the coupling of the core degradation IRSN code ICARE2 and of the thermal-hydraulics French code CATHARE2. It has been extensively used to support the level 2 Probabilistic Safety Assessment (PSA-2) of the 900 MWe PWR. This paper presents the synthesis of the ICARE/CATHARE V1 assessment which was conducted in the frame of the 'International ICARE/CATHARE Users' Club', under the management of IRSN. The ICARE/CATHARE V1 validation matrix is composed of more than 60 experiments, distributed in few thermal-hydraulics non-regression tests (to handle the front end phase of a severe accident), numerous Separate-Effect Tests, about 30 Integral Tests covering both the early and the late degradation phases, as well as a 'circuit' experiment including hydraulics loops. Finally, the simulation of the TMI-2 accident was also added to assess the code against real conditions. This validation task was aimed at assessing the ICARE/CATHARE V1 capabilities (including the stand-alone ICARE2 V3mod1 version) and also at proposing recommendations for an optimal use of this version ('Users' Guidelines'). Thus, with a correct account for the recommended guidelines, it appeared that the last ICARE/CATHARE V1 version could be reasonably used to perform best-estimate reactor studies up to a large corium slumping into the lower head. (authors)

The computer code EURDYN-1M (release 2). User's manual

International Nuclear Information System (INIS)

1982-01-01

EURDYN-1M is a finite element computer code developed at J.R.C. Ispra to compute the response of two-dimensional coupled fluid-structure configurations to transient dynamic loading for reactor safety studies. This report gives instructions for preparing input data to EURDYN-1M, release 2, and describes a test problem in order to illustrate both the input and the output of the code

Manual phased arrays for weld inspections using North American codes

International Nuclear Information System (INIS)

Moles, Michael

2008-01-01

Phased arrays are primarily a method of generating and receiving ultrasound, not a new technology. In addition, the physics of ultrasound generated by phased arrays is identical to that from conventional monocrystals. Not surprisingly, all the major North American (and some European) codes accept phased arrays, either explicitly or implicitly. However, the technique and procedures needs to be proven, typically by a Performance Demonstration. The ASME (AmeicanSociety for Mechanical Engineers) Section V and API RP2X explicitly accept phased arrays. Three ASME code cases have been written specifically fo manual phased array: Code Cases 2541. 2557 and 2558. Over and above the general requirements of Article 4, these Code Cases require full waveform calibration. This is echoed in ASTM E-2491, a Standard Guide for setting up phased arrays. In addition. details such as focusing and reporting are addressed. The American Petroleum Institute QUTE procedure did not need any modifications to be compatible with manual phased arrays. The American Welding Society (AWS) Structural Welding Code D1.1 implicitly accepts phased arrays. New technologies such as phased arrays can be proven using Annex K. Nonetheless, a manual phased array unit using the standard AWS probe and displaying 45, 60 and 70degrees waveforms would be acceptable for D1.1 a s is . Overall, most major North American codes accept phased arrays, however, the technique and procedures must be proven, often using a Performance Demonstration. (author)

Reduced levels of potential circulating biomarkers of cardiovascular diseases in apparently healthy vegetarian men.

Science.gov (United States)

Navarro, Julio Acosta; de Gouveia, Luiza Antoniazzi; Rocha-Penha, Lilliam; Cinegaglia, Naiara; Belo, Vanessa; Castro, Michele Mazzaron de; Sandrim, Valeria Cristina

2016-10-01

Several evidences report that a vegetarian diet is protector against cardiovascular diseases. Few studies have demonstrated the circulating profile of cardiovascular biomarkers in vegetarians. Therefore, the aims of the current study were compared the plasma concentrations of myeloperoxidase (MPO), metalloproteinase (MMP)-9, MMP-2, tissue inhibitor of MMP (TIMP)-1 and TIMP-2 between healthy vegetarian (Veg) and healthy omnivorous (Omn). Using ELISA and multiplexed bead immunoassay, we measured in plasma from 43 Veg and 41 Omn the cardiovascular biomarkers concentrations cited above. We found significant lower concentrations of MPO, MMP-9, MMP-2 and MMP-9/TIMP-1 ratio in Veg compared to Omn (all Pvegetarian diet is associated with a healthier profile of cardiovascular biomarkers compared to omnivorous. Copyright © 2016 Elsevier B.V. All rights reserved.

Regulation and innovation dynamics for nanoresponsible development: The case of the French code de l'environnement L 523-1 to L 523-5

International Nuclear Information System (INIS)

Auplat, C; Slimane, S Ben

2015-01-01

This paper examines one aspect of innovation dynamics for nanoresponsible development: the links between regulation and innovation dynamics. It focuses on the case of the French Code de l'environnement, Articles L. 523-1 to L. 523-3. Articles L. 523-1 to L. 523-3 of the French environment code provide for the obligation to declare the quantities and uses of substances at nanoscale produced, distributed or imported in France. This procedure is intended to improve knowledge of these substances and their uses as well as of their markets and volumes sold, to ensure traceability and to collect available information on their toxicological and ecotoxicological properties. The paper builds on recent work on the emergence of a regulatory framework for nanotechnologies to take stock of the current situation in France, in the EU and globally and to explore how this specific law package may influence innovation and the shaping of new markets for nanobased materials. The study shows that nano-regulation does have an impact on innovation. However, the impact is not the same with EU regulation and with French regulation, and while EU regulation seems to create a favourable context for innovation, French regulation seems to do the opposite. With this study we hope to bring new perspectives to the field of the strategic management of innovation, and also to shed some light on the roles and challenges of institutions to facilitate nanoresponsible development. (paper)

Case-crossover analysis of heat-coded deaths and vulnerable subpopulations: Oklahoma, 1990-2011

Science.gov (United States)

Moore, Brianna F.; Brooke Anderson, G.; Johnson, Matthew G.; Brown, Sheryll; Bradley, Kristy K.; Magzamen, Sheryl

2017-11-01

The extent of the association between temperature and heat-coded deaths, for which heat is the primary cause of death, remains largely unknown. We explored the association between temperature and heat-coded deaths and potential interactions with various demographic and environmental factors. A total of 335 heat-coded deaths that occurred in Oklahoma from 1990 through 2011 were identified using heat-related International Classification of Diseases codes, cause-of-death nomenclature, and narrative descriptions. Conditional logistic regression models examined the association between temperature and heat index on heat-coded deaths. Interaction by demographic factors (age, sex, marital status, living alone, outdoor/heavy labor occupations) and environmental factors (ozone, PM10, PM2.5) was also explored. Temperatures ≥99 °F (the median value) were associated with approximately five times higher odds of a heat-coded death as compared to temperatures effect estimates were attenuated when exposure to heat was characterized by heat index. The interaction results suggest that effect of temperature on heat-coded deaths may depend on sex and occupation. For example, the odds of a heat-coded death among outdoor/heavy labor workers exposed to temperatures ≥99 °F was greater than expected based on the sum of the individual effects (observed OR = 14.0, 95% CI 2.7, 72.0; expected OR = 4.1 [2.8 + 2.3-1.0]). Our results highlight the extent of the association between temperature and heat-coded deaths and emphasize the need for a comprehensive, multisource definition of heat-coded deaths. Furthermore, based on the interaction results, we recommend that states implement or expand heat safety programs to protect vulnerable subpopulations, such as outdoor workers.

Code-Mixing and Code Switchingin The Process of Learning

Directory of Open Access Journals (Sweden)

Diyah Atiek Mustikawati

2016-09-01

Full Text Available This study aimed to describe a form of code switching and code mixing specific form found in the teaching and learning activities in the classroom as well as determining factors influencing events stand out that form of code switching and code mixing in question.Form of this research is descriptive qualitative case study which took place in Al Mawaddah Boarding School Ponorogo. Based on the analysis and discussion that has been stated in the previous chapter that the form of code mixing and code switching learning activities in Al Mawaddah Boarding School is in between the use of either language Java language, Arabic, English and Indonesian, on the use of insertion of words, phrases, idioms, use of nouns, adjectives, clauses, and sentences. Code mixing deciding factor in the learning process include: Identification of the role, the desire to explain and interpret, sourced from the original language and its variations, is sourced from a foreign language. While deciding factor in the learning process of code, includes: speakers (O1, partners speakers (O2, the presence of a third person (O3, the topic of conversation, evoke a sense of humour, and just prestige. The significance of this study is to allow readers to see the use of language in a multilingual society, especially in AL Mawaddah boarding school about the rules and characteristics variation in the language of teaching and learning activities in the classroom. Furthermore, the results of this research will provide input to the ustadz / ustadzah and students in developing oral communication skills and the effectiveness of teaching and learning strategies in boarding schools.

An assessment of the CORCON-MOD3 code. Part 1: Thermal-hydraulic calculations

International Nuclear Information System (INIS)

Strizhov, V.; Kanukova, V.; Vinogradova, T.; Askenov, E.; Nikulshin, V.

1996-09-01

This report deals with the subject of CORCON-Mod3 code validation (thermal-hydraulic modeling capability only) based on MCCI (molten core concrete interaction) experiments conducted under different programs in the past decade. Thermal-hydraulic calculations (i.e., concrete ablation, melt temperature, melt energy, concrete temperature, and condensible and non-condensible gas generation) were performed with the code, and compared with the data from 15 experiments, conducted at different scales using both simulant (metallic and oxidic) and prototypic melt materials, using different concrete types, and with and without an overlying water pool. Sensitivity studies were performed in a few cases involving, for example, heat transfer from melt to concrete, condensed phase chemistry, etc. Further, special analysis was performed using the ACE L8 experimental data to illustrate the differences between the experimental and the reactor conditions, and to demonstrate that with proper corrections made to the code, the calculated results were in better agreement with the experimental data. Generally, in the case of dry cavity and metallic melts, CORCON-Mod3 thermal-hydraulic calculations were in good agreement with the test data. For oxidic melts in a dry cavity, uncertainties in heat transfer models played an important role for two melt configurations--a stratified geometry with segregated metal and oxide layers, and a heterogeneous mixture. Some discrepancies in the gas release data were noted in a few cases

The Drosophila gene CG9918 codes for a pyrokinin-1 receptor

DEFF Research Database (Denmark)

Cazzamali, Giuseppe; Torp, Malene; Hauser, Frank

2005-01-01

The database from the Drosophila Genome Project contains a gene, CG9918, annotated to code for a G protein-coupled receptor. We cloned the cDNA of this gene and functionally expressed it in Chinese hamster ovary cells. We tested a library of about 25 Drosophila and other insect neuropeptides......, and seven insect biogenic amines on the expressed receptor and found that it was activated by low concentrations of the Drosophila neuropeptide, pyrokinin-1 (TGPSASSGLWFGPRLamide; EC50, 5 x 10(-8) M). The receptor was also activated by other Drosophila neuropeptides, terminating with the sequence PRLamide...... (Hug-gamma, ecdysis-triggering-hormone-1, pyrokinin-2), but in these cases about six to eight times higher concentrations were needed. The receptor was not activated by Drosophila neuropeptides, containing a C-terminal PRIamide sequence (such as ecdysis-triggering-hormone-2), or PRVamide (such as capa...

OSCAR-4 Code System Application to the SAFARI-1 Reactor

International Nuclear Information System (INIS)

Stander, Gerhardt; Prinsloo, Rian H.; Tomasevic, Djordje I.; Mueller, Erwin

2008-01-01

The OSCAR reactor calculation code system consists of a two-dimensional lattice code, the three-dimensional nodal core simulator code MGRAC and related service codes. The major difference between the new version of the OSCAR system, OSCAR-4, and its predecessor, OSCAR-3, is the new version of MGRAC which contains many new features and model enhancements. In this work some of the major improvements in the nodal diffusion solution method, history tracking, nuclide transmutation and cross section models are described. As part of the validation process of the OSCAR-4 code system (specifically the new MGRAC version), some of the new models are tested by comparing computational results to SAFARI-1 reactor plant data for a number of operational cycles and for varying applications. A specific application of the new features allows correct modeling of, amongst others, the movement of fuel-follower type control rods and dynamic in-core irradiation schedules. It is found that the effect of the improved control rod model, applied over multiple cycles of the SAFARI-1 reactor operation history, has a significant effect on in-cycle reactivity prediction and fuel depletion. (authors)

Independent assessment of the TRAC-BD1/MOD1 computer code at the Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

Wilson, G.E.; Charboneau, B.L.; Dallman, R.J.; Kullberg, C.M.; Wagner, K.C.; Wheatley, P.D.

1984-01-01

Under auspices of the United States Nuclear Regulatory Commission, their primary boiling water reactor safety analysis code (TRAC-BWR) is being assessed with simulations of a wide range of experimental data. The FY-1984 assessment activities were associated with the latest version (TRAC-BD1/MOD1) of this code. Typical results of the assessment studies are given. Conclusions formulated from these results are presented. These calculations relate to the overall applicability of the current code to safety analysis, and to future work which would further enhance the code's quality and ease of use

Automated Source Code Analysis to Identify and Remove Software Security Vulnerabilities: Case Studies on Java Programs

OpenAIRE

Natarajan Meghanathan

2013-01-01

The high-level contribution of this paper is to illustrate the development of generic solution strategies to remove software security vulnerabilities that could be identified using automated tools for source code analysis on software programs (developed in Java). We use the Source Code Analyzer and Audit Workbench automated tools, developed by HP Fortify Inc., for our testing purposes. We present case studies involving a file writer program embedded with features for password validation, and ...

ESCADRE and ICARE code systems

International Nuclear Information System (INIS)

Reocreux, M.; Gauvain, J.

1992-01-01

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

Code-B-1 for stress/strain calculation for TRISO fuel particle (Contract research)

International Nuclear Information System (INIS)

Aihara, Jun; Ueta, Shohei; Shibata, Taiju; Sawa, Kazuhiro

2011-12-01

We have developed Code-B-1 for the prediction of the failure probabilities of the coated fuel particles for the high temperature gas-cooled reactors (HTGRs) under operation by modification of an existing code. A finite element method (FEM) is employed for the stress calculation part and Code-B-1 can treat the plastic deformation of the coating layer of the coated fuel particles which the existing code cannot treat. (author)

Coding in Muscle Disease.

Science.gov (United States)

Jones, Lyell K; Ney, John P

2016-12-01

Accurate coding is critically important for clinical practice and research. Ongoing changes to diagnostic and billing codes require the clinician to stay abreast of coding updates. Payment for health care services, data sets for health services research, and reporting for medical quality improvement all require accurate administrative coding. This article provides an overview of administrative coding for patients with muscle disease and includes a case-based review of diagnostic and Evaluation and Management (E/M) coding principles in patients with myopathy. Procedural coding for electrodiagnostic studies and neuromuscular ultrasound is also reviewed.

International Observe the Moon Night - An Opportunity to Participate in the Year of the Solar System While Sharing the Excitement of Lunar Science and Exploration with the Public

Science.gov (United States)

Bleacher, L.; Daou, D.; Day, B. H.; Hsu, B. C.; Jones, A. P.; Mitchell, B.; Shaner, A. J.; Shipp, S. S.

2010-12-01

International Observe the Moon Night (InOMN) is a multi-nation effort to share the excitement of recent lunar missions and new science results with education communities, amateur astronomers, space enthusiasts, and the general public. It is also intended to encourage the world to experience the thrill of observing Earth’s closest neighbor. The inaugural InOMN took place on September 18, 2010. People in over 26 countries gathered together in groups big and small to learn about the Moon through presentations by scientists, astronomers, and engineers; participate in hands-on activities; and observe the Moon through telescopes, binoculars, and the naked eye. Next year’s InOMN will take place on October 8, 2011 during the Year of the Solar System (YSS). The October 2011 YSS theme will be “Moons/Rings Across the Solar System.” InOMN is perfectly suited as an event that any museum, science center, planetarium, university, school, or other group can implement to celebrate YSS. The InOMN Coordinating Committee has developed a variety of resources and materials to make it easy to host an InOMN event of any size. Interested groups are encouraged to utilize the InOMN website (observethemoonnight.org) in planning their InOMN event for 2011/YSS. The website contains links to Moon resources, educational activities, suggestions for hosting an event, free downloads of logos and flyers for advertising an event, and contests. New for 2011 will be a discussion forum for event hosts to share their plans, tips, and experiences. Together, YSS and InOMN will enable the public to maintain its curiosity about the Moon and to gain a better understanding of the Moon’s formation, evolution, and place in the night sky.

Experimental validation for combustion analysis of GOTHIC 6.1b code in 2-dimensional premixed combustion experiments

International Nuclear Information System (INIS)

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

2003-01-01

In this study, the prediction capability of GOTHIC code for hydrogen combustion phenomena was validated with the results of two-dimensional premixed hydrogen combustion experiment executed by Seoul National University. In the experimental results, we could confirm the propagation characteristics of hydrogen flame such as buoyancy effect, flame front shape etc.. The combustion time of the tests was about 0.1 sec.. In the GOTHIC analyses results, the GOTHIC code could predict the overall hydrogen flame propagation characteristics but the buoyancy effect and flame shape did not compare well with the experimental results. Especially, in case of the flame propagate to the dead-end, GOTHIC predicted the flame did not affected by the flow and this cause quite different results in flame propagation from experimental results. Moreover the combustion time of the analyses was about 1 sec. which is ten times longer than the experimental result. To obtain more reasonable analysis results, it is necessary that combustion model parameters in GOTHIC code apply appropriately and hydrogen flame characteristics be reflected in solving governing equations

Use of diagnosis codes for detection of clinically significant opioid poisoning in the emergency department: A retrospective analysis of a surveillance case definition.

Science.gov (United States)

Reardon, Joseph M; Harmon, Katherine J; Schult, Genevieve C; Staton, Catherine A; Waller, Anna E

2016-02-08

Although fatal opioid poisonings tripled from 1999 to 2008, data describing nonfatal poisonings are rare. Public health authorities are in need of tools to track opioid poisonings in near real time. We determined the utility of ICD-9-CM diagnosis codes for identifying clinically significant opioid poisonings in a state-wide emergency department (ED) surveillance system. We sampled visits from four hospitals from July 2009 to June 2012 with diagnosis codes of 965.00, 965.01, 965.02 and 965.09 (poisoning by opiates and related narcotics) and/or an external cause of injury code of E850.0-E850.2 (accidental poisoning by opiates and related narcotics), and developed a novel case definition to determine in which cases opioid poisoning prompted the ED visit. We calculated the percentage of visits coded for opioid poisoning that were clinically significant and compared it to the percentage of visits coded for poisoning by non-opioid agents in which there was actually poisoning by an opioid agent. We created a multivariate regression model to determine if other collected triage data can improve the positive predictive value of diagnosis codes alone for detecting clinically significant opioid poisoning. 70.1 % of visits (Standard Error 2.4 %) coded for opioid poisoning were primarily prompted by opioid poisoning. The remainder of visits represented opioid exposure in the setting of other primary diseases. Among non-opioid poisoning codes reviewed, up to 36 % were reclassified as an opioid poisoning. In multivariate analysis, only naloxone use improved the positive predictive value of ICD-9-CM codes for identifying clinically significant opioid poisoning, but was associated with a high false negative rate. This surveillance mechanism identifies many clinically significant opioid overdoses with a high positive predictive value. With further validation, it may help target control measures such as prescriber education and pharmacy monitoring.

Successive Transfers Relating to Movable Tangible Assets and Acquisition of Property under Article 937, Paragraph (1 of the Civil Code

Directory of Open Access Journals (Sweden)

Mara Ioan

2015-12-01

Full Text Available Apparently article 1275, paragraph (1 of the Civil Code covers all situations that may arise in practice, without making a distinction for the constituent or transferring contracts if they are of the same or of different nature. However, we appreciate that article 1275 of the Civil Code does not apply in all situations of successive transfers relating to movable tangible property granted by the same legal subject. Corroborating this text with the norms in article 937 paragraph (1 of the Civil Code and article1273 paragraph (1 of the Civil Code it leads to the solution according to which article 1275 of the Civil Code regards only the cases where the transfer of successive property are of the same nature, the onerous primary act has not resulted in immediate transmission of real previous right of the document with the free subsidiary title and when the primal act is free, and the alternative is onerous. It is excluded, thus from the application of the rule in question when the primary onerous act had as effect the immediate transmission of the real right and then, but without having occurred the delivery of the asset by the acquirer, it was concluded a document with a free title, subsidiary.

Complete permutation Gray code implemented by finite state machine

Directory of Open Access Journals (Sweden)

Li Peng

2014-09-01

Full Text Available An enumerating method of complete permutation array is proposed. The list of n! permutations based on Gray code defined over finite symbol set Z(n = {1, 2, …, n} is implemented by finite state machine, named as n-RPGCF. An RPGCF can be used to search permutation code and provide improved lower bounds on the maximum cardinality of a permutation code in some cases.

Penal Code (Ordinance No. 12 of 1983), 1 July 1984.

Science.gov (United States)

1987-01-01

This document contains provisions of the 1984 Penal Code of Montserrat relating to sexual offenses, abortion, offenses relating to marriage, homicide and other offenses against the person, and neglect endangering life or health. Part 8 of the Code holds that a man found guilty of raping a woman is liable to life imprisonment. Rape is deemed to involve unlawful (extramarital) sexual intercourse with a woman without her consent (this is determined if the rape involved force, threats, administration of drugs, or false representation). The Code also defines offenses in cases of incest, child abuse, prostitution, abduction, controlling the actions and finances of a prostitute, and having unlawful sexual intercourse with a mentally defective woman. Part 9 of the Code outlaws abortion unless it is conducted in an approved establishment after two medical practitioners have determined that continuing the pregnancy would risk the life or physical/mental health of the pregnant woman or if a substantial risk exists that the child would have serious abnormalities. Part 10 outlaws bigamy, and part 12 holds that infanticide performed by a mother suffering postpartum imbalances can be prosecuted as manslaughter. This part also outlaws concealment of the body of a newborn, whether that child died before, at, or after birth, and aggravated assault on any child not more than 14 years old. Part 12 makes it an offense to subject any child to neglect endangering its life or health.

Combining loop unrolling strategies and code predication to reduce the worst-case execution time of real-time software

Directory of Open Access Journals (Sweden)

Andreu Carminati

2017-07-01

Full Text Available Worst-case execution time (WCET is a parameter necessary to guarantee timing constraints on real-time systems. The higher the worst-case execution time of tasks, the higher will be the resource demand for the associated system. The goal of this paper is to propose a different way to perform loop unrolling on data-dependent loops using code predication targeting WCET reduction, because existing techniques only consider loops with fixed execution counts. We also combine our technique with existing unrolling approaches. Results showed that this combination can produce aggressive WCET reductions when compared with the original code.

The PLTEMP V2.1 code

International Nuclear Information System (INIS)

Olson, A.P.

2003-01-01

Recent improvements to the computer code PLTEMP/ANL V2.1 are described. A new iterative, error-minimization solution technique is used to obtain the thermal distribution both within each fuel plate, and along the axial length of each coolant channel. A new, radial geometry solution is available for tube-type fuel assemblies. Software comparisons of these and other new models are described. Applications to Russian-designed IRT-type research reactors are described. (author)

LINK codes TRAC-BF1/PARCSv2.7 in LINUX without external communication interface

International Nuclear Information System (INIS)

Barrachina, T.; Garcia-Fenoll, M.; Abarca, A.; Miro, R.; Verdu, G.; Concejal, A.; Solar, A.

2014-01-01

The TRAC-BF1 code is still widely used by the nuclear industry for safety analysis. The plant models developed using this code are highly validated, so it is advisable to continue improving this code before migrating to another completely different code. The coupling with the NRC neutronic code PARCSv2.7 increases the simulation capabilities in transients in which the power distribution plays an important role. In this paper, the procedure for the coupling of TRAC-BF1 and PARCSv2.7 codes without PVM and in Linux is presented. (Author)

Validation of the THIRMAL-1 melt-water interaction code

Energy Technology Data Exchange (ETDEWEB)

Chu, C.C.; Sienicki, J.J.; Spencer, B.W. [Argonne National Lab., IL (United States)

1995-09-01

The THIRMAL-1 computer code has been used to calculate nonexplosive LWR melt-water interactions both in-vessel and ex-vessel. To support the application of the code and enhance its acceptability, THIRMAL-1 has been compared with available data from two of the ongoing FARO experiments at Ispra and two of the Corium Coolant Mixing (CCM) experiments performed at Argonne. THIRMAL-1 calculations for the FARO Scoping Test and Quenching Test 2 as well as the CCM-5 and -6 experiments were found to be in excellent agreement with the experiment results. This lends confidence to the modeling that has been incorporated in the code describing melt stream breakup due to the growth of both Kelvin-Helmholtz and large wave instabilities, the sizes of droplets formed, multiphase flow and heat transfer in the mixing zone surrounding and below the melt metallic phase. As part of the analysis of the FARO tests, a mechanistic model was developed to calculate the prefragmentation as it may have occurred when melt relocated from the release vessel to the water surface and the model was compared with the relevant data from FARO.

Evaluation of control room habitability in case of LOCA for Maanshan NPP using codes RADTRAD, HABIT and ALOHA

International Nuclear Information System (INIS)

Hsu, Wen-Sheng; Wang, Jong-Rong; Chen, Hsiung-Chih; Chiang, Yu; Chen, Shao-Wen; Shih, Chunkuan

2018-01-01

The method for the evaluation of the control room habitability is presented in this paper with focus on Maanshan PWR nuclear power plant (NPP) using the codes RADTRAD, HABIT, and ALOHA. Therefore, this paper is divided into two parts: The first part is the evaluation of the cumulative dose at the control room, the exclusion area boundary (EAB) and the low population zone (LPZ) in case of an design basis loss of coolant accident (DBA/LOCA). For this first part, the Maanshan NPP models of the code RADTRAD/SNAP were used for the analysis. The second part is the evaluation of the control room habitability under the assumption of CO 2 storage burst. For this part, the HABIT and ALOHA codes were used. As result it was seen that the RADTRAD calculation results are below the failure criteria of standard review plan (SRP) and 10 CFR 100.11. The HABIT and ALOHA results are below the R.G. 1.78 failure criteria. These results indicate that Maanshan NPP' habitability can be maintained under the above conditions.

26 CFR 1.997-1 - Special rules for subchapter C of the Code.

Science.gov (United States)

2010-04-01

... 26 Internal Revenue 10 2010-04-01 2010-04-01 false Special rules for subchapter C of the Code. 1.997-1 Section 1.997-1 Internal Revenue INTERNAL REVENUE SERVICE, DEPARTMENT OF THE TREASURY (CONTINUED) INCOME TAX (CONTINUED) INCOME TAXES Domestic International Sales Corporations § 1.997-1 Special rules for...

48 CFR 401.105-1 - Publication and code arrangement.

Science.gov (United States)

2010-10-01

... 48 Federal Acquisition Regulations System 4 2010-10-01 2010-10-01 false Publication and code arrangement. 401.105-1 Section 401.105-1 Federal Acquisition Regulations System DEPARTMENT OF AGRICULTURE GENERAL AGRICULTURE ACQUISITION REGULATION SYSTEM Purpose, Authority, Issuance 401.105-1 Publication and...

Validation of ICD-9 Codes for Stable Miscarriage in the Emergency Department.

Science.gov (United States)

Quinley, Kelly E; Falck, Ailsa; Kallan, Michael J; Datner, Elizabeth M; Carr, Brendan G; Schreiber, Courtney A

2015-07-01

International Classification of Disease, Ninth Revision (ICD-9) diagnosis codes have not been validated for identifying cases of missed abortion where a pregnancy is no longer viable but the cervical os remains closed. Our goal was to assess whether ICD-9 code "632" for missed abortion has high sensitivity and positive predictive value (PPV) in identifying patients in the emergency department (ED) with cases of stable early pregnancy failure (EPF). We studied females ages 13-50 years presenting to the ED of an urban academic medical center. We approached our analysis from two perspectives, evaluating both the sensitivity and PPV of ICD-9 code "632" in identifying patients with stable EPF. All patients with chief complaints "pregnant and bleeding" or "pregnant and cramping" over a 12-month period were identified. We randomly reviewed two months of patient visits and calculated the sensitivity of ICD-9 code "632" for true cases of stable miscarriage. To establish the PPV of ICD-9 code "632" for capturing missed abortions, we identified patients whose visits from the same time period were assigned ICD-9 code "632," and identified those with actual cases of stable EPF. We reviewed 310 patient records (17.6% of 1,762 sampled). Thirteen of 31 patient records assigned ICD-9 code for missed abortion correctly identified cases of stable EPF (sensitivity=41.9%), and 140 of the 142 patients without EPF were not assigned the ICD-9 code "632"(specificity=98.6%). Of the 52 eligible patients identified by ICD-9 code "632," 39 cases met the criteria for stable EPF (PPV=75.0%). ICD-9 code "632" has low sensitivity for identifying stable EPF, but its high specificity and moderately high PPV are valuable for studying cases of stable EPF in epidemiologic studies using administrative data.

SIMMER-III code-verification. Phase 1

International Nuclear Information System (INIS)

Maschek, W.

1996-05-01

SIMMER-III is a computer code to investigate core disruptive accidents in liquid metal fast reactors but should also be used to investigate safety related problems in other types of advanced reactors. The code is developed by PNC with cooperation of the European partners FZK, CEA and AEA-T. SIMMER-III is a two-dimensional, three-velocity-field, multiphase, multicomponent, Eulerian, fluid-dynamics code coupled with a space-, time-, and energy-dependent neutron dynamics model. In order to model complex flow situations in a postulated disrupting core, mass and energy conservation equations are solved for 27 density components and 16 energy components, respectively. Three velocity fields (two liquid and one vapor) are modeled to simulate the relative motion of different fluid components. An additional static field takes into account the structures available in a reactor (pins, hexans, vessel structures, internal structures etc.). The neutronics is based on the discrete ordinate method (S N method) coupled into a quasistatic dynamic model. The code assessment and verification of the fluid dynamic/thermohydraulic parts of the code is performed in several steps in a joint effort of all partners. The results of the FZK contributions to the first assessment and verification phase is reported. (orig.) [de

PASC-1, Petten AMPX-II/SCALE-3 Code System for Reactor Neutronics Calculation

International Nuclear Information System (INIS)

Yaoqing, W.; Oppe, J.; Haas, J.B.M. de; Gruppelaar, H.; Slobben, J.

1995-01-01

1 - Description of program or function: The Petten AMPX-II/SCALE-3 Code System PASC-1 is a reactor neutronics calculation programme system consisting of well known IBM-oriented codes, that have been translated into FORTRAN-77, for calculations on a CDC-CYBER computer. Thus, the portability of these codes has been increased. In this system, some AMPX-II and SCALE-3 modules, the one-dimensional transport code ANISN and the 1 to 3-dimensional diffusion code CITATION are linked together on the CDC-CYBER/855 computer. The new cell code XSDRNPM-S and the old XSDRN code are included in the system. Starting from an AMPX fine group library up to CITATION, calculations can be performed for each individual module. Existing AMPX master interface format libraries, such as CSRL-IV, JEF-1, IRI and SCALE-45, and the old XSDRN-formatted libraries such as the COBB library can be used for the calculations. The code system contains the following modules and codes at present: AIM, AJAX, MALOCS, NITAWL-S, REVERT-I, ICE-2, CONVERT, JUAN, OCTAGN, XSDRNPM-S, XSDRN, ANISN and CITATION. The system will be extended with other SCALE modules and transport codes. 2 - Method of solution: The PASC-1 system is based on AMPX-II/SCALE-3 modules. Except for some SCALE-3 modules taken from the SCALIAS package, the original AMPX-II modules were IBM versions written in FORTRAN IV. These modules have been translated into CDC FORTRAN V. In order to test these modules and link them with some codes, some of the sample problem calculations have been performed for the whole PASC-1 system. During these calculations, some FORTRAN-77 errors were found in MALOCS, REVERT, CONVERT and some subroutines of SUBLIB (FORTRAN-77 subroutine library). These errors have been corrected. Because many corrections were made for the REVERT module, it is renamed as REVERT-I (improved version of REVERT). After these corrections, the whole system is running on a CDC-CYBER Computer (NOS-BE operating system). 3 - Restrictions on the

Implementation of GNASH and auxiliary codes on the Harwell CRAY-1

International Nuclear Information System (INIS)

Muir, D.W.

1985-07-01

The report describes a version of the preequilibrium, statistical nuclear-model code GNASH which has been implemented, along with a set of small auxiliary codes, on the CRAY-1 at AERE Harwell. GNASH provides a flexible tool for calculating cross sections, isomer ratios and emission spectra. A detailed description of the current user input is provided along with a full listing of the actual FORTRAN code, as modified for this implementation. (author)

Differentially Encoded LDPC Codes—Part II: General Case and Code Optimization

Directory of Open Access Journals (Sweden)

Jing Li (Tiffany

2008-04-01

Full Text Available This two-part series of papers studies the theory and practice of differentially encoded low-density parity-check (DE-LDPC codes, especially in the context of noncoherent detection. Part I showed that a special class of DE-LDPC codes, product accumulate codes, perform very well with both coherent and noncoherent detections. The analysis here reveals that a conventional LDPC code, however, is not fitful for differential coding and does not, in general, deliver a desirable performance when detected noncoherently. Through extrinsic information transfer (EXIT analysis and a modified “convergence-constraint” density evolution (DE method developed here, we provide a characterization of the type of LDPC degree profiles that work in harmony with differential detection (or a recursive inner code in general, and demonstrate how to optimize these LDPC codes. The convergence-constraint method provides a useful extension to the conventional “threshold-constraint” method, and can match an outer LDPC code to any given inner code with the imperfectness of the inner decoder taken into consideration.

Incorporation of punishable offences against the law on protection of the environment into the penal code

International Nuclear Information System (INIS)

Moehrenschlager, M.

1979-01-01

The government bill on the sixteenth act amending the penal legislation - act on the prevention of environmental offences - is presented, with comments being given on the most important stipulations. Paragraph 328 of the penal code deals with the unauthorized handling of nuclear fuels and represents an almost complete adoption of Paragraph 45 of the Atomic Energy law. The prohibition of the operation of nuclear facilities laid down in Paragraph 45, par. 1, No.4 of the Atomic Energy law, has been incorporated into the factural characteristics of illegal operation of facilities (Paragraph 327, par. 1,3 No. 1 of the penal code). The illegal handling of other radioactive substances has been dealt with in regard to some important cases: specifically for the case of illegal waste disposal (Paragraph 326, par. 1 No. 2, par. 2 of the penal code), and for the case of irregular transport leading to concrete danger or nuisance (Paragraph 330, par. 1 No. 4 of the penal code), as well as for the case of release of ionizing radiation having any such consequences (Paragraph 330, par. 1 No. 2, item c of the penal code). According to Paragraph 324 of the penal code, radioactive contamination of waters is a punishable offence. The rule of qualification of Paragraph 330 par. 1 of the penal code is to be applied to all these facts. (orig./HP) 891 HP/orig.- 892 MB [de

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

User's manual of the REFLA-1D/MODE4 reflood thermo-hydrodynamic analysis code

International Nuclear Information System (INIS)

Hojo, Tsuneyuki; Iguchi, Tadashi; Okubo, Tsutomu; Murao, Yoshio; Sugimoto, Jun.

1986-01-01

REFLA-1D/MODE4 code has been developed by incorporating local power effect model and fuel temperature profile effect model into REFLA-1D/MODE3 code. This code can calculate the temperature transient of local rod by considering radial power profile effect in core and simulate the thermal characteristics of the nuclear fuel rod. This manual describes the outline of incorporated models, modification of the code with incorporating models and provides application information required to utilize the code. (author)

Exome-Wide Association Study Identifies New Low-Frequency and Rare UGT1A1 Coding Variants and UGT1A6 Coding Variants Influencing Serum Bilirubin in Elderly Subjects

Science.gov (United States)

Oussalah, Abderrahim; Bosco, Paolo; Anello, Guido; Spada, Rosario; Guéant-Rodriguez, Rosa-Maria; Chery, Céline; Rouyer, Pierre; Josse, Thomas; Romano, Antonino; Elia, Maurizzio; Bronowicki, Jean-Pierre; Guéant, Jean-Louis

2015-01-01

Abstract Genome-wide association studies (GWASs) have identified loci contributing to total serum bilirubin level. However, no exome-wide approaches have been performed to address this question. Using exome-wide approach, we assessed the influence of protein-coding variants on unconjugated, conjugated, and total serum bilirubin levels in a well-characterized cohort of 773 ambulatory elderly subjects from Italy. Coding variants were replicated in 227 elderly subjects from the same area. We identified 4 missense rare (minor allele frequency, MAF bilirubin level (P = 2.34 × 10−34, P = 7.02 × 10−34, and P = 8.27 × 10−34), as well as unconjugated, and conjugated bilirubin levels. We also identified UGT1A6 variants in association with total (rs6759892, p.Ser7Ala, P = 1.98 × 10−26; rs2070959, p.Thr181Ala, P = 2.87 × 10−27; and rs1105879, p.Arg184Ser, P = 3.27 × 10−29), unconjugated, and conjugated bilirubin levels. All UGT1A1 intronic variants (rs887829, rs6742078, and rs4148325) and UGT1A6 coding variants (rs6759892, rs2070959, and rs1105879) were significantly associated with gallstone-related cholecystectomy risk. The UGT1A6 variant rs2070959 (p.Thr181Ala) was associated with the highest risk of gallstone–related cholecystectomy (OR, 4.58; 95% CI, 1.58–13.28; P = 3.21 × 10−3). Using an exome-wide approach we identified coding variants on UGT1A1 and UGT1A6 genes in association with serum bilirubin level and hyperbilirubinemia risk in elderly subjects. UGT1A1 intronic single-nucleotide polymorphisms (SNPs) (rs6742078, rs887829, rs4148324) serve as proxy markers for the low-frequency and rare UGT1A1 variants, thereby providing mechanistic explanation to the relationship between UGT1A1 intronic SNPs and the UGT1A1 enzyme activity. UGT1A1 and UGT1A6 variants might be potentially associated with gallstone-related cholecystectomy risk. PMID:26039129

Three Mile Island Unit 1 Main Steam Line Break Three-Dimensional Neutronics/Thermal-Hydraulics Analysis: Application of Different Coupled Codes

International Nuclear Information System (INIS)

D'Auria, Francesco; Moreno, Jose Luis Gago; Galassi, Giorgio Maria; Grgic, Davor; Spadoni, Antonino

2003-01-01

A comprehensive analysis of the double ended main steam line break (MSLB) accident assumed to occur in the Babcock and Wilcox Three Mile Island Unit 1 (TMI-1) has been carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the University of Pisa, Italy, in cooperation with the University of Zagreb, Croatia. The overall activity has been completed within the framework of the participation in the Organization for Economic Cooperation and Development-Committee on the Safety of Nuclear Installations-Nuclear Science Committee pressurized water reactor MSLB benchmark.Thermal-hydraulic system codes (various versions of Relap5), three-dimensional (3-D) neutronics codes (Parcs, Quabbox, and Nestle), and one subchannel code (Cobra) have been adopted for the analysis. Results from the following codes (or code versions) are assumed as reference:1. Relap5/mod3.2.2, beta version, coupled with the 3-D neutron kinetics Parcs code parallel virtual machine (PVM) coupling2. Relap5/mod3.2.2, gamma version, coupled with the 3-D neutron kinetics Quabbox code (direct coupling)3. Relap5/3D code coupled with the 3-D neutron kinetics Nestle code.The influence of PVM and of direct coupling is also discussed.Boundary and initial conditions of the system, including those relevant to the fuel status, have been supplied by Pennsylvania State University in cooperation with GPU Nuclear Corporation (the utility, owner of TMI) and the U.S. Nuclear Regulatory Commission. The comparison among the results obtained by adopting the same thermal-hydraulic nodalization and the coupled code version is discussed in this paper.The capability of the control rods to recover the accident has been demonstrated in all the cases as well as the capability of all the codes to predict the time evolution of the assigned transient. However, one stuck control rod caused some 'recriticality' or 'return to power' whose magnitude is largely affected by boundary and initial conditions

An implicit Smooth Particle Hydrodynamic code

Energy Technology Data Exchange (ETDEWEB)

Knapp, Charles E. [Univ. of New Mexico, Albuquerque, NM (United States)

2000-05-01

An implicit version of the Smooth Particle Hydrodynamic (SPH) code SPHINX has been written and is working. In conjunction with the SPHINX code the new implicit code models fluids and solids under a wide range of conditions. SPH codes are Lagrangian, meshless and use particles to model the fluids and solids. The implicit code makes use of the Krylov iterative techniques for solving large linear-systems and a Newton-Raphson method for non-linear corrections. It uses numerical derivatives to construct the Jacobian matrix. It uses sparse techniques to save on memory storage and to reduce the amount of computation. It is believed that this is the first implicit SPH code to use Newton-Krylov techniques, and is also the first implicit SPH code to model solids. A description of SPH and the techniques used in the implicit code are presented. Then, the results of a number of tests cases are discussed, which include a shock tube problem, a Rayleigh-Taylor problem, a breaking dam problem, and a single jet of gas problem. The results are shown to be in very good agreement with analytic solutions, experimental results, and the explicit SPHINX code. In the case of the single jet of gas case it has been demonstrated that the implicit code can do a problem in much shorter time than the explicit code. The problem was, however, very unphysical, but it does demonstrate the potential of the implicit code. It is a first step toward a useful implicit SPH code.

Exploring Trilingual Code-Switching: The Case of "Hokaglish"

Science.gov (United States)

Gonzales, Wilkinson Daniel Wong

2016-01-01

This paper presents findings of an initial study on a trilingual code-switching (CS) phenomenon called "Hokaglish" in Binondo, Manila, The Philippines. Beginning with descriptions of multiculturalism and multilingualism in the Philippines, the discussion eventually leads to the description and survey of the code-switching phenomenon…

SASSYS-1 computer code verification with EBR-II test data

International Nuclear Information System (INIS)

Warinner, D.K.; Dunn, F.E.

1985-01-01

The EBR-II natural circulation experiment, XX08 Test 8A, is simulated with the SASSYS-1 computer code and the results for the latter are compared with published data taken during the transient at selected points in the core. The SASSYS-1 results provide transient temperature and flow responses for all points of interest simultaneously during one run, once such basic parameters as pipe sizes, initial core flows, and elevations are specified. The SASSYS-1 simulation results for the EBR-II experiment XX08 Test 8A, conducted in March 1979, are within the published plant data uncertainties and, thereby, serve as a partial verification/validation of the SASSYS-1 code

An information theoretic approach to use high-fidelity codes to calibrate low-fidelity codes

Energy Technology Data Exchange (ETDEWEB)

Lewis, Allison, E-mail: lewis.allison10@gmail.com [Department of Mathematics, North Carolina State University, Raleigh, NC 27695 (United States); Smith, Ralph [Department of Mathematics, North Carolina State University, Raleigh, NC 27695 (United States); Williams, Brian [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Figueroa, Victor [Sandia National Laboratories, Albuquerque, NM 87185 (United States)

2016-11-01

For many simulation models, it can be prohibitively expensive or physically infeasible to obtain a complete set of experimental data to calibrate model parameters. In such cases, one can alternatively employ validated higher-fidelity codes to generate simulated data, which can be used to calibrate the lower-fidelity code. In this paper, we employ an information-theoretic framework to determine the reduction in parameter uncertainty that is obtained by evaluating the high-fidelity code at a specific set of design conditions. These conditions are chosen sequentially, based on the amount of information that they contribute to the low-fidelity model parameters. The goal is to employ Bayesian experimental design techniques to minimize the number of high-fidelity code evaluations required to accurately calibrate the low-fidelity model. We illustrate the performance of this framework using heat and diffusion examples, a 1-D kinetic neutron diffusion equation, and a particle transport model, and include initial results from the integration of the high-fidelity thermal-hydraulics code Hydra-TH with a low-fidelity exponential model for the friction correlation factor.

Langley Stability and Transition Analysis Code (LASTRAC) Version 1.2 User Manual

Science.gov (United States)

Chang, Chau-Lyan

2004-01-01

LASTRAC is a general-purposed, physics-based transition prediction code released by NASA for Laminar Flow Control studies and transition research. The design and development of the LASTRAC code is aimed at providing an engineering tool that is easy to use and yet capable of dealing with a broad range of transition related issues. It was written from scratch based on the state-of-the-art numerical methods for stability analysis and modern software technologies. At low fidelity, it allows users to perform linear stability analysis and N-factor transition correlation for a broad range of flow regimes and configurations by using either the linear stability theory or linear parabolized stability equations method. At high fidelity, users may use nonlinear PSE to track finite-amplitude disturbances until the skin friction rise. This document describes the governing equations, numerical methods, code development, detailed description of input/output parameters, and case studies for the current release of LASTRAC.

Energy information data base: report number codes

Energy Technology Data Exchange (ETDEWEB)

None

1979-09-01

Each report processed by the US DOE Technical Information Center is identified by a unique report number consisting of a code plus a sequential number. In most cases, the code identifies the originating installation. In some cases, it identifies a specific program or a type of publication. Listed in this publication are all codes that have been used by DOE in cataloging reports. This compilation consists of two parts. Part I is an alphabetical listing of report codes identified with the issuing installations that have used the codes. Part II is an alphabetical listing of installations identified with codes each has used. (RWR)

Energy information data base: report number codes

International Nuclear Information System (INIS)

1979-09-01

Each report processed by the US DOE Technical Information Center is identified by a unique report number consisting of a code plus a sequential number. In most cases, the code identifies the originating installation. In some cases, it identifies a specific program or a type of publication. Listed in this publication are all codes that have been used by DOE in cataloging reports. This compilation consists of two parts. Part I is an alphabetical listing of report codes identified with the issuing installations that have used the codes. Part II is an alphabetical listing of installations identified with codes each has used

Office of Codes and Standards resource book. Section 1, Building energy codes and standards

Energy Technology Data Exchange (ETDEWEB)

Hattrup, M.P.

1995-01-01

The US Department of Energy`s (DOE`s) Office of Codes and Standards has developed this Resource Book to provide: A discussion of DOE involvement in building codes and standards; a current and accurate set of descriptions of residential, commercial, and Federal building codes and standards; information on State contacts, State code status, State building construction unit volume, and State needs; and a list of stakeholders in the building energy codes and standards arena. The Resource Book is considered an evolving document and will be updated occasionally. Users are requested to submit additional data (e.g., more current, widely accepted, and/or documented data) and suggested changes to the address listed below. Please provide sources for all data provided.

Analysis of the AD sequence in Zion plant using the March 1.1 code

International Nuclear Information System (INIS)

Oriolo, F.; Paci, S.

1985-01-01

The analyses of the AD sequences for the Zion power plant, made at the Pisa University, in the framework of the participation in the Source Tern Working Group. After a short description of the plant and the sequence under analysis, the model used for the reference computation and the results obtained using the March 1.1 code are shown. Together with the reference computation a series of parametric tests have been also made, concerning some input code variables, in order to ascertain their influence on the transient trend. The results of these analyses are shown in Appendix

The PSACOIN level 1B exercise: A probabilistic code intercomparison involving a four compartment biosphere model

International Nuclear Information System (INIS)

Klos, R.A.; Sinclair, J.E.; Torres, C.; Mobbs, S.F.; Galson, D.A.

1991-01-01

The probabilistic Systems Assessment Code (PSAC) User Group of the OECD Nuclear Energy Agency has organised a series of code intercomparison studies of relevance to the performance assessment of underground repositories for radioactive wastes - known collectively by the name PSACOIN. The latest of these to be undertaken is designated PSACOIN Level 1b, and the case specification provides a complete assessment model of the behaviour of radionuclides following release into the biosphere. PSACOIN Level 1b differs from other biosphere oriented intercomparison exercises in that individual dose is the end point of the calculations as opposed to any other intermediate quantity. The PSACOIN Level 1b case specification describes a simple source term which is used to simulate the release of activity to the biosphere from certain types of near surface waste repository, the transport of radionuclides through the biosphere and their eventual uptake by humankind. The biosphere sub model comprises 4 compartments representing top and deep soil layers, river water and river sediment. The transport of radionuclides between the physical compartments is described by ten transfer coefficients and doses to humankind arise from the simultaneous consumption of water, fish, meat, milk, and grain as well as from dust inhalation and external γ-irradiation. The parameters of the exposure pathway sub model are chosen to be representative of an individual living in a small agrarian community. (13 refs., 3 figs., 2 tabs.)

Application of the core-concrete interaction code Wechsl to reactor case

International Nuclear Information System (INIS)

Cenerino, G.

1986-09-01

The WECHSL code, developed at Kernforschungszentrum Karlsruhe, West-Germany, is used for core melt accidents in nuclear power plants. The first calculations, considering silicate and limestone/common sand concretes of different compositions, analyze the influence of the initial mass of Zirconium in the corium and, in one case, the effect of sump water ingression on the top of the melt. Moreover, for a limestone concrete, a sensitivity study is made on the melting temperature of the concrete influencing the decomposition enthalpy. The main conclusion of that paper is that, in any case, the temperature of the melt drops rapidly from the initial temperature to a temperature level close to the solidification temperature of the metal phase in a relatively short period of time (approximately 15 minutes) and then a balance between the removed heat from the melt and heating sources inside the melt is established

Pretest aerosol code comparisons for LWR aerosol containment tests LA1 and LA2

International Nuclear Information System (INIS)

Wright, A.L.; Wilson, J.H.; Arwood, P.C.

1986-01-01

The Light-Water-Reactor (LWR) Aerosol Containment Experiments (LACE) are being performed in Richland, Washington, at the Hanford Engineering Development Laboratory (HEDL) under the leadership of an international project board and the Electric Power Research Institute. These tests have two objectives: (1) to investigate, at large scale, the inherent aerosol retention behavior in LWR containments under simulated severe accident conditions, and (2) to provide an experimental data base for validating aerosol behavior and thermal-hydraulic computer codes. Aerosol computer-code comparison activities are being coordinated at the Oak Ridge National Laboratory. For each of the six LACE tests, ''pretest'' calculations (for code-to-code comparisons) and ''posttest'' calculations (for code-to-test data comparisons) are being performed. The overall goals of the comparison effort are (1) to provide code users with experience in applying their codes to LWR accident-sequence conditions and (2) to evaluate and improve the code models

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

A MacWilliams Identity for Convolutional Codes: The General Case

OpenAIRE

Gluesing-Luerssen, Heide; Schneider, Gert

2008-01-01

A MacWilliams Identity for convolutional codes will be established. It makes use of the weight adjacency matrices of the code and its dual, based on state space realizations (the controller canonical form) of the codes in question. The MacWilliams Identity applies to various notions of duality appearing in the literature on convolutional coding theory.

Benchmark studies of the gyro-Landau-fluid code and gyro-kinetic codes on kinetic ballooning modes

Energy Technology Data Exchange (ETDEWEB)

Tang, T. F. [Dalian University of Technology, Dalian 116024 (China); Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Xu, X. Q. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Ma, C. H. [Fusion Simulation Center, School of Physics, Peking University, Beijing (China); Bass, E. M.; Candy, J. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Holland, C. [University of California San Diego, La Jolla, California 92093-0429 (United States)

2016-03-15

A Gyro-Landau-Fluid (GLF) 3 + 1 model has been recently implemented in BOUT++ framework, which contains full Finite-Larmor-Radius effects, Landau damping, and toroidal resonance [Ma et al., Phys. Plasmas 22, 055903 (2015)]. A linear global beta scan has been conducted using the JET-like circular equilibria (cbm18 series), showing that the unstable modes are kinetic ballooning modes (KBMs). In this work, we use the GYRO code, which is a gyrokinetic continuum code widely used for simulation of the plasma microturbulence, to benchmark with GLF 3 + 1 code on KBMs. To verify our code on the KBM case, we first perform the beta scan based on “Cyclone base case parameter set.” We find that the growth rate is almost the same for two codes, and the KBM mode is further destabilized as beta increases. For JET-like global circular equilibria, as the modes localize in peak pressure gradient region, a linear local beta scan using the same set of equilibria has been performed at this position for comparison. With the drift kinetic electron module in the GYRO code by including small electron-electron collision to damp electron modes, GYRO generated mode structures and parity suggest that they are kinetic ballooning modes, and the growth rate is comparable to the GLF results. However, a radial scan of the pedestal for a particular set of cbm18 equilibria, using GYRO code, shows different trends for the low-n and high-n modes. The low-n modes show that the linear growth rate peaks at peak pressure gradient position as GLF results. However, for high-n modes, the growth rate of the most unstable mode shifts outward to the bottom of pedestal and the real frequency of what was originally the KBMs in ion diamagnetic drift direction steadily approaches and crosses over to the electron diamagnetic drift direction.

Code Cactus; Code Cactus

Energy Technology Data Exchange (ETDEWEB)

Fajeau, M; Nguyen, L T; Saunier, J [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

1966-09-01

This code handles the following problems: -1) Analysis of thermal experiments on a water loop at high or low pressure; steady state or transient behavior; -2) Analysis of thermal and hydrodynamic behavior of water-cooled and moderated reactors, at either high or low pressure, with boiling permitted; fuel elements are assumed to be flat plates: - Flowrate in parallel channels coupled or not by conduction across plates, with conditions of pressure drops or flowrate, variable or not with respect to time is given; the power can be coupled to reactor kinetics calculation or supplied by the code user. The code, containing a schematic representation of safety rod behavior, is a one dimensional, multi-channel code, and has as its complement (FLID), a one-channel, two-dimensional code. (authors) [French] Ce code permet de traiter les problemes ci-dessous: 1. Depouillement d'essais thermiques sur boucle a eau, haute ou basse pression, en regime permanent ou transitoire; 2. Etudes thermiques et hydrauliques de reacteurs a eau, a plaques, a haute ou basse pression, ebullition permise: - repartition entre canaux paralleles, couples on non par conduction a travers plaques, pour des conditions de debit ou de pertes de charge imposees, variables ou non dans le temps; - la puissance peut etre couplee a la neutronique et une representation schematique des actions de securite est prevue. Ce code (Cactus) a une dimension d'espace et plusieurs canaux, a pour complement Flid qui traite l'etude d'un seul canal a deux dimensions. (auteurs)

18 CFR 410.1 - Basin regulations-Water Code and Administrative Manual-Part III Water Quality Regulations.

Science.gov (United States)

2010-04-01

... Code and Administrative Manual-Part III Water Quality Regulations. 410.1 Section 410.1 Conservation of... CODE AND ADMINISTRATIVE MANUAL-PART III WATER QUALITY REGULATIONS § 410.1 Basin regulations—Water Code and Administrative Manual—Part III Water Quality Regulations. (a) The Water Code of the Delaware River...

Improvement of level-1 PSA computer code package

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae Woon; Park, C. K.; Kim, K. Y.; Han, S. H.; Jung, W. D.; Chang, S. C.; Yang, J. E.; Sung, T. Y.; Kang, D. I.; Park, J. H.; Lee, Y. H.; Kim, S. H.; Hwang, M. J.; Choi, S. Y.

1997-07-01

This year the fifth (final) year of the phase-I of the Government-sponsored Mid- and Long-term Nuclear Power Technology Development Project. The scope of this subproject titled on `The improvement of level-1 PSA Computer Codes` is divided into two main activities : (1) improvement of level-1 PSA methodology, (2) development of applications methodology of PSA techniques to operations and maintenance of nuclear power plant. Level-1 PSA code KIRAP is converted to PC-Windows environment. For the improvement of efficiency in performing PSA, the fast cutset generation algorithm and an analytical technique for handling logical loop in fault tree modeling are developed. Using about 30 foreign generic data sources, generic component reliability database (GDB) are developed considering dependency among source data. A computer program which handles dependency among data sources are also developed based on three stage bayesian updating technique. Common cause failure (CCF) analysis methods are reviewed and CCF database are established. Impact vectors can be estimated from this CCF database. A computer code, called MPRIDP, which handles CCF database are also developed. A CCF analysis reflecting plant-specific defensive strategy against CCF event is also performed. A risk monitor computer program, called Risk Monster, are being developed for the application to the operation and maintenance of nuclear power plant. The PSA application technique is applied to review the feasibility study of on-line maintenance and to the prioritization of in-service test (IST) of motor-operated valves (MOV). Finally, the root cause analysis (RCA) and reliability-centered maintenance (RCM) technologies are adopted and applied to the improvement of reliability of emergency diesel generators (EDG) of nuclear power plant. To help RCA and RCM analyses, two software programs are developed, which are EPIS and RAM Pro. (author). 129 refs., 20 tabs., 60 figs.

Improvement of level-1 PSA computer code package

International Nuclear Information System (INIS)

Kim, Tae Woon; Park, C. K.; Kim, K. Y.; Han, S. H.; Jung, W. D.; Chang, S. C.; Yang, J. E.; Sung, T. Y.; Kang, D. I.; Park, J. H.; Lee, Y. H.; Kim, S. H.; Hwang, M. J.; Choi, S. Y.

1997-07-01

This year the fifth (final) year of the phase-I of the Government-sponsored Mid- and Long-term Nuclear Power Technology Development Project. The scope of this subproject titled on 'The improvement of level-1 PSA Computer Codes' is divided into two main activities : 1) improvement of level-1 PSA methodology, 2) development of applications methodology of PSA techniques to operations and maintenance of nuclear power plant. Level-1 PSA code KIRAP is converted to PC-Windows environment. For the improvement of efficiency in performing PSA, the fast cutset generation algorithm and an analytical technique for handling logical loop in fault tree modeling are developed. Using about 30 foreign generic data sources, generic component reliability database (GDB) are developed considering dependency among source data. A computer program which handles dependency among data sources are also developed based on three stage bayesian updating technique. Common cause failure (CCF) analysis methods are reviewed and CCF database are established. Impact vectors can be estimated from this CCF database. A computer code, called MPRIDP, which handles CCF database are also developed. A CCF analysis reflecting plant-specific defensive strategy against CCF event is also performed. A risk monitor computer program, called Risk Monster, are being developed for the application to the operation and maintenance of nuclear power plant. The PSA application technique is applied to review the feasibility study of on-line maintenance and to the prioritization of in-service test (IST) of motor-operated valves (MOV). Finally, the root cause analysis (RCA) and reliability-centered maintenance (RCM) technologies are adopted and applied to the improvement of reliability of emergency diesel generators (EDG) of nuclear power plant. To help RCA and RCM analyses, two software programs are developed, which are EPIS and RAM Pro. (author). 129 refs., 20 tabs., 60 figs

Verification of the MOTIF code version 3.0

International Nuclear Information System (INIS)

Chan, T.; Guvanasen, V.; Nakka, B.W.; Reid, J.A.K.; Scheier, N.W.; Stanchell, F.W.

1996-12-01

As part of the Canadian Nuclear Fuel Waste Management Program (CNFWMP), AECL has developed a three-dimensional finite-element code, MOTIF (Model Of Transport In Fractured/ porous media), for detailed modelling of groundwater flow, heat transport and solute transport in a fractured rock mass. The code solves the transient and steady-state equations of groundwater flow, solute (including one-species radionuclide) transport, and heat transport in variably saturated fractured/porous media. The initial development was completed in 1985 (Guvanasen 1985) and version 3.0 was completed in 1986. This version is documented in detail in Guvanasen and Chan (in preparation). This report describes a series of fourteen verification cases which has been used to test the numerical solution techniques and coding of MOTIF, as well as demonstrate some of the MOTIF analysis capabilities. For each case the MOTIF solution has been compared with a corresponding analytical or independently developed alternate numerical solution. Several of the verification cases were included in Level 1 of the International Hydrologic Code Intercomparison Project (HYDROCOIN). The MOTIF results for these cases were also described in the HYDROCOIN Secretariat's compilation and comparison of results submitted by the various project teams (Swedish Nuclear Power Inspectorate 1988). It is evident from the graphical comparisons presented that the MOTIF solutions for the fourteen verification cases are generally in excellent agreement with known analytical or numerical solutions obtained from independent sources. This series of verification studies has established the ability of the MOTIF finite-element code to accurately model the groundwater flow and solute and heat transport phenomena for which it is intended. (author). 20 refs., 14 tabs., 32 figs

Analysis of NEA-NSC PWR Uncontrolled Control Rod Withdrawal at Zero Power Benchmark Cases with NODAL3 Code

Directory of Open Access Journals (Sweden)

Tagor Malem Sembiring

2017-01-01

Full Text Available The in-house coupled neutronic and thermal-hydraulic (N/T-H code of BATAN (National Nuclear Energy Agency of Indonesia, NODAL3, based on the few-group neutron diffusion equation in 3-dimensional geometry using the polynomial nodal method, has been verified with static and transient PWR benchmark cases. This paper reports the verification of NODAL3 code in the NEA-NSC PWR uncontrolled control rods withdrawal at zero power benchmark. The objective of this paper is to determine the accuracy of NODAL3 code in solving the continuously slow and fast reactivity insertions due to single and group of control rod bank withdrawn while the power and temperature increment are limited by the Doppler coefficient. The benchmark is chosen since many organizations participated using various methods and approximations, so the calculation results of NODAL3 can be compared to other codes’ results. The calculated parameters are performed for the steady-state, transient core averaged, and transient hot pellet results. The influence of radial and axial nodes number was investigated for all cases. The results of NODAL3 code are in very good agreement with the reference solutions if the radial and axial nodes number is 2 × 2 and 2 × 18 (total axial layers, respectively.

A FACSIMILE code for calculating void swelling, version VS1

International Nuclear Information System (INIS)

Windsor, M.; Bullough, R.; Wood, M.H.

1979-11-01

VS1 is the first of a series of FACSIMILE codes that are being made available to predict the swelling of materials under irradiation at different temperatures, using chemical rate equations for the point defect losses to voids, interstitial loops, dislocation network, grain boundaries and foil surfaces. In this report the rate equations used in the program are given together with a detailed description of the code and directions for its use. (author)

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)

UNSAT-H Version 1.0: unsaturated flow code documentation and applications for the Hanford Site

International Nuclear Information System (INIS)

Fayer, M.J.; Gee, G.W.; Jones, T.L.

1986-08-01

Waste mangement practices at the Hanford Site have relied havily on near-surface burial. Predicting the future performance of any burial site in terms of the migration of buried contaminants requires a model capable of simulating water flow in the unsaturated soils above the buried waste. The model currently being developed to meet this need is UNSAT-H, which was developed at Pacific Northwest Laboratory for assessing the water dynamics of near-surface waste-disposal sites at the Hanfrod Site. The code will primarily be used to predict deep drainage (i.e., recharge) as a function of environmental conditions such as climate, soil type, and vegetation. UNSAT-H will also simulate various waste-management practices such as placing surface barriers over waste sites. UNSAT-H is a one-dimensional model that simulates the dynamics processes of infiltration, drainage, redistribution, surface evaporation, and uptake of water from soil by plants. UNSAT-H is designed to utilize two auxiliary codes. These codes are DATAINH, which is used to process the input data, and DATAOUT, which is used to process the UNSAT-H output. Operation of the code requires three separate steps. First, the problem to be simulated must be conceptualized in terms of boundary conditions, available data, and soil properties. Next, the data must be correctly formatted for input. Finally, the unput data must be processed, UNSAT-H run, and the output data processed for analysis. This report includes three examples of code use. In the first example, a benchmark test case is run in which the results of UNSAT-H simulations of infiltration are compared with an analytical solution and a numerical solution. The comparisons show excellent agreement for the specific test case, and this agreement provides vertification of the infiltration portion of the UNSAT-H code. The other two examples of code use are a simulation of a layered soil and one of plant transpiration

THYDE-B1/MOD2: a computer code for analysis of small-break loss-of-coolant accidents of boiling water reactors

International Nuclear Information System (INIS)

Nakamura, Hideo; Muramatsu, Ken; Kukita, Yutaka; Tasaka, Kanji

1988-04-01

THYDE-B1/MOD2 is a fast-running best estimate (BE) computer code to analyze thermal-hydraulic behaviors of the reactor cooling system of a boiling water reactor (BWR), mainly, during a small-break loss-of-coolant accident (SBLOCA) with a special emphasis on the behavior of pressure and mixture level in the pressure vessel. The coolant behavior is simulated with a volume-and-junction method based on assumptions of thermal equilibrium and homogeneous conditions for two-phase flow. A characteristic feature of this code is a three-region representation of the state of the coolant in a control volume, in which three regions consist of subcooled liquid, saturated mixture and saturated steam regions from the volume bottom. The regions are separated by two horizontal moving boundaries which are tracked by mass and energy balances for each region. With this three region node model, the interior of the pressure vessel can be represented by only two volumes: one for inside of the shroud and the other for outside, while other portions of the system are treated with homogeneous node model. This method, although it seems to be very simple, has been verified to be adequate for cases of BWR SBLOCAs in which the thermal-hydraulic behavior is relatively slow and gravity controlled. The code has been improved and modified from the last version of the code, THYDE-B1/MOD1, especially in the phase separation model which is used in the mixture level calculation in the three region node model. Then, a good predictability of the code has been indicated through the comparison of calculated results with various SBLOCA test data including ROSA-III of JAERI and FIST of the General Electric Co. This report presents the code modifications and input data requirements of the THYDE-B1/MOD2 code. (author)

Evaluation Codes from an Affine Veriety Code Perspective

DEFF Research Database (Denmark)

Geil, Hans Olav

2008-01-01

Evaluation codes (also called order domain codes) are traditionally introduced as generalized one-point geometric Goppa codes. In the present paper we will give a new point of view on evaluation codes by introducing them instead as particular nice examples of affine variety codes. Our study...... includes a reformulation of the usual methods to estimate the minimum distances of evaluation codes into the setting of affine variety codes. Finally we describe the connection to the theory of one-pointgeometric Goppa codes. Contents 4.1 Introduction...... . . . . . . . . . . . . . . . . . . . . . . . 171 4.9 Codes form order domains . . . . . . . . . . . . . . . . . . . . . . . . . . . . 173 4.10 One-point geometric Goppa codes . . . . . . . . . . . . . . . . . . . . . . . . 176 4.11 Bibliographical Notes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 178 References...

Simulation of transport in the ignited ITER with 1.5-D predictive code

International Nuclear Information System (INIS)

Becker, G.

1995-01-01

The confinement in the bulk and scrape-off layer plasmas of the ITER EDA and CDA is investigated with special versions of the 1.5-D BALDUR predictive transport code for the case of peaked density profiles (C υ = 1.0). The code self-consistently computes 2-D equilibria and solves 1-D transport equations with empirical transport coefficients for the ohmic, L and ELMy H mode regimes. Self-sustained steady state thermonuclear burn is demonstrated for up to 500 s. It is shown to be compatible with the strong radiation losses for divertor heat load reduction caused by the seeded impurities iron, neon and argon. The corresponding global and local energy and particle transport are presented. The required radiation corrected energy confinement times of the EDA and CDA are found to be close to 4 s. In the reference cases, the steady state helium fraction is 7%. The fractions of iron, neon and argon needed for the prescribed radiative power loss are given. It is shown that high radiative losses from the confinement zone, mainly by bremsstrahlung, cannot be avoided. The radiation profiles of iron and argon are found to be the same, with two thirds of the total radiation being emitted from closed flux surfaces. Fuel dilution due to iron and argon is small. The neon radiation is more peripheral. But neon is found to cause high fuel dilution. The combined dilution effect by helium and neon conflicts with burn control, self-sustained burn and divertor power reduction. Raising the helium fraction above 10% leads to the same difficulties owing to fuel dilution. The high helium levels of the present EDA design are thus unacceptable. The bootstrap current has only a small impact on the current profile. The sawtooth dominated region is found to cover 35% of the plasma cross-section. Local stability analysis of ideal ballooning modes shows that the plasma is everywhere well below the stability limit. 23 refs, 34 figs, 3 tabs

Computer codes for level 1 probabilistic safety assessment

International Nuclear Information System (INIS)

1990-06-01

Probabilistic Safety Assessment (PSA) entails several laborious tasks suitable for computer codes assistance. This guide identifies these tasks, presents guidelines for selecting and utilizing computer codes in the conduct of the PSA tasks and for the use of PSA results in safety management and provides information on available codes suggested or applied in performing PSA in nuclear power plants. The guidance is intended for use by nuclear power plant system engineers, safety and operating personnel, and regulators. Large efforts are made today to provide PC-based software systems and PSA processed information in a way to enable their use as a safety management tool by the nuclear power plant overall management. Guidelines on the characteristics of software needed for management to prepare a software that meets their specific needs are also provided. Most of these computer codes are also applicable for PSA of other industrial facilities. The scope of this document is limited to computer codes used for the treatment of internal events. It does not address other codes available mainly for the analysis of external events (e.g. seismic analysis) flood and fire analysis. Codes discussed in the document are those used for probabilistic rather than for phenomenological modelling. It should be also appreciated that these guidelines are not intended to lead the user to selection of one specific code. They provide simply criteria for the selection. Refs and tabs

Code requirements document: MODFLOW 2.1: A program for predicting moderator flow patterns

International Nuclear Information System (INIS)

Peterson, P.F.

1992-03-01

Sudden changes in the temperature of flowing liquids can result in transient buoyancy forces which strongly impact the flow hydrodynamics via flow stratification. These effects have been studied for the case of potential flow of stratified liquids to line sinks, but not for moderator flow in SRS reactors. Standard codes, such as TRAC and COMMIX, do not have the capability to capture the stratification effect, due to strong numerical diffusion which smears away the hot/cold fluid interface. A related problem with standard codes is the inability to track plumes injected into the liquid flow, again due to numerical diffusion. The combined effects of buoyant stratification and plume dispersion have been identified as being important in operation of the Supplementary Safety System which injects neutron-poison ink into SRS reactors to provide safe shutdown in the event of safety rod failure. The MODFLOW code discussed here provides transient moderator flow pattern information with stratification effects, and tracks the location of ink plumes in the reactor. The code, written in Fortran, is compiled for Macintosh II computers, and includes subroutines for interactive control and graphical output. Removing the graphics capabilities, the code can also be compiled on other computers. With graphics, in addition to the capability to perform safety related computations, MODFLOW also provides an easy tool for becoming familiar with flow distributions in SRS reactors

Modelling magnetic islands in the H-1NF heliac with the hint code

International Nuclear Information System (INIS)

Lloyd, S.S.; Gardner, H.J.

1999-01-01

Full text: Recent progress in the theoretical modelling of the effects of plasma pressure on the growth and change in geometry of magnetic islands in the H-1NF Heliac will be reviewed. The HINT magnetohydrodynamic equilibrium code, which has become a standard workhorse in the stellarator community for problems of this type, has been modified to incorporate an interpolation algorithm which significantly accelerates its convergence. This has enabled the critical evaluation of earlier results, and of some conventional wisdom. In many ways the treatment of magnetic islands in low shear fusion reactors, such as H-1NF, is an ideal case study in computational science - the devil is in the details and the devil is important: the existence or otherwise of island self-healing at reactor pressures could significantly affect the design of future experiments. (author)

MARS 1.3 system analysis code coupling with CONTEMPT4/MOD5/PCCS containment analysis code using dynamic link library

International Nuclear Information System (INIS)

Chung, Bub Dong; Jeong, Jae Jun; Lee, Won Jae

1998-01-01

The two independent codes, MARS 1.3 and CONTEMPT4/MOD5/PCCS, have been coupled using the method of dynamic-link-library (DLL) technique. Overall configuration of the code system is designed so that MARS will be a main driver program which use CONTEMPT as associated routines. Using Digital Visual Fortran compiler, DLL was generated from the CONTEMPT source code with the interfacing routine names and arguments. Coupling of MARS with CONTEMPT was realized by calling the DLL routines at the appropriate step in the MARS code. Verification of coupling was carried out for LBLOCA transient of a typical plant design. It was found that the DLL technique is much more convenient than the UNIX process control techniques and effective for Window operating system. Since DLL can be used by more than one application and an application program can use many DLLs simultaneously, this technique would enable the existing codes to use more broadly with linking others

MARS 1.3 system analysis code coupling with CONTEMPT4/MOD5/PCCS containment analysis code using dynamic link library

Energy Technology Data Exchange (ETDEWEB)

Chung, Bub Dong; Jeong, Jae Jun; Lee, Won Jae [KAERI, Taejon (Korea, Republic of)

1998-10-01

The two independent codes, MARS 1.3 and CONTEMPT4/MOD5/PCCS, have been coupled using the method of dynamic-link-library (DLL) technique. Overall configuration of the code system is designed so that MARS will be a main driver program which use CONTEMPT as associated routines. Using Digital Visual Fortran compiler, DLL was generated from the CONTEMPT source code with the interfacing routine names and arguments. Coupling of MARS with CONTEMPT was realized by calling the DLL routines at the appropriate step in the MARS code. Verification of coupling was carried out for LBLOCA transient of a typical plant design. It was found that the DLL technique is much more convenient than the UNIX process control techniques and effective for Window operating system. Since DLL can be used by more than one application and an application program can use many DLLs simultaneously, this technique would enable the existing codes to use more broadly with linking others.

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.

Challenges and opportunities in coding the commons: problems, procedures, and potential solutions in large-N comparative case studies

Directory of Open Access Journals (Sweden)

Elicia Ratajczyk

2016-09-01

Full Text Available On-going efforts to understand the dynamics of coupled social-ecological (or more broadly, coupled infrastructure systems and common pool resources have led to the generation of numerous datasets based on a large number of case studies. This data has facilitated the identification of important factors and fundamental principles which increase our understanding of such complex systems. However, the data at our disposal are often not easily comparable, have limited scope and scale, and are based on disparate underlying frameworks inhibiting synthesis, meta-analysis, and the validation of findings. Research efforts are further hampered when case inclusion criteria, variable definitions, coding schema, and inter-coder reliability testing are not made explicit in the presentation of research and shared among the research community. This paper first outlines challenges experienced by researchers engaged in a large-scale coding project; then highlights valuable lessons learned; and finally discusses opportunities for further research on comparative case study analysis focusing on social-ecological systems and common pool resources.

A MacWilliams Identity for Convolutional Codes : The General Case

NARCIS (Netherlands)

Gluesing-Luerssen, Heide; Schneider, Gert

A MacWilliams Identity for convolutional codes will be established. It makes use of the weight adjacency matrices of the code and its dual, based on state space realizations (the controller canonical form) of the codes in question. The MacWilliams Identity applies to various notions of duality

Sensitivity analysis of FRAPCON-1 computer code to some parameters

International Nuclear Information System (INIS)

Chia, C.T.; Silva, C.F. da.

1987-05-01

A sensibility study of the code FRAPCON-1 was done for the following inout data: number of axial nodes, number of time steps and the axial power shape. Their influence in the code response concerning to the fuel center line temperature, stored energy, internal gas pressure, clad hoop strain and gap width were analyzed. The number of axial nodes has little influence, but care must be taken in the choice of the power axial profile and the time step length. (Author) [pt

MicroHH 1.0: a computational fluid dynamics code for direct numerical simulation and large-eddy simulation of atmospheric boundary layer flows

Science.gov (United States)

van Heerwaarden, Chiel C.; van Stratum, Bart J. H.; Heus, Thijs; Gibbs, Jeremy A.; Fedorovich, Evgeni; Mellado, Juan Pedro

2017-08-01

This paper describes MicroHH 1.0, a new and open-source (www.microhh.org) computational fluid dynamics code for the simulation of turbulent flows in the atmosphere. It is primarily made for direct numerical simulation but also supports large-eddy simulation (LES). The paper covers the description of the governing equations, their numerical implementation, and the parameterizations included in the code. Furthermore, the paper presents the validation of the dynamical core in the form of convergence and conservation tests, and comparison of simulations of channel flows and slope flows against well-established test cases. The full numerical model, including the associated parameterizations for LES, has been tested for a set of cases under stable and unstable conditions, under the Boussinesq and anelastic approximations, and with dry and moist convection under stationary and time-varying boundary conditions. The paper presents performance tests showing good scaling from 256 to 32 768 processes. The graphical processing unit (GPU)-enabled version of the code can reach a speedup of more than an order of magnitude for simulations that fit in the memory of a single GPU.

Vectorization and improvement of nuclear codes. 3. DGR, STREAM V3.1, Cella, GGR

Energy Technology Data Exchange (ETDEWEB)

Nemoto, Toshiyuki; Eguchi, Norikuni; Watanabe, Hideo; Machida, Masahiko; Yokokawa, Mitsuo; Fujii, Minoru [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1995-01-01

Four nuclear codes have been vectorized and improved in order to realize the high speed performance on the VP2600 supercomputer at Computing and Information Systems Center of JAERI in the fiscal year 1993. Molecular Dynamics simulation code DGR which simulates the irradiation damage on diamond crystalline, three-dimensional non-steady compressible fluid dynamics code STREAM V3.1, two-dimensional fluid simulation code using Cell Automaton model Cella and Molecular Dynamics code GGR which simulates the irradiation damage on black carbon crystalline have been vectorized and improved, respectively. Speed up ratios by the vectorization to scalar mode on VP2600 show 2.8, 6.8-14.8, 15-16 and 1.23 times for DGR, STREAM V3.1, Cella and GGR, respectively. In this report, we present vectorization techniques, vectorization effects, evaluations of the numerical results and techniques for the improvement. (author).

Differentially Encoded LDPC Codes—Part II: General Case and Code Optimization

Directory of Open Access Journals (Sweden)

Li (Tiffany Jing

2008-01-01

Full Text Available This two-part series of papers studies the theory and practice of differentially encoded low-density parity-check (DE-LDPC codes, especially in the context of noncoherent detection. Part I showed that a special class of DE-LDPC codes, product accumulate codes, perform very well with both coherent and noncoherent detections. The analysis here reveals that a conventional LDPC code, however, is not fitful for differential coding and does not, in general, deliver a desirable performance when detected noncoherently. Through extrinsic information transfer (EXIT analysis and a modified "convergence-constraint" density evolution (DE method developed here, we provide a characterization of the type of LDPC degree profiles that work in harmony with differential detection (or a recursive inner code in general, and demonstrate how to optimize these LDPC codes. The convergence-constraint method provides a useful extension to the conventional "threshold-constraint" method, and can match an outer LDPC code to any given inner code with the imperfectness of the inner decoder taken into consideration.

Development of a multi-dimensional realistic thermal-hydraulic system analysis code, MARS 1.3 and its verification

Energy Technology Data Exchange (ETDEWEB)

Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-06-01

A multi-dimensional realistic thermal-hydraulic system analysis code, MARS version 1.3 has been developed. Main purpose of MARS 1.3 development is to have the realistic analysis capability of transient two-phase thermal-hydraulics of Pressurized Water Reactors (PWRs) especially during Large Break Loss of Coolant Accidents (LBLOCAs) where the multi-dimensional phenomena domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, three-dimensional (3D) reactor vessel analysis code, and RELAP5/MOD3.2.1.2, one-dimensional (1D) reactor system analysis code., Developmental requirements for MARS are chosen not only to best utilize the existing capability of the codes but also to have the enhanced capability in code maintenance, user accessibility, user friendliness, code portability, code readability, and code flexibility. For the maintenance of existing codes capability and the enhancement of code maintenance capability, user accessibility and user friendliness, MARS has been unified to be a single code consisting of 1D module (RELAP5) and 3D module (COBRA-TF). This is realized by implicitly integrating the system pressure matrix equations of hydrodynamic models and solving them simultaneously, by modifying the 1D/3D calculation sequence operable under a single Central Processor Unit (CPU) and by unifying the input structure and the light water property routines of both modules. In addition, the code structure of 1D module is completely restructured using the modular data structure of standard FORTRAN 90, which greatly improves the code maintenance capability, readability and portability. For the code flexibility, a dynamic memory management scheme is applied in both modules. MARS 1.3 now runs on PC/Windows and HP/UNIX platforms having a single CPU, and users have the options to select the 3D module to model the 3D thermal-hydraulics in the reactor vessel or other

Analysis on ingress of coolant event in vacuum vessel using modified TRAC-BF1 code

International Nuclear Information System (INIS)

Ajima, Toshio; Kurihara, Ryoichi; Seki, Yasushi

1999-08-01

The Transient Reactor Analysis Code (TRAC-BF1) was modified on the basis of ICE experimental results so as to analyze the Ingress of Coolant Event (ICE) in the vacuum vessel of a nuclear fusion reactor. In the previous report, the TRAC-BF1 code, which was originally developed for the safety analysis of a light water reactor, had been modified for the ICE of the fusion reactor. And the addition of the flat structural plate model to the VESSEL component and arbitrary appointment of the gravity direction had been added in the TRAC-BF1 code. This TRAC-BF1 code was further modified. The flat structural plate model of the VESSEL component was enabled to divide in multi layers having different materials, and a part of the multi layers could take a buried heater into consideration. Moreover, the TRAC-BF1 code was modified to analyze under the low-pressure condition close to vacuum within range of the steam table. This paper describes additional functions of the modified TRAC-BF1 code, analytical evaluation using ICE experimental data and the ITER model with final design report (FDR) data. (author)

Average Likelihood Methods of Classification of Code Division Multiple Access (CDMA)

Science.gov (United States)

2016-05-01

subject to code matrices that follows the structure given by (113). [⃗ yR y⃗I ] = √ Es 2L [ GR1 −GI1 GI2 GR2 ] [ QR −QI QI QR ] [⃗ bR b⃗I ] + [⃗ nR n⃗I... QR ] [⃗ b+ b⃗− ] + [⃗ n+ n⃗− ] (115) The average likelihood for type 4 CDMA (116) is a special case of type 1 CDMA with twice the code length and...AVERAGE LIKELIHOOD METHODS OF CLASSIFICATION OF CODE DIVISION MULTIPLE ACCESS (CDMA) MAY 2016 FINAL TECHNICAL REPORT APPROVED FOR PUBLIC RELEASE

List Decoding of Matrix-Product Codes from nested codes: an application to Quasi-Cyclic codes

DEFF Research Database (Denmark)

Hernando, Fernando; Høholdt, Tom; Ruano, Diego

2012-01-01

A list decoding algorithm for matrix-product codes is provided when $C_1,..., C_s$ are nested linear codes and $A$ is a non-singular by columns matrix. We estimate the probability of getting more than one codeword as output when the constituent codes are Reed-Solomon codes. We extend this list...... decoding algorithm for matrix-product codes with polynomial units, which are quasi-cyclic codes. Furthermore, it allows us to consider unique decoding for matrix-product codes with polynomial units....

Validation of full core geometry model of the NODAL3 code in the PWR transient Benchmark problems

International Nuclear Information System (INIS)

T-M Sembiring; S-Pinem; P-H Liem

2015-01-01

The coupled neutronic and thermal-hydraulic (T/H) code, NODAL3 code, has been validated in some PWR static benchmark and the NEACRP PWR transient benchmark cases. However, the NODAL3 code have not yet validated in the transient benchmark cases of a control rod assembly (CR) ejection at peripheral core using a full core geometry model, the C1 and C2 cases. By this research work, the accuracy of the NODAL3 code for one CR ejection or the unsymmetrical group of CRs ejection case can be validated. The calculations by the NODAL3 code have been carried out by the adiabatic method (AM) and the improved quasistatic method (IQS). All calculated transient parameters by the NODAL3 code were compared with the reference results by the PANTHER code. The maximum relative difference of 16 % occurs in the calculated time of power maximum parameter by using the IQS method, while the relative difference of the AM method is 4 % for C2 case. All calculation results by the NODAL3 code shows there is no systematic difference, it means the neutronic and T/H modules are adopted in the code are considered correct. Therefore, all calculation results by using the NODAL3 code are very good agreement with the reference results. (author)

MODIF-a code for completely reflected cylindrical reactors

International Nuclear Information System (INIS)

Gaafar, M.; Mechail, I.; Tadrus, S.

1981-01-01

MODIF-Code is a computer program for calculating the reflector saving, material buckling, and effective multiplication constant of completely reflected cylindrical reactors. The calculational method is based on a modified iterative algorithm which has been deduced from the general analytical solution of the two group diffusion equations. The code has been written in FORTRAN language suited for the ICL-1906 computer facility at Cairo University. The computer time required to solve a problem of actual reactor is less than 1 minute. The problem converges within five iteration steps. The accuracy in determining the effective multiplication constant lies within +-10 -5 . The code has been applied to the case of UA-RR-1 reactor, the results confirm the validity and accuracy of the calculational method

Coding and decoding for code division multiple user communication systems

Science.gov (United States)

Healy, T. J.

1985-01-01

A new algorithm is introduced which decodes code division multiple user communication signals. The algorithm makes use of the distinctive form or pattern of each signal to separate it from the composite signal created by the multiple users. Although the algorithm is presented in terms of frequency-hopped signals, the actual transmitter modulator can use any of the existing digital modulation techniques. The algorithm is applicable to error-free codes or to codes where controlled interference is permitted. It can be used when block synchronization is assumed, and in some cases when it is not. The paper also discusses briefly some of the codes which can be used in connection with the algorithm, and relates the algorithm to past studies which use other approaches to the same problem.

TRAC-BD1: transient reactor analysis code for boiling-water systems

International Nuclear Information System (INIS)

Spore, J.W.; Weaver, W.L.; Shumway, R.W.; Giles, M.M.; Phillips, R.E.; Mohr, C.M.; Singer, G.L.; Aguilar, F.; Fischer, S.R.

1981-01-01

The Boiling Water Reactor (BWR) version of the Transient Reactor Analysis Code (TRAC) is being developed at the Idaho National Engineering Laboratory (INEL) to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in BWRs. The TRAC-BD1 program provides the Loss of Coolant Accident (LOCA) analysis capability for BWRs and for many BWR related thermal hydraulic experimental facilities. This code features a three-dimensional treatment of the BWR pressure vessel; a detailed model of a BWR fuel bundle including multirod, multibundle, radiation heat transfer, leakage path modeling capability, flow-regime-dependent constitutive equation treatment, reflood tracking capability for both falling films and bottom flood quench fronts, and consistent treatment of the entire accident sequence. The BWR component models in TRAC-BD1 are described and comparisons with data presented. Application of the code to a BWR6 LOCA is also presented

Accelerating execution of the integrated TIGER series Monte Carlo radiation transport codes

Science.gov (United States)

Smith, L. M.; Hochstedler, R. D.

1997-02-01

Execution of the integrated TIGER series (ITS) of coupled electron/photon Monte Carlo radiation transport codes has been accelerated by modifying the FORTRAN source code for more efficient computation. Each member code of ITS was benchmarked and profiled with a specific test case that directed the acceleration effort toward the most computationally intensive subroutines. Techniques for accelerating these subroutines included replacing linear search algorithms with binary versions, replacing the pseudo-random number generator, reducing program memory allocation, and proofing the input files for geometrical redundancies. All techniques produced identical or statistically similar results to the original code. Final benchmark timing of the accelerated code resulted in speed-up factors of 2.00 for TIGER (the one-dimensional slab geometry code), 1.74 for CYLTRAN (the two-dimensional cylindrical geometry code), and 1.90 for ACCEPT (the arbitrary three-dimensional geometry code).

Accelerating execution of the integrated TIGER series Monte Carlo radiation transport codes

International Nuclear Information System (INIS)

Smith, L.M.; Hochstedler, R.D.

1997-01-01

Execution of the integrated TIGER series (ITS) of coupled electron/photon Monte Carlo radiation transport codes has been accelerated by modifying the FORTRAN source code for more efficient computation. Each member code of ITS was benchmarked and profiled with a specific test case that directed the acceleration effort toward the most computationally intensive subroutines. Techniques for accelerating these subroutines included replacing linear search algorithms with binary versions, replacing the pseudo-random number generator, reducing program memory allocation, and proofing the input files for geometrical redundancies. All techniques produced identical or statistically similar results to the original code. Final benchmark timing of the accelerated code resulted in speed-up factors of 2.00 for TIGER (the one-dimensional slab geometry code), 1.74 for CYLTRAN (the two-dimensional cylindrical geometry code), and 1.90 for ACCEPT (the arbitrary three-dimensional geometry code)

Application of low bitrate image coding to surveillance of electric power facilities. Part 1. Proposal of low bitrate coding for surveillance of electric power facilities and examination of facilities region extraction method; Denryoku setsubi kanshi eno tei rate fugoka hoshiki no tekiyo. 1. Setsubi kanshiyo fugoka hoshiki no teian to setsubi ryoiki chushutsuho no kento

Energy Technology Data Exchange (ETDEWEB)

Murata, H.; Ishino, R. [Central Research Institute of Electric Power Industry, Tokyo (Japan)

1996-03-01

Current status of low bitrate image coding has been investigated, and a low bitrate coding suitable for the surveillance of electric power facilities has been proposed, to extract its problems to be solved. For the conventional image coding, the waveform coding has been used by which the images are processed as signals. While, for the MPEG-4, a coding method with considering the image information has been proposed. For these coding methods, however, image information lacks details primarily, when lowering the bitrate. Accordingly, these methods can not be applied when the details in the images are important, such as in the case of surveillance of facilities. Then, the coding method has been proposed by expanding the partially detailed coding, and by separating constituent images of facilities, such as power cables and steel towers, designated by operators. It is the special feature of this method that the method can easily respond to the low bitrate and the detailed information can be conserved by using the structure extraction coding for the designated partial image which is generally processed by the low bitrate waveform coding. 29 refs., 17 figs., 1 tab.

Vectorization of LWR transient analysis code RELAP5/MOD1 and its effect

International Nuclear Information System (INIS)

Ishiguro, Misako; Harada, Hiroo; Shinozawa, Naohisa; Naraoka, Ken-itsu

1985-03-01

The RELAP5/MOD1 is a large thermal-hydraulic code to analyze LWR LOCA and non-LOCA transients. The code originally was designed for use on a CDC Cyber-176. This report documents vectorization of the RELAP5/MOD1 code conducted for the purpose of efficient use of VP-100 (peak speed 250 MFLOPS, clock period 7.5 ns) at the JAERI. The code was vectorized using the junction and volume level parallelisms in the hydrodynamic calculations, and the heat-structure and heat-mesh level in the heat conduction calculations. The vectorized version runs as much as 2.4 to 2.8 times faster than the original scalar version, while the speedup ratio is dependent on the number of spactial cells included in the problem. (author)

Reasons for Adopting or Revising a Journalism Ethics Code: The Case of Three Ethics Codes in the Netherlands

OpenAIRE

Poler Kovačič, Melita; van Putten, Anne-Marie

2011-01-01

The authors of this article approached the dilemma of whether or not a universal code of journalism ethics should be drafted based on the existence of factors prompting the need for a new ethics code in a national environment. Semi-structured interviews were performed with the key persons involved in the process of drafting or revising three ethics codes in the Netherlands from 2007 onwards: the Journalism Guideline by the Press Council, the Journalism Code by the Society of Chief-Editors and...

The computer code Eurdyn - 1 M. (Release 1) Part 2: User's Manual

International Nuclear Information System (INIS)

Donea, J.; Giuliani, S.

1979-01-01

This report is the user's manual for the computer code Eurdyn-1 M developed at the J.R.C. Ispra for use in containment and fuel subassembly analyses for fast reactor safety studies. The input data are defined and a test problem is presented to illustrate both the input and the output of results

World-Wide Outreach through International Observe the Moon Night

Science.gov (United States)

Buxner, S.; Jones, A. P.; Bleacher, L.; Shaner, A. J.; Day, B. H.; Wenger, M.; Joseph, E.; Canipe, M.

2016-12-01

International Observe the Moon Night (InOMN) is an annual worldwide public event that encourages observation, appreciation, and understanding of our Moon and its connection to NASA planetary science and exploration. Everyone on Earth is invited to join the celebration by hosting or attending an InOMN event - and uniting on one day each year to look at and learn about the Moon together. Events are hosted by a variety of institutions including astronomy clubs, observatories, schools, and universities, museums, planetaria, schools, universities, observatories, parks, private businesses and private homes. Events hosts are supported with event flyers, information sheets, Moon maps for observing, activities to use during events, presentations, certificates of participation, and evaluation materials to be used by hosts. 2016 is the seventh year of worldwide participation in InOMN which will be held on October 8th. In the last six years, over 3,000 events were registered worldwide from almost 100 different countries and almost all 50 states and the District of Columbia in the United States. Evaluation of InOMN is conducted by an external evaluation group and includes analysis of event registrations, facilitator surveys, and visitor surveys. Evaluation results demonstrate that InOMN events are successful in raising visitors' awareness of lunar science and exploration, providing audiences with information about lunar science and exploration, and inspiring visitors to want to learn more about the Moon. Additionally, preliminary analysis of social media has shown that there is a virtual network of individuals connecting about InOMN. A large fraction of events have been held by institutions for more than one year showing sustained interest in participation. During this presentation, we will present data for all seven years of InOMN including lessons learned through supporting and evaluating a worldwide event. InOMN is sponsored by NASA's Lunar Reconnaissance Orbiter, NASA

Structuring and coding in health care records: a qualitative analysis using diabetes as a case study

Directory of Open Access Journals (Sweden)

Ann R R Robertson

2015-03-01

Full Text Available Background   Globally, diabetes mellitus presents a substantial burden to individuals and healthcare systems. Structuring and/or coding of medical records underpin attempts to improve information sharing and searching, potentially bringing clinical and secondary uses benefits.Aims and objectives   We investigated if, how and why records for adults with diabetes were structured and/or coded, and explored stakeholders’ perceptions of current practice.Methods   We carried out a qualitative, theoretically-informed case study of documenting healthcare information for diabetes patients in family practice and hospital settings, using semi-structured interviews, observations, systems demonstrations and documentary data.Results   We conducted 22 interviews and four on-site observations, and reviewed 25 documents. For secondary uses – research, audit, public health and service planning – the benefits of highly structured and coded diabetes data were clearly articulated. Reported clinical benefits in terms of managing and monitoring diabetes, and perhaps encouraging patient self-management, were modest. We observed marked differences in levels of record structuring and/or coding between settings, and found little evidence that these data were being exploited to improve information sharing between them.Conclusions   Using high levels of data structuring and coding in medical records for diabetes patients has potential to be exploited more fully, and lessons might be learned from successful developments elsewhere in the UK.

The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 1: Basic Models

Science.gov (United States)

Mosunova, N. A.

2018-05-01

The article describes the basic models included in the EUCLID/V1 integrated code intended for safety analysis of liquid metal (sodium, lead, and lead-bismuth) cooled fast reactors using fuel rods with a gas gap and pellet dioxide, mixed oxide or nitride uranium-plutonium fuel under normal operation, under anticipated operational occurrences and accident conditions by carrying out interconnected thermal-hydraulic, neutronics, and thermal-mechanical calculations. Information about the Russian and foreign analogs of the EUCLID/V1 integrated code is given. Modeled objects, equation systems in differential form solved in each module of the EUCLID/V1 integrated code (the thermal-hydraulic, neutronics, fuel rod analysis module, and the burnup and decay heat calculation modules), the main calculated quantities, and also the limitations on application of the code are presented. The article also gives data on the scope of functions performed by the integrated code's thermal-hydraulic module, using which it is possible to describe both one- and twophase processes occurring in the coolant. It is shown that, owing to the availability of the fuel rod analysis module in the integrated code, it becomes possible to estimate the performance of fuel rods in different regimes of the reactor operation. It is also shown that the models implemented in the code for calculating neutron-physical processes make it possible to take into account the neutron field distribution over the fuel assembly cross section as well as other features important for the safety assessment of fast reactors.

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

Computations for Truck Sliding with TRUCK 3.1 Code

Science.gov (United States)

1989-08-01

16 REFERENCES 1. L u. \\Villiam N.. Hobbs. Norman P. and Atkinson, Michael. TRUCK 3.1-An Improrcd Digital (’oiputtr Program for Calculating the Response...for Operations and Plans ATIN: Technical Libary Director of Chemical & Nuear Operations Dpartnt of the AIW Waskbington, DC 20310 1 Cocaeder US Ay...Lawrenoe Livermore Lab. ATIN: Code 2124, Tedhnical ATTN: Tech Info Dept L-3 Reports Libary P.O. Be 808 Monterey, CA 93940 Livermore, CA 94550 AFSC

Tests of the TRAC code against known analytical solutions for stratified flow

International Nuclear Information System (INIS)

Black, P.S.; Leslie, D.C.; Hewitt, G.F.

1987-01-01

The area averaged equations for gas-liquid flow are briefly summarized and related, for the specific case of stratified flow, to the shallow water equations commonly used in hydraulics. These equations are then compared to the equations used in TRAC-PF/MOD1 and are shown to differ in their treatment of the gravity head terms. A modification of the TRAC code is therefore necessary to bring it into line with established shallow water theory. The corrected form of the code was compared with a number of specific cases, each of which throws further light on the code behavior. The following areas are discussed in the paper: (1) the dam break problem; (2) Kelvin-Helmholtz instability; (3) counter-current flow; and (4) slug flow. It is concluded that detailed comparisons of the code with known analytic solutions and with a number of the more complex phenomenological experiments can give useful insights into its behavior

Use of computer codes for system reliability analysis

International Nuclear Information System (INIS)

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

1989-01-01

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

Use of computer codes for system reliability analysis

Energy Technology Data Exchange (ETDEWEB)

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

1989-01-01

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

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

Psacoin level 1A intercomparison probabilistic system assessment code (PSAC) user group

International Nuclear Information System (INIS)

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

1990-01-01

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

Design specifications for ASME B and PV Code Section III nuclear class 1 piping

International Nuclear Information System (INIS)

Richardson, J.A.

1978-01-01

ASME B and PV Code Section III code regulations for nuclear piping requires that a comprehensive Design Specification be developed for ensuring that the design and installation of the piping meets all code requirements. The intent of this paper is to describe the code requirements, discuss the implementation of these requirements in a typical Class 1 piping design specification, and to report on recent piping failures in operating light water nuclear power plants in the US. (author)

Simulation of ROCOM Experiment using CUPID Code

Energy Technology Data Exchange (ETDEWEB)

Cho, Yun Je; Lee, Jae Ryong; Yoon, Han Young [KAERI, Daejeon (Korea, Republic of)

2016-10-15

KAERI has developed CUPID, which is a three dimensional thermal hydraulics code for the transient analysis of two-phase flows in nuclear reactor components. To validate the capability of CUPID for simulation of multi-dimensional flow mixing behavior, ROCOM (ROssenforf COolant Mixing) test was simulated. ROCOM test has been conducted in the OECD PKL2 Project to investigate in more detail the thermal hydraulic behavior inside the RPV. Thus far, many researchers used the ROCOM data to validate the CFD code capability of thermal mixing behavior. In this study, a hybrid grid was generated using SALOME software and the ROCOM simulation was performed using CUPID. In addition, the effect of turbulence model was also investigated. Test ROCOM 2.1 and 1.2 cases were simulated using the CUPID code. It was shown that CUPID had capabilities to properly simulate the thermal mixing behavior in the case where the cold water is injected asymmetrically. As the result of calculations, it was found that the mixing efficiency in the downcomer and lower plenum was varied according to the turbulence model. In particular, the calculation results showed that the low Reynolds number turbulence model resulted in good agreement with the experimental data. The further works may involve the finer grid generation and the test of other turbulence models.

Preliminary Coupling of MATRA Code for Multi-physics Analysis

International Nuclear Information System (INIS)

Kim, Seongjin; Choi, Jinyoung; Yang, Yongsik; Kwon, Hyouk; Hwang, Daehyun

2014-01-01

The boundary conditions such as the inlet temperature, mass flux, averaged heat flux, power distributions of the rods, and core geometry is given by constant values or functions of time. These conditions are separately calculated and provided by other codes, such as a neutronics or a system codes, into the MATRA code. In addition, the coupling of several codes in the different physics field is focused and embodied. In this study, multiphysics coupling methods were developed for a subchannel code (MATRA) with neutronics codes (MASTER, DeCART) and a fuel performance code (FRAPCON-3). Preliminary evaluation results for representative sample cases are presented. The MASTER and DeCART codes provide the power distribution of the rods in the core to the MATRA code. In case of the FRAPCON-3 code, the variation of the rod diameter induced by the thermal expansion is yielded and provided. The MATRA code transfers the thermal-hydraulic conditions that each code needs. Moreover, the coupling method with each code is described

Numerical Tokamak Project code comparison

International Nuclear Information System (INIS)

Waltz, R.E.; Cohen, B.I.; Beer, M.A.

1994-01-01

The Numerical Tokamak Project undertook a code comparison using a set of TFTR tokamak parameters. Local radial annulus codes of both gyrokinetic and gyrofluid types were compared for both slab and toroidal case limits assuming ion temperature gradient mode turbulence in a pure plasma with adiabatic electrons. The heat diffusivities were found to be in good internal agreement within ± 50% of the group average over five codes

1 CFR 21.14 - Deviations from standard organization of the Code of Federal Regulations.

Science.gov (United States)

2010-01-01

... 1 General Provisions 1 2010-01-01 2010-01-01 false Deviations from standard organization of the... CODIFICATION General Numbering § 21.14 Deviations from standard organization of the Code of Federal Regulations. (a) Any deviation from standard Code of Federal Regulations designations must be approved in advance...

Code ATOM for calculation of atomic characteristics

International Nuclear Information System (INIS)

Vainshtein, L.A.

1990-01-01

In applying atomic physics to problems of plasma diagnostics, it is necessary to determine some atomic characteristics, including energies and transition probabilities, for very many atoms and ions. Development of general codes for calculation of many types of atomic characteristics has been based on general but comparatively simple approximate methods. The program ATOM represents an attempt at effective use of such a general code. This report gives a brief description of the methods used, and the possibilities of and limitations to the code are discussed. Characteristics of the following processes can be calculated by ATOM: radiative transitions between discrete levels, radiative ionization and recombination, collisional excitation and ionization by electron impact, collisional excitation and ionization by point heavy particle (Born approximation only), dielectronic recombination, and autoionization. ATOM explores Born (for z=1) or Coulomb-Born (for z>1) approximations. In both cases exchange and normalization can be included. (N.K.)

SCDAP/RELAP5/MOD 3.1 code manual: Interface theory. Volume 1

International Nuclear Information System (INIS)

Coryell, E.W.

1995-06-01

The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code 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 off-site power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume describes the organization and manner of the interface between severe accident models which are resident in the SCDAP portion of the code and hydrodynamic models which are resident in the RELAP5 portion of the code. A description of the organization and structure of SCDAP/RELAP5 is presented. Additional information is provided regarding the manner in which models in one portion of the code impact other parts of the code, and models which are dependent on and derive information from other subcodes

Oligonephronia and Wolf-Hirschhorn syndrome: A further observation.

Science.gov (United States)

Gatto, Antonio; Ferrara, Pietro; Leoni, Chiara; Onesimo, Roberta; Zollino, Marcella; Emma, Francesco; Zampino, Giuseppe

2018-02-01

Wolf-Hirschhorn syndrome (WHS) is a rare chromosomal disorder caused by a partial deletion of chromosome 4 (4p16.3p16.2). We describe a case of a male 9 years old children with WHS proteinuria and hypertension. Laboratory data showed creatinine 1.05 mg/dl, GFR 65.9 ml/min/1.73 m 2 , cholesterol 280 mg/dl, triglyceride 125 mg/dl with electrolytes in the normal range. Urine collection showed protein 2.72 g/L with a urine protein/creatinine ratio (U P /U Cr ratio) of 4.2 and diuresis of 1,100 ml. Renal ultrasound showed reduced kidney dimensions with diffusely hyperechogenic cortex and poorly visualized pyramids. Renal biopsy showed oligonephronia with focal segmental glomerulosclerosis associated with initial tubulointerstitial sclerotic atrophy. The child began therapy with Angiotensin-converting enzyme inhibitors (ACE-inhibitors) to reduce proteinuria and progression of chronic kidney disease. In the literature the anomalies of number of glomeruli oligonephronia and oligomeganephronia (OMN) are described in two forms, one without any associated anomalies, sporadic, and solitary and the other with one or more anomalies. Our review of the literature shows that the pathogenesis of this anomaly is unknown but the role of chromosome 4 is very relevant. Many cases of OMN are associated with anomalies on this chromosome, in the literature cases series we observed this association in 14/48 cases (29.2%) and in 7 of these 14 cases with WHS. Our case and the review of literature demonstrate how periodic urinalysis and renal ultrasound monitoring is recommended in patients affected by WHS and the renal biopsy must be performed when there is the onset of proteinuria. © 2017 Wiley Periodicals, Inc.

Severe damage analysis of VVER 1000 following large break LOCA using Astec code

International Nuclear Information System (INIS)

Chatterjee, B.; Mukhopadhyay, D.; Lele, H.G.; Ghosh, A.K.; Kushwaha, H.S.

2007-01-01

Severe accident analysis of a reactor is an important aspect in the evaluation of source term. This in turn helps in emergency planning. An analysis has been carried out for VVER-1000 (V320) reactor following Large Break LOCA (loss of coolant accident) along with Station Blackout (SBO). Computer code ASTEC (jointly developed by IRSN, France, and GRS, Germany) is used for analyzing the transient. This integral code has been designed to be used as reference code for PSA2 studies. Severe accident analysis is carried out for an accident initiated by Large break LOCA along with SBO. Two cases have been analysed with the version ASTEC V1.2-rev1. In the first case hydro-accumulators are considered not available while the second case has been analysed with hydro accumulators. In this paper, ASTEC predictions have been studied for the in-vessel phase of the accident till vessel failure. The vessel failure was observed at 6979 s when accumulators were assumed not available. The vessel failure was quite delayed (19294 s) with operating accumulators. The hydrogen production was found to be very large (22% of total Zr inventory) in the case with accumulators compared to the case without accumulators (1.5% of total Zr inventory)

Codes and standards and other guidance cited in regulatory documents. Revision 1

International Nuclear Information System (INIS)

Ankrum, A.; Nickolaus, J.; Vinther, R.; Maguire-Moffitt, N.; Hammer, J.; Sherfey, L.; Warner, R.

1994-08-01

As part of the US Nuclear Regulatory Commission (NRC) Standard Review Plan Update and Development Program, Pacific Northwest Laboratory developed a listing of industry consensus codes and standards and other government and industry guidance referred to in regulatory documents. In addition to updating previous information, Revision 1 adds citations from the NRC Inspection Manual and the Improved Standard Technical Specifications. This listing identifies the version of the code or standard cited in the regulatory document, the regulatory document, and the current version of the code or standard. It also provides a summary characterization of the nature of the citation. This listing was developed from electronic searches of the Code of Federal Regulations and the NRC's Bulletins, Information Notices, Circulars, Generic Letters, Policy Statements, Regulatory Guides, and the Standard Review Plan (NUREG-0800)

Codes and standards and other guidance cited in regulatory documents. Revision 1

Energy Technology Data Exchange (ETDEWEB)

Ankrum, A.; Nickolaus, J.; Vinther, R.; Maguire-Moffitt, N.; Hammer, J.; Sherfey, L.; Warner, R. [Pacific Northwest Lab., Richland, WA (United States)

1994-08-01

As part of the US Nuclear Regulatory Commission (NRC) Standard Review Plan Update and Development Program, Pacific Northwest Laboratory developed a listing of industry consensus codes and standards and other government and industry guidance referred to in regulatory documents. In addition to updating previous information, Revision 1 adds citations from the NRC Inspection Manual and the Improved Standard Technical Specifications. This listing identifies the version of the code or standard cited in the regulatory document, the regulatory document, and the current version of the code or standard. It also provides a summary characterization of the nature of the citation. This listing was developed from electronic searches of the Code of Federal Regulations and the NRC`s Bulletins, Information Notices, Circulars, Generic Letters, Policy Statements, Regulatory Guides, and the Standard Review Plan (NUREG-0800).

Validation of ICD-9-CM coding algorithm for improved identification of hypoglycemia visits

Directory of Open Access Journals (Sweden)

Lieberman Rebecca M

2008-04-01

Full Text Available Abstract Background Accurate identification of hypoglycemia cases by International Classification of Diseases, Ninth Revision, Clinical Modification (ICD-9-CM codes will help to describe epidemiology, monitor trends, and propose interventions for this important complication in patients with diabetes. Prior hypoglycemia studies utilized incomplete search strategies and may be methodologically flawed. We sought to validate a new ICD-9-CM coding algorithm for accurate identification of hypoglycemia visits. Methods This was a multicenter, retrospective cohort study using a structured medical record review at three academic emergency departments from July 1, 2005 to June 30, 2006. We prospectively derived a coding algorithm to identify hypoglycemia visits using ICD-9-CM codes (250.3, 250.8, 251.0, 251.1, 251.2, 270.3, 775.0, 775.6, and 962.3. We confirmed hypoglycemia cases by chart review identified by candidate ICD-9-CM codes during the study period. The case definition for hypoglycemia was documented blood glucose 3.9 mmol/l or emergency physician charted diagnosis of hypoglycemia. We evaluated individual components and calculated the positive predictive value. Results We reviewed 636 charts identified by the candidate ICD-9-CM codes and confirmed 436 (64% cases of hypoglycemia by chart review. Diabetes with other specified manifestations (250.8, often excluded in prior hypoglycemia analyses, identified 83% of hypoglycemia visits, and unspecified hypoglycemia (251.2 identified 13% of hypoglycemia visits. The absence of any predetermined co-diagnosis codes improved the positive predictive value of code 250.8 from 62% to 92%, while excluding only 10 (2% true hypoglycemia visits. Although prior analyses included only the first-listed ICD-9 code, more than one-quarter of identified hypoglycemia visits were outside this primary diagnosis field. Overall, the proposed algorithm had 89% positive predictive value (95% confidence interval, 86–92 for

SACRD: a data base for fast reactor safety computer codes, contents and glossary of Version 1 of the system

International Nuclear Information System (INIS)

Greene, N.M.; Forsberg, V.M.; Raiford, G.B.; Arwood, J.W.; Flanagan, G.F.

1979-01-01

SACRD is a data base of material properties and other handbook data needed in computer codes used for fast reactor safety studies. This document lists the contents of Version 1 and also serves as a glossary of terminology used in the data base. Data are available in the thermodynamics, heat transfer, fluid mechanics, structural mechanics, aerosol transport, meteorology, neutronics and dosimetry areas. Tabular, graphical and parameterized data are provided in many cases

Non-coding RNAs and heme oxygenase-1 in vaccinia virus infection

International Nuclear Information System (INIS)

Meseda, Clement A.; Srinivasan, Kumar; Wise, Jasen; Catalano, Jennifer; Yamada, Kenneth M.; Dhawan, Subhash

2014-01-01

Highlights: • Heme oxygenase-1 (HO-1) induction inhibited vaccinia virus infection of macrophages. • Reduced infectivity inversely correlated with increased expression of non-coding RNAs. • The regulation of HO-1 and ncRNAs suggests a novel host defense response against vaccinia virus infection. - Abstract: Small nuclear RNAs (snRNAs) are <200 nucleotide non-coding uridylate-rich RNAs. Although the functions of many snRNAs remain undetermined, a population of snRNAs is produced during the early phase of infection of cells by vaccinia virus. In the present study, we demonstrate a direct correlation between expression of the cytoprotective enzyme heme oxygenase-1 (HO-1), suppression of selective snRNA expression, and inhibition of vaccinia virus infection of macrophages. Hemin induced HO-1 expression, completely reversed virus-induced host snRNA expression, and suppressed vaccinia virus infection. This involvement of specific virus-induced snRNAs and associated gene clusters suggests a novel HO-1-dependent host-defense pathway in poxvirus infection

Validation of thermohydraulic codes by comparison of experimental results with computer simulations

International Nuclear Information System (INIS)

Madeira, A.A.; Galetti, M.R.S.; Pontedeiro, A.C.

1989-01-01

The results obtained by simulation of three cases from CANON depressurization experience, using the TRAC-PF1 computer code, version 7.6, implanted in the VAX-11/750 computer of Brazilian CNEN, are presented. The CANON experience was chosen as first standard problem in thermo-hydraulic to be discussed at ENFIR for comparing results from different computer codes with results obtained experimentally. The ability of TRAC-PF1 code to prevent the depressurization phase of a loss of primary collant accident in pressurized water reactors is evaluated. (M.C.K.) [pt

Nanoparticle based bio-bar code technology for trace analysis of aflatoxin B1 in Chinese herbs

Directory of Open Access Journals (Sweden)

Yu-yan Yu

2018-04-01

Full Text Available A novel and sensitive assay for aflatoxin B1 (AFB1 detection has been developed by using bio-bar code assay (BCA. The method that relies on polyclonal antibodies encoded with DNA modified gold nanoparticle (NP and monoclonal antibodies modified magnetic microparticle (MMP, and subsequent detection of amplified target in the form of bio-bar code using a fluorescent quantitative polymerase chain reaction (FQ-PCR detection method. First, NP probes encoded with DNA that was unique to AFB1, MMP probes with monoclonal antibodies that bind AFB1 specifically were prepared. Then, the MMP-AFB1-NP sandwich compounds were acquired, dehybridization of the oligonucleotides on the nanoparticle surface allows the determination of the presence of AFB1 by identifying the oligonucleotide sequence released from the NP through FQ-PCR detection. The bio-bar code techniques system for detecting AFB1 was established, and the sensitivity limit was about 10−8 ng/mL, comparable ELISA assays for detecting the same target, it showed that we can detect AFB1 at low attomolar levels with the bio-bar-code amplification approach. This is also the first demonstration of a bio-bar code type assay for the detection of AFB1 in Chinese herbs. Keywords: Aflatoxin B1, Bio-bar code assay, Chinese herbs, Magnetic microparticle probes, Nanoparticle probes

Simulations of corrosion product transfer with the OSCAR V1.2 code

International Nuclear Information System (INIS)

Dacquait, F.; Francescatto, J.; Broutin, F.; Genin, J.B.; Benier, G.; Girard, M.; You, D.; Ranchoux, G.; Bonnefon, J.; Bachet, M.; Riot, G.

2012-09-01

Activated Corrosion Products (ACPs) generate a radiation field in PWRs, which is the major contributor to the dose absorbed by nuclear power plant staff working during shutdown operations and maintenance. Therefore, a thorough understanding of the mechanisms that control the corrosion product transfer is of the highest importance. Since the 1970's, the R and D strategy in France has been based on experiments in test loops representative of PWR conditions, on in-situ gamma spectrometry measurements of the PWR primary system contamination and on simulation code development. The simulation of corrosion product transfers in PWR primary circuits is a major challenge since it involves many physical and chemical phenomena including: corrosion, dissolution, precipitation, erosion, deposition, convection, activation... In addition to the intrinsic difficulty of multi-physics modelling, the primary systems present severe operating conditions (300 deg. C, 150 bar, neutron flux, fluid velocity up to 15 m.s -1 and very low corrosion product concentrations). The purpose of the OSCAR code, developed by the CEA in cooperation with EDF and AREVA NP, is to predict the PWR primary system contamination by corrosion and fission products. The OSCAR code is considered to be not only a tool for numerical simulations and predictions (operational practices improvements and new-built PWRs design) but also one that might combine and organise all new knowledge useful to progress on contamination. The OSCAR code for Products of Corrosion, OSCAR PC, allows researchers to analyse the corrosion product behaviour and to calculate the ACP volume and surface activities of the primary and auxiliary systems. In the new version, OSCAR PC V1.2, the corrosion product transfer in the particulate form is enhanced and a new feature is the possibility to simulate cold shutdowns. In order to validate this version, the contamination transfer has been simulated in 5 French PWRs with different operating and

TRAC-PF1 code verification with data from the OTIS test facility

International Nuclear Information System (INIS)

Childerson, M.T.; Fujita, R.K.

1985-01-01

A computer code (TRAC-PF1/MOD1) developed for predicting transient thermal and hydraulic integral nuclear steam supply system (NSSS) response was benchmarked. Post-small break loss-of-coolant accident (LOCA) data from a scaled, experimental facility, designated the One-Through Integral System (OTIS), were obtained for the Babcock and Wilcox NSSS and compared to TRAC predictions. The OTIS tests provided a challenging small break LOCA data set for TRAC verification. The major phases of a small break LOCA observed in the OTIS tests included pressurizer draining and loop saturation, intermittent reactor coolant system circulation, boiler-condenser mode, and the initial stages of refill. The TRAC code was successful in predicting OTIS loop conditions (system pressures and temperatures) after modification of the steam generator model. In particular, the code predicted both pool and auxiliary-feedwater initiated boiler-condenser mode heat transfer

Phase 1 Validation Testing and Simulation for the WEC-Sim Open Source Code

Science.gov (United States)

Ruehl, K.; Michelen, C.; Gunawan, B.; Bosma, B.; Simmons, A.; Lomonaco, P.

2015-12-01

WEC-Sim is an open source code to model wave energy converters performance in operational waves, developed by Sandia and NREL and funded by the US DOE. The code is a time-domain modeling tool developed in MATLAB/SIMULINK using the multibody dynamics solver SimMechanics, and solves the WEC's governing equations of motion using the Cummins time-domain impulse response formulation in 6 degrees of freedom. The WEC-Sim code has undergone verification through code-to-code comparisons; however validation of the code has been limited to publicly available experimental data sets. While these data sets provide preliminary code validation, the experimental tests were not explicitly designed for code validation, and as a result are limited in their ability to validate the full functionality of the WEC-Sim code. Therefore, dedicated physical model tests for WEC-Sim validation have been performed. This presentation provides an overview of the WEC-Sim validation experimental wave tank tests performed at the Oregon State University's Directional Wave Basin at Hinsdale Wave Research Laboratory. Phase 1 of experimental testing was focused on device characterization and completed in Fall 2015. Phase 2 is focused on WEC performance and scheduled for Winter 2015/2016. These experimental tests were designed explicitly to validate the performance of WEC-Sim code, and its new feature additions. Upon completion, the WEC-Sim validation data set will be made publicly available to the wave energy community. For the physical model test, a controllable model of a floating wave energy converter has been designed and constructed. The instrumentation includes state-of-the-art devices to measure pressure fields, motions in 6 DOF, multi-axial load cells, torque transducers, position transducers, and encoders. The model also incorporates a fully programmable Power-Take-Off system which can be used to generate or absorb wave energy. Numerical simulations of the experiments using WEC-Sim will be

Acceptance and validation test report for HANSF code version 1.3.2

International Nuclear Information System (INIS)

PIEPHO, M.G.

2001-01-01

The HANSF code, Version 1.3.2, is a stand-along code that runs only in DOS. As a result, it runs on any Windows' platform, since each Windows(trademark) platform can create a DOS-prompt window and execute HANSF in the DOS window. The HANSF code is proprietary to Fauske and Associates, Inc., (FAI) of Burr Ridge, IL, the developers of the code. The SNF Project has a license from FAI to run the HANSF code on any computer for only work related to SNF Project. The SNF Project owns the MCO.FOR routine, which is the main routine in HANSF for CVDF applications. The HANSF code calculates physical variables such as temperature, pressure, oxidation rates due to chemical reactions of uranium metal/fuel with water or oxygen. The code is used by the Spent Nuclear Fuel (SNF) Project at Hanford; for example, the report Thermal Analysis of Cold Vacuum Drying of Spent Nuclear Fuel (HNF-SD-SNF-CN-023). The primary facilities of interest are the K-Basins, Cold Vacuum Drying Facility (CVDF), Canister Storage Building (CSB) and T Plant. The overall Summary is presented in Section 2.0, Variances in Section 3.0, Comprehensive Assessment in Section 4.0, Results in Section 5.0, Evaluation in Section 6.0, and Summary of Activities in Section 7.0

18 CFR Table 1 to Part 301 - Functionalization and Escalation Codes

Science.gov (United States)

2010-04-01

... 18 Conservation of Power and Water Resources 1 2010-04-01 2010-04-01 false Functionalization and Escalation Codes 1 Table 1 to Part 301 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY COMMISSION, DEPARTMENT OF ENERGY REGULATIONS FOR FEDERAL POWER MARKETING ADMINISTRATIONS AVERAGE SYSTEM COST...

Entanglement-assisted quantum MDS codes from negacyclic codes

Science.gov (United States)

Lu, Liangdong; Li, Ruihu; Guo, Luobin; Ma, Yuena; Liu, Yang

2018-03-01

The entanglement-assisted formalism generalizes the standard stabilizer formalism, which can transform arbitrary classical linear codes into entanglement-assisted quantum error-correcting codes (EAQECCs) by using pre-shared entanglement between the sender and the receiver. In this work, we construct six classes of q-ary entanglement-assisted quantum MDS (EAQMDS) codes based on classical negacyclic MDS codes by exploiting two or more pre-shared maximally entangled states. We show that two of these six classes q-ary EAQMDS have minimum distance more larger than q+1. Most of these q-ary EAQMDS codes are new in the sense that their parameters are not covered by the codes available in the literature.

Coding in Stroke and Other Cerebrovascular Diseases.

Science.gov (United States)

Korb, Pearce J; Jones, William

2017-02-01

Accurate coding is critical for clinical practice and research. Ongoing changes to diagnostic and billing codes require the clinician to stay abreast of coding updates. Payment for health care services, data sets for health services research, and reporting for medical quality improvement all require accurate administrative coding. This article provides an overview of coding principles for patients with strokes and other cerebrovascular diseases and includes an illustrative case as a review of coding principles in a patient with acute stroke.

Sequence Coding and Search System for licensee event reports: user's guide. Volume 1, Revision 1

International Nuclear Information System (INIS)

Greene, N.M.; Mays, G.T.; Johnson, M.P.

1985-04-01

Operating experience data from nuclear power plants are essential for safety and reliability analyses, especially analyses of trends and patterns. The licensee event reports (LERs) that are submitted to the Nuclear Regulatory Commission (NRC) by the nuclear power plant utilities contain much of this data. The NRC's Office for Analysis and Evaluation of Operational Data (AEOD) has developed, under contract with NSIC, a system for codifying the events reported in the LERs. The primary objective of the Sequence Coding and Search System (SCSS) is to reduce the descriptive text of the LERs to coded sequences that are both computer-readable and computer-searchable. This system provides a structured format for detailed coding of component, system, and unit effects as well as personnel errors. The database contains all current LERs submitted by nuclear power plant utilities for events occurring since 1981 and is updated on a continual basis. This four volume report documents and describes SCSS in detail. Volume 1 is a User's Guide for searching the SCSS database. This volume contains updated material through February 1985 of the working version of ORNL/NSIC-223, Vol. 1

The activation cross section library UKACT1 and the inventory code FISPACT

International Nuclear Information System (INIS)

Forrest, R.A.

1989-01-01

The UK activation library for fusion applications, UKACT1, supersedes the existing UKCTRIIIA library. It contains neutron induced reaction data for 8719 reactions on 625 target nuclides. The library is used by the inventory code FISPACT which is a modified version of the existing code FISPIN. A library of decay information for all the 1314 nuclides involved is also required for calculations and this is also briefly described. UKACT1 will be used for irradiation calculations and as the starting point for a new version which will contain improved data for the most important reactions. These will be identified using the sensitivity subroutine in FISPACT. 16 refs, 1 fig., 2 tabs

Building energy performance analysis by an in-house developed dynamic simulation code: An investigation for different case studies

International Nuclear Information System (INIS)

Buonomano, Annamaria; Palombo, Adolfo

2014-01-01

Highlights: • A new dynamic simulation code for building energy performance analysis is presented. • The thermal behavior of each building element is modeled by a thermal RC network. • The physical models implemented in the code are illustrated. • The code was validated by the BESTEST standard procedure. • We investigate residential buildings, offices and stores in different climates. - Abstract: A novel dynamic simulation model for the building envelope energy performance analysis is presented in this paper. This tool helps the investigation of many new building technologies to increase the system energy efficiency and it can be carried out for scientific research purposes. In addition to the yearly heating and cooling load and energy demand, the obtained output is the dynamic temperature profile of indoor air and surfaces and the dynamic profile of the thermal fluxes through the building elements. The presented simulation model is also validated through the BESTEST standard procedure. Several new case studies are developed for assessing, through the presented code, the energy performance of three different building envelopes with several different weather conditions. In particular, dwelling and commercial buildings are analysed. Light and heavyweight envelopes as well as different glazed surfaces areas have been used for every case study. With the achieved results interesting design and operating guidelines can be obtained. Such data have been also compared vs. those calculated by TRNSYS and EnergyPlus. The detected deviation of the obtained results vs. those of such standard tools are almost always lower than 10%

Code development of total sensitivity and uncertainty analysis for reactor physics calculations

International Nuclear Information System (INIS)

Wan, C.; Cao, L.; Wu, H.; Zu, T.; Shen, W.

2015-01-01

Sensitivity and uncertainty analysis are essential parts for reactor system to perform risk and policy analysis. In this study, total sensitivity and corresponding uncertainty analysis for responses of neutronics calculations have been accomplished and developed the S&U analysis code named UNICORN. The UNICORN code can consider the implicit effects of multigroup cross sections on the responses. The UNICORN code has been applied to typical pin-cell case in this paper, and can be proved correct by comparison the results with those of the TSUNAMI-1D code. (author)

Code development of total sensitivity and uncertainty analysis for reactor physics calculations

Energy Technology Data Exchange (ETDEWEB)

Wan, C.; Cao, L.; Wu, H.; Zu, T., E-mail: chenghuiwan@stu.xjtu.edu.cn, E-mail: caolz@mail.xjtu.edu.cn, E-mail: hongchun@mail.xjtu.edu.cn, E-mail: tiejun@mail.xjtu.edu.cn [Xi' an Jiaotong Univ., School of Nuclear Science and Technology, Xi' an (China); Shen, W., E-mail: Wei.Shen@cnsc-ccsn.gc.ca [Xi' an Jiaotong Univ., School of Nuclear Science and Technology, Xi' an (China); Canadian Nuclear Safety Commission, Ottawa, ON (Canada)

2015-07-01

Sensitivity and uncertainty analysis are essential parts for reactor system to perform risk and policy analysis. In this study, total sensitivity and corresponding uncertainty analysis for responses of neutronics calculations have been accomplished and developed the S&U analysis code named UNICORN. The UNICORN code can consider the implicit effects of multigroup cross sections on the responses. The UNICORN code has been applied to typical pin-cell case in this paper, and can be proved correct by comparison the results with those of the TSUNAMI-1D code. (author)

REFLA-1D/MODE3: a computer code for reflood thermo-hydrodynamic analysis during PWR-LOCA

International Nuclear Information System (INIS)

Murao, Yoshio; Okubo, Tsutomu; Sugimoto, Jun; Iguchi, Tadashi; Sudoh, Takashi.

1985-02-01

This manual describes the REFLA-1D/MODE3 reflood system analysis code. This code can solve the core thermo-hydrodynamics under forced flooding conditions and gravity feed conditions in a system similar to FLECHT-SET Phase A. This manual describes the REFLA-1D/MODE3 models and provides application information required to utilize the code. (author)

Consistency and accuracy of diagnostic cancer codes generated by automated registration: comparison with manual registration

Directory of Open Access Journals (Sweden)

Codazzi Tiziana

2006-09-01

Full Text Available Abstract Background Automated procedures are increasingly used in cancer registration, and it is important that the data produced are systematically checked for consistency and accuracy. We evaluated an automated procedure for cancer registration adopted by the Lombardy Cancer Registry in 1997, comparing automatically-generated diagnostic codes with those produced manually over one year (1997. Methods The automatically generated cancer cases were produced by Open Registry algorithms. For manual registration, trained staff consulted clinical records, pathology reports and death certificates. The social security code, present and checked in both databases in all cases, was used to match the files in the automatic and manual databases. The cancer cases generated by the two methods were compared by manual revision. Results The automated procedure generated 5027 cases: 2959 (59% were accepted automatically and 2068 (41% were flagged for manual checking. Among the cases accepted automatically, discrepancies in data items (surname, first name, sex and date of birth constituted 8.5% of cases, and discrepancies in the first three digits of the ICD-9 code constituted 1.6%. Among flagged cases, cancers of female genital tract, hematopoietic system, metastatic and ill-defined sites, and oropharynx predominated. The usual reasons were use of specific vs. generic codes, presence of multiple primaries, and use of extranodal vs. nodal codes for lymphomas. The percentage of automatically accepted cases ranged from 83% for breast and thyroid cancers to 13% for metastatic and ill-defined cancer sites. Conclusion Since 59% of cases were accepted automatically and contained relatively few, mostly trivial discrepancies, the automatic procedure is efficient for routine case generation effectively cutting the workload required for routine case checking by this amount. Among cases not accepted automatically, discrepancies were mainly due to variations in coding practice.

Lattice Index Coding

OpenAIRE

Natarajan, Lakshmi; Hong, Yi; Viterbo, Emanuele

2014-01-01

The index coding problem involves a sender with K messages to be transmitted across a broadcast channel, and a set of receivers each of which demands a subset of the K messages while having prior knowledge of a different subset as side information. We consider the specific case of noisy index coding where the broadcast channel is Gaussian and every receiver demands all the messages from the source. Instances of this communication problem arise in wireless relay networks, sensor networks, and ...

The Cost of Enforcing Building Energy Codes: Phase 1

Energy Technology Data Exchange (ETDEWEB)

Williams, Alison [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Vine, Ed [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Price, Sarah [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Sturges, Andrew [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Rosenquist, Greg [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

2013-04-01

The purpose of this literature review is to summarize key findings regarding the costs associated with enforcing building energy code compliance—primarily focusing on costs borne by local government. The review takes into consideration over 150 documents that discuss, to some extent, code enforcement. This review emphasizes those documents that specifically focus on costs associated with energy code enforcement. Given the low rates of building energy code compliance that have been reported in existing studies, as well as the many barriers to both energy code compliance and enforcement, this study seeks to identify the costs of initiatives to improve compliance and enforcement. Costs are reported primarily as presented in the original source. Some costs are given on a per home or per building basis, and others are provided for jurisdictions of a certain size. This literature review gives an overview of state-based compliance rates, barriers to code enforcement, and U.S. Department of Energy (DOE) and key stakeholder involvement in improving compliance with building energy codes. In addition, the processes and costs associated with compliance and enforcement of building energy codes are presented. The second phase of this study, which will be presented in a different report, will consist of surveying 34 experts in the building industry at the national and state or local levels in order to obtain additional cost information, building on the findings from the first phase, as well as recommendations for where to most effectively spend money on compliance and enforcement.

Separate Turbo Code and Single Turbo Code Adaptive OFDM Transmissions

Directory of Open Access Journals (Sweden)

Lei Ye

2009-01-01

Full Text Available This paper discusses the application of adaptive modulation and adaptive rate turbo coding to orthogonal frequency-division multiplexing (OFDM, to increase throughput on the time and frequency selective channel. The adaptive turbo code scheme is based on a subband adaptive method, and compares two adaptive systems: a conventional approach where a separate turbo code is used for each subband, and a single turbo code adaptive system which uses a single turbo code over all subbands. Five modulation schemes (BPSK, QPSK, 8AMPM, 16QAM, and 64QAM are employed and turbo code rates considered are 1/2 and 1/3. The performances of both systems with high (10−2 and low (10−4 BER targets are compared. Simulation results for throughput and BER show that the single turbo code adaptive system provides a significant improvement.

Under-coding of secondary conditions in coded hospital health data: Impact of co-existing conditions, death status and number of codes in a record.

Science.gov (United States)

Peng, Mingkai; Southern, Danielle A; Williamson, Tyler; Quan, Hude

2017-12-01

This study examined the coding validity of hypertension, diabetes, obesity and depression related to the presence of their co-existing conditions, death status and the number of diagnosis codes in hospital discharge abstract database. We randomly selected 4007 discharge abstract database records from four teaching hospitals in Alberta, Canada and reviewed their charts to extract 31 conditions listed in Charlson and Elixhauser comorbidity indices. Conditions associated with the four study conditions were identified through multivariable logistic regression. Coding validity (i.e. sensitivity, positive predictive value) of the four conditions was related to the presence of their associated conditions. Sensitivity increased with increasing number of diagnosis code. Impact of death on coding validity is minimal. Coding validity of conditions is closely related to its clinical importance and complexity of patients' case mix. We recommend mandatory coding of certain secondary diagnosis to meet the need of health research based on administrative health data.

The computer code EURDYN - 1 M (release 1) for transient dynamic fluid-structure interaction. Pt.1: governing equations and finite element modelling

International Nuclear Information System (INIS)

Donea, J.; Fasoli-Stella, P.; Giuliani, S.; Halleux, J.P.; Jones, A.V.

1980-01-01

This report describes the governing equations and the finite element modelling used in the computer code EURDYN - 1 M. The code is a non-linear transient dynamic program for the analysis of coupled fluid-structure systems; It is designed for safety studies on LMFBR components (primary containment and fuel subassemblies)

Calculations of IAEA-CRP-6 Benchmark Case 1 through 7 for a TRISO-Coated Fuel Particle

International Nuclear Information System (INIS)

Kim, Young Min; Lee, Y. W.; Chang, J. H.

2005-01-01

IAEA-CRP-6 is a coordinated research program of IAEA on Advances in HTGR fuel technology. The CRP examines aspects of HTGR fuel technology, ranging from design and fabrication to characterization, irradiation testing, performance modeling, as well as licensing and quality control issues. The benchmark section of the program treats simple analytical cases, pyrocarbon layer behavior, single TRISO-coated fuel particle behavior, and benchmark calculations of some irradiation experiments performed and planned. There are totally seventeen benchmark cases in the program. Member countries are participating in the benchmark calculations of the CRP with their own developed fuel performance analysis computer codes. Korea is also taking part in the benchmark calculations using a fuel performance analysis code, COPA (COated PArticle), which is being developed in Korea Atomic Energy Research Institute. The study shows the calculational results of IAEACRP- 6 benchmark cases 1 through 7 which describe the structural behaviors for a single fuel particle

CONTAIN LMR/1B-Mod.1, A computer code for containment analysis of accidents in liquid-metal-cooled nuclear reactors

International Nuclear Information System (INIS)

Murata, K.K.; Carroll, D.E.; Bergeron, K.D.; Valdez, G.D.

1993-01-01

The CONTAIN computer code is a best-estimate, integrated analysis tool for predicting the physical, chemical, and radiological conditions inside a nuclear reactor containment building following the release of core material from the primary system. CONTAIN is supported primarily by the U. S. Nuclear Regulatory Commission (USNRC), and the official code versions produced with this support are intended primarily for the analysis of light water reactors (LWR). The present manual describes CONTAIN LMR/1B-Mod. 1, a code version designed for the analysis of reactors with liquid metal coolant. It is a variant of the official CONTAIN 1.11 LWR code version. Some of the features of CONTAIN-LMR for treating the behavior of liquid metal coolant are in fact present in the LWR code versions but are discussed here rather than in the User's Manual for the LWR versions. These features include models for sodium pool and spray fires. In addition to these models, new or substantially improved models have been installed in CONTAIN-LMR. The latter include models for treating two condensables (sodium and water) simultaneously, sodium atmosphere and pool chemistry, sodium condensation on aerosols, heat transfer from core-debris beds and to sodium pools, and sodium-concrete interactions. A detailed description of each of the above models is given, along with the code input requirements

RIA Fuel Codes Benchmark - Volume 1

International Nuclear Information System (INIS)

Marchand, Olivier; Georgenthum, Vincent; Petit, Marc; Udagawa, Yutaka; Nagase, Fumihisa; Sugiyama, Tomoyuki; Arffman, Asko; Cherubini, Marco; Dostal, Martin; Klouzal, Jan; Geelhood, Kenneth; Gorzel, Andreas; Holt, Lars; Jernkvist, Lars Olof; Khvostov, Grigori; Maertens, Dietmar; Spykman, Gerold; Nakajima, Tetsuo; Nechaeva, Olga; Panka, Istvan; Rey Gayo, Jose M.; Sagrado Garcia, Inmaculada C.; Shin, An-Dong; Sonnenburg, Heinz Guenther; Umidova, Zeynab; Zhang, Jinzhao; Voglewede, John

2013-01-01

Reactivity-initiated accident (RIA) fuel rod codes have been developed for a significant period of time and they all have shown their ability to reproduce some experimental results with a certain degree of adequacy. However, they sometimes rely on different specific modelling assumptions the influence of which on the final results of the calculations is difficult to evaluate. The NEA Working Group on Fuel Safety (WGFS) is tasked with advancing the understanding of fuel safety issues by assessing the technical basis for current safety criteria and their applicability to high burnup and to new fuel designs and materials. The group aims at facilitating international convergence in this area, including the review of experimental approaches as well as the interpretation and use of experimental data relevant for safety. As a contribution to this task, WGFS conducted a RIA code benchmark based on RIA tests performed in the Nuclear Safety Research Reactor in Tokai, Japan and tests performed or planned in CABRI reactor in Cadarache, France. Emphasis was on assessment of different modelling options for RIA fuel rod codes in terms of reproducing experimental results as well as extrapolating to typical reactor conditions. This report provides a summary of the results of this task. (authors)

1-D hybrid code for FRM dynamics

International Nuclear Information System (INIS)

Stark, R.A.; Miley, G.H.

1985-01-01

A 1-D radial hybrid code has been written to study the start-up of the FRM via neutral-beam injection. This code, named FROST (Field Reversed One-dimensional STart-up), models the plasma as azimuthal symmetric with no axial dependence. A multi-group method in energy and canonical angular momentum describes the large-orbit ions from the beam. This method is designed to be more efficient than those employing particle tracking, since the characteristic timescale of the simulation is the ion slowing down time, rather than the much shorter cyclotron period. A time-differentiated Grad-Shafranov equation couples the ion current to massless fluid equations describing electrons and low energy ions. Flux coordinates are used in this fluid model, in preference to an Eulerian framework, so that coupling of plasma at the two different radii of a closed flux surface may be treated with ease. Since a fluid treatment for electrons is invalid near a field null, a separate model for the electron current has been included for this region, a unique feature. Results of simulation of injection into a 2XIIB-like plasma are discussed. Electron currents are found to retard, but not prevent reversal of the magnetic field at the plasma center

POPCYCLE: a computer code for calculating nuclear and fossil plant levelized life-cycle power costs

International Nuclear Information System (INIS)

Hardie, R.W.

1982-02-01

POPCYCLE, a computer code designed to calculate levelized life-cycle power costs for nuclear and fossil electrical generating plants is described. Included are (1) derivations of the equations and a discussion of the methodology used by POPCYCLE, (2) a description of the input required by the code, (3) a listing of the input for a sample case, and (4) the output for a sample case

Evaluation of Computational Fluids Dynamics (CFD) code Open FOAM in the study of the pressurized thermal stress of PWR reactors. Comparison with the commercial code Ansys-CFX

International Nuclear Information System (INIS)

Martinez, M.; Barrachina, T.; Miro, R.; Verdu Martin, G.; Chiva, S.

2012-01-01

In this work is proposed to evaluate the potential of the OpenFOAM code for the simulation of typical fluid flows in reactors PWR, in particular for the study of pressurized thermal stress. Test T1-1 has been simulated , within the OECD ROSA project, with the objective of evaluating the performance of the code OpenFOAM and models of turbulence that has implemented to capture the effect of the thrust forces in the case study.

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

Refactoring test code

NARCIS (Netherlands)

A. van Deursen (Arie); L.M.F. Moonen (Leon); A. van den Bergh; G. Kok

2001-01-01

textabstractTwo key aspects of extreme programming (XP) are unit testing and merciless refactoring. Given the fact that the ideal test code / production code ratio approaches 1:1, it is not surprising that unit tests are being refactored. We found that refactoring test code is different from

Recent Updates to the MELCOR 1.8.2 Code for ITER Applications

Energy Technology Data Exchange (ETDEWEB)

Merrill, Brad J

2007-05-01

This report documents recent changes made to the MELCOR 1.8.2 computer code for application to the International Thermonuclear Experimental Reactor (ITER), as required by ITER Task Agreement ITA 81-18. There are four areas of change documented by this report. The first area is the addition to this code of a model for transporting HTO. The second area is the updating of the material oxidation correlations to match those specified in the ITER Safety Analysis Data List (SADL). The third area replaces a modification to an aerosol tranpsort subroutine that specified the nominal aerosol density internally with one that now allows the user to specify this density through user input. The fourth area corrected an error that existed in an air condensation subroutine of previous versions of this modified MELCOR code. The appendices of this report contain FORTRAN listings of the coding for these modifications.

Recent Updates to the MELCOR 1.8.2 Code for ITER Applications

Energy Technology Data Exchange (ETDEWEB)

Merrill, Brad J

2007-04-01

This report documents recent changes made to the MELCOR 1.8.2 computer code for application to the International Thermonuclear Experimental Reactor (ITER), as required by ITER Task Agreement ITA 81-18. There are four areas of change documented by this report. The first area is the addition to this code of a model for transporting HTO. The second area is the updating of the material oxidation correlations to match those specified in the ITER Safety Analysis Data List (SADL). The third area replaces a modification to an aerosol tranpsort subroutine that specified the nominal aerosol density internally with one that now allows the user to specify this density through user input. The fourth area corrected an error that existed in an air condensation subroutine of previous versions of this modified MELCOR code. The appendices of this report contain FORTRAN listings of the coding for these modifications.

Structuring and coding in health care records: a qualitative analysis using diabetes as a case study.

Science.gov (United States)

Robertson, Ann R R; Fernando, Bernard; Morrison, Zoe; Kalra, Dipak; Sheikh, Aziz

2015-03-27

Globally, diabetes mellitus presents a substantial and increasing burden to individuals, health care systems and society. Structuring and coding of information in the electronic health record underpin attempts to improve sharing and searching for information. Digital records for those with long-term conditions are expected to bring direct and secondary uses benefits, and potentially to support patient self-management. We sought to investigate if how and why records for adults with diabetes were structured and coded and to explore a range of UK stakeholders' perceptions of current practice in the National Health Service. We carried out a qualitative, theoretically informed case study of documenting health care information for diabetes in family practice and hospital settings in England, using semi-structured interviews, observations, systems demonstrations and documentary data. We conducted 22 interviews and four on-site observations. With respect to secondary uses - research, audit, public health and service planning - interviewees clearly articulated the benefits of highly structured and coded diabetes data and it was believed that benefits would expand through linkage to other datasets. Direct, more marginal, clinical benefits in terms of managing and monitoring diabetes and perhaps encouraging patient self-management were also reported. We observed marked differences in levels of record structuring and/or coding between family practices, where it was high, and the hospital. We found little evidence that structured and coded data were being exploited to improve information sharing between care settings. Using high levels of data structuring and coding in records for diabetes patients has the potential to be exploited more fully, and lessons might be learned from successful developments elsewhere in the UK. A first step would be for hospitals to attain levels of health information technology infrastructure and systems use commensurate with family practices.

Performance Study of Monte Carlo Codes on Xeon Phi Coprocessors — Testing MCNP 6.1 and Profiling ARCHER Geometry Module on the FS7ONNi Problem

Science.gov (United States)

Liu, Tianyu; Wolfe, Noah; Lin, Hui; Zieb, Kris; Ji, Wei; Caracappa, Peter; Carothers, Christopher; Xu, X. George

2017-09-01

This paper contains two parts revolving around Monte Carlo transport simulation on Intel Many Integrated Core coprocessors (MIC, also known as Xeon Phi). (1) MCNP 6.1 was recompiled into multithreading (OpenMP) and multiprocessing (MPI) forms respectively without modification to the source code. The new codes were tested on a 60-core 5110P MIC. The test case was FS7ONNi, a radiation shielding problem used in MCNP's verification and validation suite. It was observed that both codes became slower on the MIC than on a 6-core X5650 CPU, by a factor of 4 for the MPI code and, abnormally, 20 for the OpenMP code, and both exhibited limited capability of strong scaling. (2) We have recently added a Constructive Solid Geometry (CSG) module to our ARCHER code to provide better support for geometry modelling in radiation shielding simulation. The functions of this module are frequently called in the particle random walk process. To identify the performance bottleneck we developed a CSG proxy application and profiled the code using the geometry data from FS7ONNi. The profiling data showed that the code was primarily memory latency bound on the MIC. This study suggests that despite low initial porting e_ort, Monte Carlo codes do not naturally lend themselves to the MIC platform — just like to the GPUs, and that the memory latency problem needs to be addressed in order to achieve decent performance gain.

Significant issues and changes for ANSI/ASME OM-1 1981, part 1, ASME OMc code-1994, and ASME OM Code-1995, Appendix I, inservice testing of pressure relief devices in light water reactor power plants

Energy Technology Data Exchange (ETDEWEB)

Seniuk, P.J.

1996-12-01

This paper identifies significant changes to the ANSI/ASME OM-1 1981, Part 1, and ASME Omc Code-1994 and ASME OM Code-1995, Appendix I, {open_quotes}Inservice Testing of Pressure Relief Devices in Light-Water Reactor Power Plants{close_quotes}. The paper describes changes to different Code editions and presents insights into the direction of the code committee and selected topics to be considered by the ASME O&M Working Group on pressure relief devices. These topics include scope issues, thermal relief valve issues, as-found and as-left set-pressure determinations, exclusions from testing, and cold setpoint bench testing. The purpose of this paper is to describe some significant issues being addressed by the O&M Working Group on Pressure Relief Devices (OM-1). The writer is currently the chair of OM-1 and the statements expressed herein represents his personal opinion.

Significant issues and changes for ANSI/ASME OM-1 1981, part 1, ASME OMc code-1994, and ASME OM Code-1995, Appendix I, inservice testing of pressure relief devices in light water reactor power plants

International Nuclear Information System (INIS)

Seniuk, P.J.

1996-01-01

This paper identifies significant changes to the ANSI/ASME OM-1 1981, Part 1, and ASME Omc Code-1994 and ASME OM Code-1995, Appendix I, open-quotes Inservice Testing of Pressure Relief Devices in Light-Water Reactor Power Plantsclose quotes. The paper describes changes to different Code editions and presents insights into the direction of the code committee and selected topics to be considered by the ASME O ampersand M Working Group on pressure relief devices. These topics include scope issues, thermal relief valve issues, as-found and as-left set-pressure determinations, exclusions from testing, and cold setpoint bench testing. The purpose of this paper is to describe some significant issues being addressed by the O ampersand M Working Group on Pressure Relief Devices (OM-1). The writer is currently the chair of OM-1 and the statements expressed herein represents his personal opinion

Entanglement-assisted quantum MDS codes constructed from negacyclic codes

Science.gov (United States)

Chen, Jianzhang; Huang, Yuanyuan; Feng, Chunhui; Chen, Riqing

2017-12-01

Recently, entanglement-assisted quantum codes have been constructed from cyclic codes by some scholars. However, how to determine the number of shared pairs required to construct entanglement-assisted quantum codes is not an easy work. In this paper, we propose a decomposition of the defining set of negacyclic codes. Based on this method, four families of entanglement-assisted quantum codes constructed in this paper satisfy the entanglement-assisted quantum Singleton bound, where the minimum distance satisfies q+1 ≤ d≤ n+2/2. Furthermore, we construct two families of entanglement-assisted quantum codes with maximal entanglement.

Building 1D resonance broadened quasilinear (RBQ) code for fast ions Alfvénic relaxations

Science.gov (United States)

Gorelenkov, Nikolai; Duarte, Vinicius; Berk, Herbert

2016-10-01

The performance of the burning plasma is limited by the confinement of superalfvenic fusion products, e.g. alpha particles, which are capable of resonating with the Alfvénic eigenmodes (AEs). The effect of AEs on fast ions is evaluated using a resonance line broadened diffusion coefficient. The interaction of fast ions and AEs is captured for cases where there are either isolated or overlapping modes. A new code RBQ1D is being built which constructs diffusion coefficients based on realistic eigenfunctions that are determined by the ideal MHD code NOVA. The wave particle interaction can be reduced to one-dimensional dynamics where for the Alfvénic modes typically the particle kinetic energy is nearly constant. Hence to a good approximation the Quasi-Linear (QL) diffusion equation only contains derivatives in the angular momentum. The diffusion equation is then one dimensional that is efficiently solved simultaneously for all particles with the equation for the evolution of the wave angular momentum. The evolution of fast ion constants of motion is governed by the QL diffusion equations which are adapted to find the ion distribution function.

Light water reactor fuel analysis code FEMAXI-V (Ver.1)

International Nuclear Information System (INIS)

Suzuki, Motoe

2000-09-01

A light water fuel analysis code FEMAXI-V is an advanced version which has been produced by integrating FEMAXI-IV(Ver.2), high burn-up fuel code EXBURN-I, and a number of functional improvements and extensions, to predict fuel rod behavior in normal and transient (not accident) conditions. The present report describes in detail the basic theories and structure, models and numerical solutions applied, improvements and extensions, and the material properties adopted in FEMAXI-V(Ver.1). FEMAXI-V deals with a single fuel rod. It predicts thermal and mechanical response of fuel rod to irradiation, including FP gas release. The thermal analysis predicts rod temperature distribution on the basis of pellet heat generation, changes in pellet thermal conductivity and gap thermal conductance, (transient) change in surface heat transfer to coolant, using radial one-dimensional geometry. The heat generation density profile of pellet can be determined by adopting the calculated results of burning analysis code. The mechanical analysis performs elastic/plastic, creep and PCMI calculations by FEM. The FP gas release model calculates diffusion of FP gas atoms and accumulation in bubbles, release and increase in internal pressure of rod. In every analysis, it is possible to allow some materials properties and empirical equations to depend on the local burnup or heat flux, which enables particularly analysis of high burnup fuel behavior and boiling transient of BWR rod. In order to facilitate effective and wide-ranging application of the code, formats and methods of input/output of the code are also described, and a sample output in an actual form is included. (author)

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

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

Application of the NJOY code for unresolved resonance treatment in the MCNP utility code

International Nuclear Information System (INIS)

Milosevic, M.; Greenspan, E.; Vujic, J. . E-mail addresses of corresponding authors: mmilos@vin.bg.ac.yu , vujic@nuc.berkeley.edu ,; Milosevic, M.; Vujic, J.)

2005-01-01

There are numerous uncertainties in the prediction of neutronic characteristics of reactor cores, particularly in the case of innovative reactor designs, arising from approximations used in the solution of the transport equation, and in nuclear data processing and cross section libraries generation. This paper describes the problems encountered in the analysis of the Encapsulated Nuclear Heat Source (ENHS) benchmark core and the new procedures and cross section libraries developed to overcome these problems. The ENHS is a new lead-bismuth or lead cooled novel reactor concept that is fuelled with metallic alloy of Pu, U and Zr, and it is designed to operate for 20 effective full power years without refuelling and with very small burnup reactivity swing. The computational tools benchmarked include: MOCUP - a coupled MCNP-4C and ORIGEN2.1 utility codes with MCNP data libraries based on the ENDF/B-VI evaluations; and KWO2 - a coupled KENO-V.a and ORIGEN2.1 code with ENDFB-V.2 based 238 group library. Calculations made for the ENHS benchmark have shown that the differences between the results obtained using different code systems and cross section libraries are significant and should be taken into account in assessing the quality of nuclear data libraries. (author)

Audit of accuracy of clinical coding in oral surgery.

Science.gov (United States)

Naran, S; Hudovsky, A; Antscherl, J; Howells, S; Nouraei, S A R

2014-10-01

We aimed to study the accuracy of clinical coding within oral surgery and to identify ways in which it can be improved. We undertook did a multidisciplinary audit of a sample of 646 day case patients who had had oral surgery procedures between 2011 and 2012. We compared the codes given with their case notes and amended any discrepancies. The accuracy of coding was assessed for primary and secondary diagnoses and procedures, and for health resource groupings (HRGs). The financial impact of coding Subjectivity, Variability and Error (SVE) was assessed by reference to national tariffs. The audit resulted in 122 (19%) changes to primary diagnoses. The codes for primary procedures changed in 224 (35%) cases; 310 (48%) morbidities and complications had been missed, and 266 (41%) secondary procedures had been missed or were incorrect. This led to at least one change of coding in 496 (77%) patients, and to the HRG changes in 348 (54%) patients. The financial impact of this was £114 in lost revenue per patient. There is a high incidence of coding errors in oral surgery because of the large number of day cases, a lack of awareness by clinicians of coding issues, and because clinical coders are not always familiar with the large number of highly specialised abbreviations used. Accuracy of coding can be improved through the use of a well-designed proforma, and standards can be maintained by the use of an ongoing data quality assurance programme. Copyright © 2014. Published by Elsevier Ltd.

Nanoparticle based bio-bar code technology for trace analysis of aflatoxin B1 in Chinese herbs.

Science.gov (United States)

Yu, Yu-Yan; Chen, Yuan-Yuan; Gao, Xuan; Liu, Yuan-Yuan; Zhang, Hong-Yan; Wang, Tong-Ying

2018-04-01

A novel and sensitive assay for aflatoxin B1 (AFB1) detection has been developed by using bio-bar code assay (BCA). The method that relies on polyclonal antibodies encoded with DNA modified gold nanoparticle (NP) and monoclonal antibodies modified magnetic microparticle (MMP), and subsequent detection of amplified target in the form of bio-bar code using a fluorescent quantitative polymerase chain reaction (FQ-PCR) detection method. First, NP probes encoded with DNA that was unique to AFB1, MMP probes with monoclonal antibodies that bind AFB1 specifically were prepared. Then, the MMP-AFB1-NP sandwich compounds were acquired, dehybridization of the oligonucleotides on the nanoparticle surface allows the determination of the presence of AFB1 by identifying the oligonucleotide sequence released from the NP through FQ-PCR detection. The bio-bar code techniques system for detecting AFB1 was established, and the sensitivity limit was about 10 -8  ng/mL, comparable ELISA assays for detecting the same target, it showed that we can detect AFB1 at low attomolar levels with the bio-bar-code amplification approach. This is also the first demonstration of a bio-bar code type assay for the detection of AFB1 in Chinese herbs. Copyright © 2017. Published by Elsevier B.V.

Monte Carlo simulation in UWB1 depletion code

International Nuclear Information System (INIS)

Lovecky, M.; Prehradny, J.; Jirickova, J.; Skoda, R.

2015-01-01

U W B 1 depletion code is being developed as a fast computational tool for the study of burnable absorbers in the University of West Bohemia in Pilsen, Czech Republic. In order to achieve higher precision, the newly developed code was extended by adding a Monte Carlo solver. Research of fuel depletion aims at development and introduction of advanced types of burnable absorbers in nuclear fuel. Burnable absorbers (BA) allow the compensation of the initial reactivity excess of nuclear fuel and result in an increase of fuel cycles lengths with higher enriched fuels. The paper describes the depletion calculations of VVER nuclear fuel doped with rare earth oxides as burnable absorber based on performed depletion calculations, rare earth oxides are divided into two equally numerous groups, suitable burnable absorbers and poisoning absorbers. According to residual poisoning and BA reactivity worth, rare earth oxides marked as suitable burnable absorbers are Nd, Sm, Eu, Gd, Dy, Ho and Er, while poisoning absorbers include Sc, La, Lu, Y, Ce, Pr and Tb. The presentation slides have been added to the article

Use of computer codes for system reliability analysis

International Nuclear Information System (INIS)

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

1988-01-01

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

Fast decoders for qudit topological codes

International Nuclear Information System (INIS)

Anwar, Hussain; Brown, Benjamin J; Campbell, Earl T; Browne, Dan E

2014-01-01

Qudit toric codes are a natural higher-dimensional generalization of the well-studied qubit toric code. However, standard methods for error correction of the qubit toric code are not applicable to them. Novel decoders are needed. In this paper we introduce two renormalization group decoders for qudit codes and analyse their error correction thresholds and efficiency. The first decoder is a generalization of a ‘hard-decisions’ decoder due to Bravyi and Haah (arXiv:1112.3252). We modify this decoder to overcome a percolation effect which limits its threshold performance for many-level quantum systems. The second decoder is a generalization of a ‘soft-decisions’ decoder due to Poulin and Duclos-Cianci (2010 Phys. Rev. Lett. 104 050504), with a small cell size to optimize the efficiency of implementation in the high dimensional case. In each case, we estimate thresholds for the uncorrelated bit-flip error model and provide a comparative analysis of the performance of both these approaches to error correction of qudit toric codes. (paper)

ScintSim1: a new Monte Carlo simulation code for transport of optical photons in 2D arrays of scintillation detectors

International Nuclear Information System (INIS)

Mosleh-Shirazi, Mohammad Amin; Karbasi, Sareh; Zarrini-Monfared, Zinat; Zamani, Ali

2014-01-01

Two-dimensional (2D) arrays of thick segmented scintillators are of interest as X-ray detectors for both 2D and 3D image-guided radiotherapy (IGRT). Their detection process involves ionizing radiation energy deposition followed by production and transport of optical photons. Only a very limited number of optical Monte Carlo simulation models exist, which has limited the number of modeling studies that have considered both stages of the detection process. We present ScintSim1, an in-house optical Monte Carlo simulation code for 2D arrays of scintillation crystals, developed in the MATLAB programming environment. The code was rewritten and revised based on an existing program for single-element detectors, with the additional capability to model 2D arrays of elements with configurable dimensions, material, etc., The code generates and follows each optical photon history through the detector element (and, in case of cross-talk, the surrounding ones) until it reaches a configurable receptor, or is attenuated. The new model was verified by testing against relevant theoretically known behaviors or quantities and the results of a validated single-element model. For both sets of comparisons, the discrepancies in the calculated quantities were all <1%. The results validate the accuracy of the new code, which is a useful tool in scintillation detector optimization. (author)

NSURE code

International Nuclear Information System (INIS)

Rattan, D.S.

1993-11-01

NSURE stands for Near-Surface Repository code. NSURE is a performance assessment code. developed for the safety assessment of near-surface disposal facilities for low-level radioactive waste (LLRW). Part one of this report documents the NSURE model, governing equations and formulation of the mathematical models, and their implementation under the SYVAC3 executive. The NSURE model simulates the release of nuclides from an engineered vault, their subsequent transport via the groundwater and surface water pathways tot he biosphere, and predicts the resulting dose rate to a critical individual. Part two of this report consists of a User's manual, describing simulation procedures, input data preparation, output and example test cases

Quantum philosophy understanding and interpreting contemporary science

CERN Document Server

Omnès, Roland

2002-01-01

In this magisterial work, Roland Omnès takes us from the academies of ancient Greece to the laboratories of modern science as he seeks to do no less than rebuild the foundations of the philosophy of knowledge. One of the world's leading quantum physicists, Omnès reviews the history and recent development of mathematics, logic, and the physical sciences to show that current work in quantum theory offers new answers to questions that have puzzled philosophers for centuries: Is the world ultimately intelligible? Are all events caused? Do objects have definitive locations? Omnès addresses these p

CH-TRU Waste Content Codes

Energy Technology Data Exchange (ETDEWEB)

Washington TRU Solutions LLC

2008-01-16

The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

Calculation of the RSG-GAS core using computer code citation-3D

International Nuclear Information System (INIS)

Taryo, T.; Rokhmadi

1998-01-01

Since core reactivity is one of the reactor safety parameters, this R and D has been carried out. To carry out the R and D, the code called WIMSD4 was used respectively for generating cross section and diffusion parameters. The code CITATION was then applied to estimate core reactivity in the RSG-GAS core. To verify the result of the calculation, data and information of the RSG-GAS Typical Working Core Were used. To Prove the codes reliably used, the case of all control elements down in the reactor core and that of all control rods up in the core were applied. The result taking into account those cases showed respectively that K eff are less and greater than unity (K eff eff >1)

Percolation bounds for decoding thresholds with correlated erasures in quantum LDPC codes

Science.gov (United States)

Hamilton, Kathleen; Pryadko, Leonid

Correlations between errors can dramatically affect decoding thresholds, in some cases eliminating the threshold altogether. We analyze the existence of a threshold for quantum low-density parity-check (LDPC) codes in the case of correlated erasures. When erasures are positively correlated, the corresponding multi-variate Bernoulli distribution can be modeled in terms of cluster errors, where qubits in clusters of various size can be marked all at once. In a code family with distance scaling as a power law of the code length, erasures can be always corrected below percolation on a qubit adjacency graph associated with the code. We bound this correlated percolation transition by weighted (uncorrelated) percolation on a specially constructed cluster connectivity graph, and apply our recent results to construct several bounds for the latter. This research was supported in part by the NSF Grant PHY-1416578 and by the ARO Grant W911NF-14-1-0272.

RELAP5/MOD3 code manual: Summaries and reviews of independent code assessment reports. Volume 7, Revision 1

International Nuclear Information System (INIS)

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

1996-10-01

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

Validation of system codes RELAP5 and SPECTRA for natural convection boiling in narrow channels

Energy Technology Data Exchange (ETDEWEB)

Stempniewicz, M.M., E-mail: stempniewicz@nrg.eu; Slootman, M.L.F.; Wiersema, H.T.

2016-10-15

Highlights: • Computer codes RELAP5/Mod3.3 and SPECTRA 3.61 validated for boiling in narrow channels. • Validated codes can be used for LOCA analyses in research reactors. • Code validation based on natural convection boiling in narrow channels experiments. - Abstract: Safety analyses of LOCA scenarios in nuclear power plants are performed with so called thermal–hydraulic system codes, such as RELAP5. Such codes are validated for typical fuel geometries applied in nuclear power plants. The question considered by this article is if the codes can be applied for LOCA analyses in research reactors, in particular exceeding CHF in very narrow channels. In order to answer this question, validation calculations were performed with two thermal–hydraulic system codes: RELAP and SPECTRA. The validation was based on natural convection boiling in narrow channels experiments, performed by Prof. Monde et al. in the years 1990–2000. In total 42 vertical tube and annulus experiments were simulated with both codes. A good agreement of the calculated values with the measured data was observed. The main conclusions are: • The computer codes RELAP5/Mod 3.3 (US NRC version) and SPECTRA 3.61 have been validated for natural convection boiling in narrow channels using experiments of Monde. The dimensions applied in the experiments were performed for a range that covers the values observed in typical research reactors. Therefore it is concluded that both codes are validated and can be used for LOCA analyses in research reactors, including natural convection boiling. The applicability range of the present validation is: hydraulic diameters of 1.1 ⩽ D{sub hyd} ⩽ 9.0 mm, heated lengths of 0.1 ⩽ L ⩽ 1.0 m, pressures of 0.10 ⩽ P ⩽ 0.99 MPa. In most calculations the burnout was predicted to occur at lower power than that observed in the experiments. In several cases the burnout was observed at higher power. The overprediction was not larger than 16% in RELAP and 15% in

SUMMARY OF GENERAL WORKING GROUP A+B+D: CODES BENCHMARKING.

Energy Technology Data Exchange (ETDEWEB)

WEI, J.; SHAPOSHNIKOVA, E.; ZIMMERMANN, F.; HOFMANN, I.

2006-05-29

Computer simulation is an indispensable tool in assisting the design, construction, and operation of accelerators. In particular, computer simulation complements analytical theories and experimental observations in understanding beam dynamics in accelerators. The ultimate function of computer simulation is to study mechanisms that limit the performance of frontier accelerators. There are four goals for the benchmarking of computer simulation codes, namely debugging, validation, comparison and verification: (1) Debugging--codes should calculate what they are supposed to calculate; (2) Validation--results generated by the codes should agree with established analytical results for specific cases; (3) Comparison--results from two sets of codes should agree with each other if the models used are the same; and (4) Verification--results from the codes should agree with experimental measurements. This is the summary of the joint session among working groups A, B, and D of the HI32006 Workshop on computer codes benchmarking.

Greedy vs. L1 convex optimization in sparse coding

DEFF Research Database (Denmark)

Ren, Huamin; Pan, Hong; Olsen, Søren Ingvor

2015-01-01

Sparse representation has been applied successfully in many image analysis applications, including abnormal event detection, in which a baseline is to learn a dictionary from the training data and detect anomalies from its sparse codes. During this procedure, sparse codes which can be achieved...... solutions. Considering the property of abnormal event detection, i.e., only normal videos are used as training data due to practical reasons, effective codes in classification application may not perform well in abnormality detection. Therefore, we compare the sparse codes and comprehensively evaluate...... their performance from various aspects to better understand their applicability, including computation time, reconstruction error, sparsity, detection...

Specification of a test problem for HYDROCOIN [Hydrologic Code Intercomparison] Level 3 Case 2: Sensitivity analysis for deep disposal in partially saturated, fractured tuff

International Nuclear Information System (INIS)

Prindle, R.W.

1987-08-01

The international Hydrologic Code Intercomparison Project (HYDROCOIN) was formed to evaluate hydrogeologic models and computer codes and their use in performance assessment for high-level radioactive waste repositories. Three principal activities in the HYDROCOIN Project are Level 1, verification and benchmarking of hydrologic codes; Level 2, validation of hydrologic models; and Level 3, sensitivity and uncertainty analyses of the models and codes. This report presents a test case defined for the HYDROCOIN Level 3 activity to explore the feasibility of applying various sensitivity-analysis methodologies to a highly nonlinear model of isothermal, partially saturated flow through fractured tuff, and to develop modeling approaches to implement the methodologies for sensitivity analysis. These analyses involve an idealized representation of a repository sited above the water table in a layered sequence of welded and nonwelded, fractured, volcanic tuffs. The analyses suggested here include one-dimensional, steady flow; one-dimensional, nonsteady flow; and two-dimensional, steady flow. Performance measures to be used to evaluate model sensitivities are also defined; the measures are related to regulatory criteria for containment of high-level radioactive waste. 14 refs., 5 figs., 4 tabs

Uncertainty and sensitivity analysis using probabilistic system assessment code. 1

International Nuclear Information System (INIS)

Honma, Toshimitsu; Sasahara, Takashi.

1993-10-01

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

Z₂-double cyclic codes

OpenAIRE

Borges, J.

2014-01-01

A binary linear code C is a Z2-double cyclic code if the set of coordinates can be partitioned into two subsets such that any cyclic shift of the coordinates of both subsets leaves invariant the code. These codes can be identified as submodules of the Z2[x]-module Z2[x]/(x^r − 1) × Z2[x]/(x^s − 1). We determine the structure of Z2-double cyclic codes giving the generator polynomials of these codes. The related polynomial representation of Z2-double cyclic codes and its duals, and the relation...

Reduced subclinical carotid vascular disease and arterial stiffness in vegetarian men: The CARVOS Study.

Science.gov (United States)

Acosta-Navarro, Julio; Antoniazzi, Luiza; Oki, Adriana Midori; Bonfim, Maria Carlos; Hong, Valeria; Acosta-Cardenas, Pedro; Strunz, Celia; Brunoro, Eleonora; Miname, Marcio Hiroshi; Filho, Wilson Salgado; Bortolotto, Luiz Aparecido; Santos, Raul D

2017-03-01

Dietary habits play an important role in the development of atherosclerosis, the most important cause of morbidity and mortality in the world. The objective of this study was to verify if vegetarian (VEG) diet could be related a better profile of subclinical vascular disease evaluated by arterial stiffness and functional and structural properties of carotid arteries, compared to omnivorous (OMN) diet. In this cross-sectional study, 44 VEG and 44 OMN apparently healthy men ≥35years of age, in order to not have confounding risk factors of subclinical atherosclerosis, were assessed for anthropometric data, blood pressure, blood lipids, glucose, C reactive protein (CRP), and arterial stiffness determined by carotid-femoral pulse wave velocity (PWV). Also, carotid intima-media thickness (c-IMT) and distensibility were evaluated. VEG men had lower body mass index, systolic and diastolic blood pressures, fasting serum total cholesterol, LDL and non-HDL-cholesterol, apolipoprotein B, glucose and glycated hemoglobin values in comparison with OMN individuals (all p values <0.05). Markers of vascular structure and function were different between VEG and OMN: PWV 7.1±0.8m/s vs. 7.7±0.9m/s (p<0.001); c-IMT 593±94 vs. 661±128μm (p=0.003); and relative carotid distensibility 6.39±1.7 vs. 5.72±1.8% (p=0.042), respectively. After a multivariate linear regression analysis, a VEG diet was independently and negatively associated with PWV (p value 0.005). A VEG diet is associated with a more favorable cardiovascular diseases biomarker profile and better vascular structural and functional parameters. Copyright © 2016 Elsevier Ireland Ltd. All rights reserved.

Modeling of the installation with the code MELCOR 1.8.4

International Nuclear Information System (INIS)

Pomier Baez, L.E.; Nunez Mc Leod, J.E.

1998-01-01

Full text: The calculation code MELCOR 1.8.4 is an integrated program that allow to simulate the development of accidents in nuclear plants with refrigerated reactors with light water. This code can simulate the whole spectrum of phenomenons. This work carried out the validation of the packages of the code MELCOR dedicated to evaluate the behaviour under conditions of two-phase flow, through the comparison of the results of the simulation with the experimental data of the installation TPTF (Two-Phase Test Facility) (ROSA-IV) of the Institute JAERI (Japan Atomic Energy Research Institute) of Japan. The main objective of the experiments TPTF is obtain data on the thermohydraulic behaviour from light water reactors (PWR) during an accident with small loss of coolant (SBLOCA), and the capacity of MELCOR code was evaluated in the simulation of these kind of accidents. Diverse options of the code were studies, in order to analyzing the behaviour of the feigned phenomenon. The effect of the change in the nodalization of the nuclear installation was studies, as well as the management of diverse control functions. The results of the evaluation show a good concordance with the experimental data, especially in the prediction of the behaviour of the steam fraction in relation with the mass flow, the quality of the steam and the mixture level in the exit volume that represent two possibilities state in the vessel reactor during the accidental situation. (author) [es

Ethical and educational considerations in coding hand surgeries.

Science.gov (United States)

Lifchez, Scott D; Leinberry, Charles F; Rivlin, Michael; Blazar, Philip E

2014-07-01

To assess treatment coding knowledge and practices among residents, fellows, and attending hand surgeons. Through the use of 6 hypothetical cases, we developed a coding survey to assess coding knowledge and practices. We e-mailed this survey to residents, fellows, and attending hand surgeons. In additionally, we asked 2 professional coders to code these cases. A total of 71 participants completed the survey out of 134 people to whom the survey was sent (response rate = 53%). We observed marked disparity in codes chosen among surgeons and among professional coders. Results of this study indicate that coding knowledge, not just its ethical application, had a major role in coding procedures accurately. Surgical coding is an essential part of a hand surgeon's practice and is not well learned during residency or fellowship. Whereas ethical issues such as deliberate unbundling and upcoding may have a role in inaccurate coding, lack of knowledge among surgeons and coders has a major role as well. Coding has a critical role in every hand surgery practice. Inconstancies among those polled in this study reveal that an increase in education on coding during training and improvement in the clarity and consistency of the Current Procedural Terminology coding rules themselves are needed. Copyright © 2014 American Society for Surgery of the Hand. Published by Elsevier Inc. All rights reserved.

Evaluation of the Trac-PF1 code for simulating the Neptun reflooding experiment

International Nuclear Information System (INIS)

Pontedeiro, A.C.; Galetti, M.R.S.

1991-01-01

The present work presents an assessment of the TRAC-BF1 code using the results of the NEPTUN experiment which simulates the reflooding in a loss-of-coolant accident (LOCA) in a PWR. The NEPTUN experiment is composed of an array of electrically-heated tubes where the reflooding condition can be tested. Two types of tests results are presented and compared with the values obtained with the TRAC-BF1 code. From this comparison it is concluded that TRAC is suitable for verifying accident analysis. (author)

Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

International Nuclear Information System (INIS)

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

1997-12-01

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

Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-01

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

Audit of Clinical Coding of Major Head and Neck Operations

Science.gov (United States)

Mitra, Indu; Malik, Tass; Homer, Jarrod J; Loughran, Sean

2009-01-01

INTRODUCTION Within the NHS, operations are coded using the Office of Population Censuses and Surveys (OPCS) classification system. These codes, together with diagnostic codes, are used to generate Healthcare Resource Group (HRG) codes, which correlate to a payment bracket. The aim of this study was to determine whether allocated procedure codes for major head and neck operations were correct and reflective of the work undertaken. HRG codes generated were assessed to determine accuracy of remuneration. PATIENTS AND METHODS The coding of consecutive major head and neck operations undertaken in a tertiary referral centre over a retrospective 3-month period were assessed. Procedure codes were initially ascribed by professional hospital coders. Operations were then recoded by the surgical trainee in liaison with the head of clinical coding. The initial and revised procedure codes were compared and used to generate HRG codes, to determine whether the payment banding had altered. RESULTS A total of 34 cases were reviewed. The number of procedure codes generated initially by the clinical coders was 99, whereas the revised codes generated 146. Of the original codes, 47 of 99 (47.4%) were incorrect. In 19 of the 34 cases reviewed (55.9%), the HRG code remained unchanged, thus resulting in the correct payment. Six cases were never coded, equating to £15,300 loss of payment. CONCLUSIONS These results highlight the inadequacy of this system to reward hospitals for the work carried out within the NHS in a fair and consistent manner. The current coding system was found to be complicated, ambiguous and inaccurate, resulting in loss of remuneration. PMID:19220944

Specialized Monte Carlo codes versus general-purpose Monte Carlo codes

International Nuclear Information System (INIS)

Moskvin, Vadim; DesRosiers, Colleen; Papiez, Lech; Lu, Xiaoyi

2002-01-01

The possibilities of Monte Carlo modeling for dose calculations and optimization treatment are quite limited in radiation oncology applications. The main reason is that the Monte Carlo technique for dose calculations is time consuming while treatment planning may require hundreds of possible cases of dose simulations to be evaluated for dose optimization. The second reason is that general-purpose codes widely used in practice, require an experienced user to customize them for calculations. This paper discusses the concept of Monte Carlo code design that can avoid the main problems that are preventing wide spread use of this simulation technique in medical physics. (authors)

Simulation and interpretation codes for the JET ECE diagnostic. Part 1: physics of the codes' operation

International Nuclear Information System (INIS)

Bartlett, D.V.

1983-06-01

The codes which have been developed for the analysis of electron cyclotron emission measurements in JET are described. Their principal function is to interpret the spectra measured by the diagnostic so as to give the spatial distribution of the electron temperature in the poloidal cross-section. Various systematic effects in the data are corrected using look-up tables generated by an elaborate simulation code. The part of this code responsible for the accurate calculation of single-pass emission and refraction has been written at CNR-Milan and is described in a separate report. The present report is divided into two parts. This first part describes the methods used for the simulation and interpretation of spectra, the physical/mathematical basis of the codes written at CEA-Fontenay and presents some illustrative results

A first accident simulation for Angra-1 power plant using the ALMOD computer code

International Nuclear Information System (INIS)

Camargo, C.T.M.

1981-02-01

The acquisition of the Almod computer code from GRS-Munich to CNEN has permited doing calculations of transients in PWR nuclear power plants, in which doesn't occur loss of coolant. The implementation of the german computer code Almod and its application in the calculation of Angra-1, a nuclear power plant different from the KWU power plants, demanded study and models adaptation; and due to economic reasons simplifications and optimizations were necessary. The first results define the analytical potential of the computer code, confirm the adequacy of the adaptations done and provide relevant conclusions about the Angra-1 safety analysis, showing at the same time areas in which the model can be applied or simply improved. (Author) [pt

An evaluation of TRAC-PF1/MOD1 computer code performance during posttest simulations of Semiscale MOD-2C feedwater line break transients

International Nuclear Information System (INIS)

Hall, D.G.; Watkins, J.C.

1987-01-01

This report documents an evaluation of the TRAC-PF1/MOD1 reactor safety analysis computer code during computer simulations of feedwater line break transients. The experimental data base for the evaluation included the results of three bottom feedwater line break tests performed in the Semiscale Mod-2C test facility. The tests modeled 14.3% (S-FS-7), 50% (S-FS-11), and 100% (S-FS-6B) breaks. The test facility and the TRAC-PF1/MOD1 model used in the calculations are described. Evaluations of the accuracy of the calculations are presented in the form of comparisons of measured and calculated histories of selected parameters associated with the primary and secondary systems. In addition to evaluating the accuracy of the code calculations, the computational performance of the code during the simulations was assessed. A conclusion was reached that the code is capable of making feedwater line break transient calculations efficiently, but there is room for significant improvements in the simulations that were performed. Recommendations are made for follow-on investigations to determine how to improve future feedwater line break calculations and for code improvements to make the code easier to use

Implantation of TRAC-PF1 computer code of VAX-11/750

International Nuclear Information System (INIS)

Madeira, A.A.; Souza Gouvea, L. de; Galetti, M.R.S.

1988-01-01

The implantation of TRAC-PF1 code, IBM version, in VAX-11/750 is described. This work provides Reator Department with an advanced best-estimate tool to perform loss-of-coolant accident analysis. (Author) [pt

Use of AERIN code for determining internal doses of transuranic isotopes

International Nuclear Information System (INIS)

King, W.C.

1980-01-01

The AERIN computer code is a mathematical expression of the ICRP Lung Model. The code was developed at the Lawrence Livermore National Laboratory to compute the body organ burdens and absorbed radiation doses resulting from the inhalation of transuranic isotopes and to predict the amount of activity excreted in the urine and feces as a function of time. Over forty cases of internal exposure have been studied using the AERIN code. The code, as modified, has proven to be extremely versatile. The case studies presented demonstrate the excellent correlation that can be obtained between code predictions and observed bioassay data. In one case study a discrepancy was observed between an in vivo count of the whole body and the application of the code using urine and fecal data as input. The discrepancy was resolved by in vivo skull counts that showed the code had predicted the correct skeletal burden

The PP1 binding code: a molecular-lego strategy that governs specificity.

Science.gov (United States)

Heroes, Ewald; Lesage, Bart; Görnemann, Janina; Beullens, Monique; Van Meervelt, Luc; Bollen, Mathieu

2013-01-01

Ser/Thr protein phosphatase 1 (PP1) is a single-domain hub protein with nearly 200 validated interactors in vertebrates. PP1-interacting proteins (PIPs) are ubiquitously expressed but show an exceptional diversity in brain, testis and white blood cells. The binding of PIPs is mainly mediated by short motifs that dock to surface grooves of PP1. Although PIPs often contain variants of the same PP1 binding motifs, they differ in the number and combination of docking sites. This molecular-lego strategy for binding to PP1 creates holoenzymes with unique properties. The PP1 binding code can be described as specific, universal, degenerate, nonexclusive and dynamic. PIPs control associated PP1 by interference with substrate recruitment or access to the active site. In addition, some PIPs have a subcellular targeting domain that promotes dephosphorylation by increasing the local concentration of PP1. The diversity of the PP1 interactome and the properties of the PP1 binding code account for the exquisite specificity of PP1 in vivo. © 2012 The Authors Journal compilation © 2012 FEBS.

Procedure and code for calculating black control rods taking into account epithermal absorption, code CAS-1; Postupak i program za proracun crnih kontrolnih sipki, uzimajuci u obzir i epitermalnu apsorpciju, CAS-1

Energy Technology Data Exchange (ETDEWEB)

Martinc, R; Trivunac, N; Zivkovic, Z [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

1964-12-15

This report describes the computer code CAS-1, calculation method and procedure applied for calculating the black control rods taking into account the epithermal neutron absorption. Results obtained for supercell method applied for regular lattice reflected in the multiplication medium is part of this report in addition to the computer code manual.

Motion-adaptive intraframe transform coding of video signals

NARCIS (Netherlands)

With, de P.H.N.

1989-01-01

Spatial transform coding has been widely applied for image compression because of its high coding efficiency. However, in many intraframe systems, in which every TV frame is independently processed, coding of moving objects in the case of interlaced input signals is not addressed. In this paper, we

TESLA: Large Signal Simulation Code for Klystrons

International Nuclear Information System (INIS)

Vlasov, Alexander N.; Cooke, Simon J.; Chernin, David P.; Antonsen, Thomas M. Jr.; Nguyen, Khanh T.; Levush, Baruch

2003-01-01

TESLA (Telegraphist's Equations Solution for Linear Beam Amplifiers) is a new code designed to simulate linear beam vacuum electronic devices with cavities, such as klystrons, extended interaction klystrons, twistrons, and coupled cavity amplifiers. The model includes a self-consistent, nonlinear solution of the three-dimensional electron equations of motion and the solution of time-dependent field equations. The model differs from the conventional Particle in Cell approach in that the field spectrum is assumed to consist of a carrier frequency and its harmonics with slowly varying envelopes. Also, fields in the external cavities are modeled with circuit like equations and couple to fields in the beam region through boundary conditions on the beam tunnel wall. The model in TESLA is an extension of the model used in gyrotron code MAGY. The TESLA formulation has been extended to be capable to treat the multiple beam case, in which each beam is transported inside its own tunnel. The beams interact with each other as they pass through the gaps in their common cavities. The interaction is treated by modification of the boundary conditions on the wall of each tunnel to include the effect of adjacent beams as well as the fields excited in each cavity. The extended version of TESLA for the multiple beam case, TESLA-MB, has been developed for single processor machines, and can run on UNIX machines and on PC computers with a large memory (above 2GB). The TESLA-MB algorithm is currently being modified to simulate multiple beam klystrons on multiprocessor machines using the MPI (Message Passing Interface) environment. The code TESLA has been verified by comparison with MAGIC for single and multiple beam cases. The TESLA code and the MAGIC code predict the same power within 1% for a simple two cavity klystron design while the computational time for TESLA is orders of magnitude less than for MAGIC 2D. In addition, recently TESLA was used to model the L-6048 klystron, code

Writing the Live Coding Book

DEFF Research Database (Denmark)

Blackwell, Alan; Cox, Geoff; Lee, Sang Wong

2016-01-01

This paper is a speculation on the relationship between coding and writing, and the ways in which technical innovations and capabilities enable us to rethink each in terms of the other. As a case study, we draw on recent experiences of preparing a book on live coding, which integrates a wide range...... of personal, historical, technical and critical perspectives. This book project has been both experimental and reflective, in a manner that allows us to draw on critical understanding of both code and writing, and point to the potential for new practices in the future....

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

Improvement of Level-1 PSA computer code package -A study for nuclear safety improvement-

International Nuclear Information System (INIS)

Park, Chang Kyu; Kim, Tae Woon; Ha, Jae Joo; Han, Sang Hoon; Cho, Yeong Kyun; Jeong, Won Dae; Jang, Seung Cheol; Choi, Young; Seong, Tae Yong; Kang, Dae Il; Hwang, Mi Jeong; Choi, Seon Yeong; An, Kwang Il

1994-07-01

This year is the second year of the Government-sponsored Mid- and Long-Term Nuclear Power Technology Development Project. The scope of this subproject titled on 'The Improvement of Level-1 PSA Computer Codes' is divided into three main activities : (1) Methodology development on the under-developed fields such as risk assessment technology for plant shutdown and external events, (2) Computer code package development for Level-1 PSA, (3) Applications of new technologies to reactor safety assessment. At first, in the area of PSA methodology development, foreign PSA reports on shutdown and external events have been reviewed and various PSA methodologies have been compared. Level-1 PSA code KIRAP and CCF analysis code COCOA are converted from KOS to Windows. Human reliability database has been also established in this year. In the area of new technology applications, fuzzy set theory and entropy theory are used to estimate component life and to develop a new measure of uncertainty importance. Finally, in the field of application study of PSA technique to reactor regulation, a strategic study to develop a dynamic risk management tool PEPSI and the determination of inspection and test priority of motor operated valves based on risk importance worths have been studied. (Author)

MELCOR computer code manuals: Primer and user's guides, Version 1.8.3 September 1994. Volume 1

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 US 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

Compiler Driven Code Comments and Refactoring

DEFF Research Database (Denmark)

Larsen, Per; Ladelsky, Razya; Karlsson, Sven

2011-01-01

. We demonstrate the ability of our tool to trans- form code, and suggest code refactoring that increase its amenability to optimization. The preliminary results shows that, with our tool-set, au- tomatic loop parallelization with the GNU C compiler, gcc, yields 8.6x best-case speedup over...

Modelling of blackout sequence at Atucha-1 using the MARCH3 code

International Nuclear Information System (INIS)

Baron, J.; Bastianelli, B.

1997-01-01

This paper presents the modelling of a complete blackout at the Atucha-1 NPP as preliminary phase for a Level II safety probabilistic analysis. The MARCH3 code of the STCP (Source Term Code Package) is used, based on a plant model made in accordance with particularities of the plant design. The analysis covers all the severe accident phases. The results allow to view the time sequence of the events, and provide the basis for source term studies. (author). 6 refs., 2 figs

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

Self-complementary circular codes in coding theory.

Science.gov (United States)

Fimmel, Elena; Michel, Christian J; Starman, Martin; Strüngmann, Lutz

2018-04-01

Self-complementary circular codes are involved in pairing genetic processes. A maximal [Formula: see text] self-complementary circular code X of trinucleotides was identified in genes of bacteria, archaea, eukaryotes, plasmids and viruses (Michel in Life 7(20):1-16 2017, J Theor Biol 380:156-177, 2015; Arquès and Michel in J Theor Biol 182:45-58 1996). In this paper, self-complementary circular codes are investigated using the graph theory approach recently formulated in Fimmel et al. (Philos Trans R Soc A 374:20150058, 2016). A directed graph [Formula: see text] associated with any code X mirrors the properties of the code. In the present paper, we demonstrate a necessary condition for the self-complementarity of an arbitrary code X in terms of the graph theory. The same condition has been proven to be sufficient for codes which are circular and of large size [Formula: see text] trinucleotides, in particular for maximal circular codes ([Formula: see text] trinucleotides). For codes of small-size [Formula: see text] trinucleotides, some very rare counterexamples have been constructed. Furthermore, the length and the structure of the longest paths in the graphs associated with the self-complementary circular codes are investigated. It has been proven that the longest paths in such graphs determine the reading frame for the self-complementary circular codes. By applying this result, the reading frame in any arbitrary sequence of trinucleotides is retrieved after at most 15 nucleotides, i.e., 5 consecutive trinucleotides, from the circular code X identified in genes. Thus, an X motif of a length of at least 15 nucleotides in an arbitrary sequence of trinucleotides (not necessarily all of them belonging to X) uniquely defines the reading (correct) frame, an important criterion for analyzing the X motifs in genes in the future.

TRANSURANUS: A fuel rod analysis code ready for use

Energy Technology Data Exchange (ETDEWEB)

Lassmann, K; O` Carroll, C; Van de Laar, J [Commission of the European Communities, Karlsruhe (Germany). European Inst. for Transuranium Elements; Ott, C [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1994-12-31

The basic concepts of fuel rod performance codes are discussed. The TRANSURANUS code developed at the Institute for Transuranium Elements, Karlsruhe (GE) is presented. It is a quasi two-dimensional (1{sub 1/2}-D) code designed for treatment of a whole fuel rod for any type of reactor and any situation. The fuel rods found in the majority of test- or power reactors can be analyzed for very different situations (normal, off-normal and accidental). The time scale of the problems to be treated may range from milliseconds to years. The TRANSURANUS code consists of a clearly defined mechanical/mathematical framework into which physical models can easily be incorporated. This framework has been extensively tested and the programming very clearly reflects this structure. The code is well structured and easy to understand. It has a comprehensive material data bank for different fuels, claddings, coolants and their properties. The code can be employed in a deterministic and a statistical version. It is written in standard FORTRAN 77. The code system includes: 2 preprocessor programs (MAKROH and AXORDER) for setting up new data cases; the post-processor URPLOT for plotting all important quantities as a function of the radius, the axial coordinate or the time; the post-processor URSTART evaluating statistical analyses. The TRANSURANUS code exhibits short running times. A new WINDOWS-based interactive interface is under development. The code is now in use in various European institutions and is available to all interested parties. 7 figs., 15 refs.

SCDAP/RELAP5/MOD 3.1 Code Manual: Developmental assessment. Volume 5

Energy Technology Data Exchange (ETDEWEB)

Hohorst, J.K.; Johnsen, E.C. [eds.; Allison, C.M. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

1995-06-01

The SCDAP/RELAP5 code has been developed for best estimate transient simulation of Light Water Reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the 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. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume contains detailed code-to-data calculations performed using SCDAP/RELAP5/MOD3.1, as well as comparison calculations performed with earlier code versions. Results of full plant calculations which include Surry, TMI-2, and Browns Ferry are described. Results of a nodalization study, which accounted for both axial and radial nodalization of the core, are also reported.

SCDAP/RELAP5/MOD 3.1 Code Manual: Developmental assessment. Volume 5

International Nuclear Information System (INIS)

Hohorst, J.K.; Johnsen, E.C.

1995-06-01

The SCDAP/RELAP5 code has been developed for best estimate transient simulation of Light Water Reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the 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. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume contains detailed code-to-data calculations performed using SCDAP/RELAP5/MOD3.1, as well as comparison calculations performed with earlier code versions. Results of full plant calculations which include Surry, TMI-2, and Browns Ferry are described. Results of a nodalization study, which accounted for both axial and radial nodalization of the core, are also reported

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

Deterministic dense coding with partially entangled states

Science.gov (United States)

Mozes, Shay; Oppenheim, Jonathan; Reznik, Benni

2005-01-01

The utilization of a d -level partially entangled state, shared by two parties wishing to communicate classical information without errors over a noiseless quantum channel, is discussed. We analytically construct deterministic dense coding schemes for certain classes of nonmaximally entangled states, and numerically obtain schemes in the general case. We study the dependency of the maximal alphabet size of such schemes on the partially entangled state shared by the two parties. Surprisingly, for d>2 it is possible to have deterministic dense coding with less than one ebit. In this case the number of alphabet letters that can be communicated by a single particle is between d and 2d . In general, we numerically find that the maximal alphabet size is any integer in the range [d,d2] with the possible exception of d2-1 . We also find that states with less entanglement can have a greater deterministic communication capacity than other more entangled states.

Medical Ethics Code: an Analysis from Ethical-Disciplinary Cases Against Medical Professionals within the Specialty of Psychiatry

Directory of Open Access Journals (Sweden)

Giselle Crosara Gracindo

2018-02-01

Full Text Available Objective: To identify the nature of infractions committed by doctors working within the field of psychiatry, between 2010 and 2016, from the scope of appeals within ethical-disciplinary cases judged at the Plenary Tribunal of the Federal Medical Council, based on the medical ethics code, and to list some elements that make it possible to outline the professional profile of those involved. Method: This was a document-based investigation in the form of a retrospective and descriptive study. Data were gathered using the Federal Medical Council (CFM database and from consultation of judgments issued by the Plenary Body of the Medical Ethics Tribunal (TSEM, of the CFM. The investigation used a sample consisting of 206 appeals and 19 referrals, totaling 224 appeals by doctors who underwent trials. We took into account cases judged between April 13, 2010 and August 3, 2016. Three databases were used in the investigation: cases (224; doctors facing charges (191 and cases/penalties (146. Based on the records of the 191 doctors charged, the ethical-disciplinary cases of seven doctors working in psychiatry were analyzed specifically for the present study, whether or not they had a specialist title. Characterization of infractions committed encompassed references to the articles of the medical ethics code most frequently infringed in the field of psychiatry, along with a survey of the motives for these infractions and some characteristics relating to these professionals’ profile. Results: Among the findings from this investigation, infractions of the articles of the medical ethics code can be highlighted, such as article 30 “[...] Use of the profession to corrupt customs and to commit or favor crime [...]” and article 40 “[...] Taking advantage of situations arising from the doctor-patient relationship to obtain physical, emotional, financial or any other advantage [...]”. The professional profile of those involved in these cases was also shown

Validation of the AZTRAN 1.1 code with problems Benchmark of LWR reactors

International Nuclear Information System (INIS)

Vallejo Q, J. A.; Bastida O, G. E.; Francois L, J. L.; Xolocostli M, J. V.; Gomez T, A. M.

2016-09-01

The AZTRAN module is a computational program that is part of the AZTLAN platform (Mexican modeling platform for the analysis and design of nuclear reactors) and that solves the neutron transport equation in 3-dimensional using the discrete ordinates method S_N, steady state and Cartesian geometry. As part of the activities of Working Group 4 (users group) of the AZTLAN project, this work validates the AZTRAN code using the 2002 Yamamoto Benchmark for LWR reactors. For comparison, the commercial code CASMO-4 and the free code Serpent-2 are used; in addition, the results are compared with the data obtained from an article of the PHYSOR 2002 conference. The Benchmark consists of a fuel pin, two UO_2 cells and two other of MOX cells; there is a problem of each cell for each type of reactor PWR and BWR. Although the AZTRAN code is at an early stage of development, the results obtained are encouraging and close to those reported with other internationally accepted codes and methodologies. (Author)

Is Self-Regulation Sufficient? Case of the German Transparency Code

Directory of Open Access Journals (Sweden)

Kristin Buske

2016-02-01

Full Text Available The German pharmaceutical industry is stepping ahead with its implementation of a new transparency disclosure code for cooperation between pharmaceutical companies and health care professionals (HCPs and health care organisations (HCOs. In Germany, this transparency code (“Transparenzkodex” is applicable since January 2015, and data will be publicly available around mid-2016. No empirical work has been done that addresses the impact of the transparency code on cooperation between HCPs, HCOs and the pharmaceutical companies, including the possibilities of competitive analysis of the available data. In this paper, we interviewed experts from 11 pharmaceutical companies representing small, medium-sized as well as multinational corporations which represent 80% of the German pharmaceutical market. Besides interviews, the authors designed a game to evaluate possible financial investments in key opinion leaders. The market can be regarded as a zero sum game. By allowing public identification of such key HCPs and HCOs, the amount spent on them might increase and not decrease. In a way, the transparency code may foster more and not less spending; in our simulation game, the financial investment in marketing key HCPs and HCOs exceeded sustainable limits.

microRNA-9 targets the long non-coding RNA MALAT1 for degradation in the nucleus

DEFF Research Database (Denmark)

Leucci, Eleonora; Patella, Francesca; Waage, Johannes

2013-01-01

-coding RNAs. Here we report that microRNA-9 (miR-9) regulates the expression of the Metastasis Associated Lung Adenocarcinoma Transcript 1 (MALAT-1), one of the most abundant and conserved long non-coding RNAs. Intriguingly, we find that miR-9 targets AGO2-mediated regulation of MALAT1 in the nucleus. Our...

Integrated Tiger Series of electron/photon Monte Carlo transport codes: a user's guide for use on IBM mainframes

International Nuclear Information System (INIS)

Kirk, B.L.

1985-12-01

The ITS (Integrated Tiger Series) Monte Carlo code package developed at Sandia National Laboratories and distributed as CCC-467/ITS by the Radiation Shielding Information Center (RSIC) at Oak Ridge National Laboratory (ORNL) consists of eight codes - the standard codes, TIGER, CYLTRAN, ACCEPT; the P-codes, TIGERP, CYLTRANP, ACCEPTP; and the M-codes ACCEPTM, CYLTRANM. The codes have been adapted to run on the IBM 3081, VAX 11/780, CDC-7600, and Cray 1 with the use of the update emulator UPEML. This manual should serve as a guide to a user running the codes on IBM computers having 370 architecture. The cases listed were tested on the IBM 3033, under the MVS operating system using the VS Fortran Level 1.3.1 compiler

User Effect on Code Application and Qualification Needs

International Nuclear Information System (INIS)

D'Auria, F.; Salah, A.B.

2008-01-01

Experience with some code assessment case studies and also additional ISPs have shown the dominant effect of the code user on the predicted system behavior. The general findings of the user effect investigations on some of the case studies indicate, specifically, that in addition to user effects, there are other reasons which affect the results of the calculations and are hidden under the general title of user effects. The specific characteristics of experimental facilities, i.e. limitations as far as code assessment is concerned; limitations of the used thermal-hydraulic codes to simulate certain system behavior or phenomena; limitations due to interpretation of experimental data by the code user, i.e. interpretation of experimental data base. On the basis of the discussions in this paper, the following conclusions and recommendations can be made: More dialogue appears to be necessary with the experimenters in the planning of code assessment calculations, e.g. ISPs.; User guidelines are not complete for the codes and the lack of sufficient and detailed user guidelines are observed with some of the case studies; More extensive user instruction and training, improved user guidelines, or quality assurance procedures may partially reduce some of the subjective user influence on the calculated results; The discrepancies between experimental data and code predictions are due both to the intrinsic code limit and to the so called user effects. There is a worthful need to quantify the percentage of disagreement due to the poor utilization of the code and due to the code itself. This need especially arises for the uncertainty evaluation studies (e.g. [18]) which do not take into account the mentioned user effects; A much focused investigation, based on the results of comparison calculations e.g. ISPs, analyzing the experimental data and the results of the specific code in order to evaluate the user effects and the related experimental aspects should be integral part of the

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

Science.gov (United States)

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

2018-02-01

systems were used in Europe (59%) and North America (57%). 34% of the reviewed coding systems were utilized in at least 1 of the 16 pharmacoepidemiology databases of interest evaluated. 21% of coding systems had studies evaluating the validity and consistency of their use in research within pharmacoepidemiology databases of interest. The most prevalent validation method was comparison with a review of patient charts, case notes or medical records (64% of reviewed validation studies). The reported performance measures in the reviewed studies varied across a large range of values (PPV 0-100%, NPV 6-100%, sensitivity 0-100%, specificity 23-100% and accuracy 16-100%) and were dependent on many factors including coding system(s), therapeutic area, pharmacoepidemiology database, and outcome. Coding systems vary by type of information captured, clinical setting, and pharmacoepidemiology database and region of use. Of the 57 reviewed coding systems, few are routinely and widely applied in pharmacoepidemiology database research. Indication and outcome dependent heterogeneity in coding system performance suggest that accurate definitions and algorithms for capturing specific exposures and outcomes within large healthcare datasets should be developed on a case-by-case basis and in consultation with clinical experts. Schattauer GmbH.

Theoretical Atomic Physics code development II: ACE: Another collisional excitation code

International Nuclear Information System (INIS)

Clark, R.E.H.; Abdallah, J. Jr.; Csanak, G.; Mann, J.B.; Cowan, R.D.

1988-12-01

A new computer code for calculating collisional excitation data (collision strengths or cross sections) using a variety of models is described. The code uses data generated by the Cowan Atomic Structure code or CATS for the atomic structure. Collisional data are placed on a random access file and can be displayed in a variety of formats using the Theoretical Atomic Physics Code or TAPS. All of these codes are part of the Theoretical Atomic Physics code development effort at Los Alamos. 15 refs., 10 figs., 1 tab

Lattice-Like Total Perfect Codes

Directory of Open Access Journals (Sweden)

Araujo Carlos

2014-02-01

Full Text Available A contribution is made to the classification of lattice-like total perfect codes in integer lattices Λn via pairs (G, Φ formed by abelian groups G and homomorphisms Φ: Zn → G. A conjecture is posed that the cited contribution covers all possible cases. A related conjecture on the unfinished work on open problems on lattice-like perfect dominating sets in Λn with induced components that are parallel paths of length > 1 is posed as well.

Development of a friendly interface for ORIGEN 2.1 code using MatLab software

International Nuclear Information System (INIS)

Vieira, Joao Paulo

2011-01-01

In an event of accidental liberation of radioactive material to the environment from a nuclear power plant, decisions must be taken quickly to supply the need of mitigating actions. Thus, it is important a fast, clear and safe access to all information about the source term. This work describes the initiative to develop a graphic interface to output data for ORIGEN 2.1 code, intending a friendly and secure approach with the output data and other important parameters for an analysis in emergency case, known the historic of operation of a nuclear power plant type PWR. By using the software MATLAB it is possible to develop an output routine with graphic presentation to some necessary data for an emergency analysis. The interface output must be able of fix up the ORIGEN conventional tables in graphics. In advance, preliminary results will be presented. (author)

Adventure Code Camp: Library Mobile Design in the Backcountry

OpenAIRE

Ward, David; Hahn, James; Mestre, Lori

2014-01-01

This article presents a case study exploring the use of a student Coding Camp as a bottom-up mobile design process to generate library mobile apps. A code camp sources student programmer talent and ideas for designing software services and features.  This case study reviews process, outcomes, and next steps in mobile web app coding camps. It concludes by offering implications for services design beyond the local camp presented in this study. By understanding how patrons expect to integrate li...

Development, validation and application of NAFA 2D-CFD code

International Nuclear Information System (INIS)

Vaidya, A.M.; Maheshwari, N.K.; Vijayan, P.K.; Saha, D.

2010-01-01

A 2D axi-symmetric code named NAFA (Version 1.0) is developed for studying the pipe flow under various conditions. It can handle laminar/ turbulent flows, with or without heat transfer, under sub-critical/super-critical conditions. The code solves for momentum, energy equations with standard k-ε turbulence model (with standard wall functions). It solves pipe flow subjected to 'velocity inlet', 'wall', 'axis' and 'pressure outlet' boundary conditions. It is validated for several cases by comparing its results with experimental data/analytical solutions/correlations. The code has excellent convergence characteristics as verified from fall of equation residual in each case. It has proven capability of generating mesh independent results for laminar as well as turbulent flows. The code is applied to supercritical flows. For supercritical flows, the effect of mesh size on prediction of heat transfer coefficient is studied. With grid refinement, the Y + reduces and reaches the limiting value of 11.63. Hence the accuracy is found to increase with grid refinement. NAFA is able to qualitatively predict the effect of heat flux and operating pressure on heat transfer coefficient. The heat transfer coefficient matches well with experimental values under various conditions. (author)

Quantum computation with Turaev-Viro codes

International Nuclear Information System (INIS)

Koenig, Robert; Kuperberg, Greg; Reichardt, Ben W.

2010-01-01

For a 3-manifold with triangulated boundary, the Turaev-Viro topological invariant can be interpreted as a quantum error-correcting code. The code has local stabilizers, identified by Levin and Wen, on a qudit lattice. Kitaev's toric code arises as a special case. The toric code corresponds to an abelian anyon model, and therefore requires out-of-code operations to obtain universal quantum computation. In contrast, for many categories, such as the Fibonacci category, the Turaev-Viro code realizes a non-abelian anyon model. A universal set of fault-tolerant operations can be implemented by deforming the code with local gates, in order to implement anyon braiding. We identify the anyons in the code space, and present schemes for initialization, computation and measurement. This provides a family of constructions for fault-tolerant quantum computation that are closely related to topological quantum computation, but for which the fault tolerance is implemented in software rather than coming from a physical medium.

Development of an analysis code for pressure wave propagation, (1)

International Nuclear Information System (INIS)

Tanaka, Yoshihisa; Sakano, Kosuke; Shindo, Yoshihisa

1974-11-01

We analyzed the propagation of the pressure-wave in the piping system of SWAT-1B rig by using SWAC-5 Code. We carried out analyses on the following parts. 1) A straight pipe 2) Branches 3) A piping system The results obtained in these analyses are as follows. 1) The present our model simulates well the straight pipe and the branch with the same diameters. 2) The present our model simulates approximately the branch with the different diameters and the piping system. (auth.)

CALMAR: A New Versatile Code Library for Adjustment from Measurements

Directory of Open Access Journals (Sweden)

Grégoire G.

2016-01-01

Full Text Available CALMAR, a new library for adjustment has been developed. This code performs simultaneous shape and level adjustment of an initial prior spectrum from measured reactions rates of activation foils. It is written in C++ using the ROOT data analysis framework,with all linear algebra classes. STAYSL code has also been reimplemented in this library. Use of the code is very flexible : stand-alone, inside a C++ code, or driven by scripts. Validation and test cases are under progress. Theses cases will be included in the code package that will be available to the community. Future development are discussed. The code should support the new Generalized Nuclear Data (GND format. This new format has many advantages compared to ENDF.

How Effective Are Codes of Ethics? A Look at Three Newsrooms.

Science.gov (United States)

Boeyink, David E.

1994-01-01

Presents case studies of three newsrooms and their codes of ethics. Finds that while the content of codes is not irrelevant, the value of codes needs to be examined in the context of the newsroom environment; and that ethical discussions and debate within the newsroom are critical to bridge the gap between ethical standards and concrete cases. (RS)

Cardiorespiratory Fitness and Peak Torque Differences between Vegetarian and Omnivore Endurance Athletes: A Cross-Sectional Study.

Science.gov (United States)

Lynch, Heidi M; Wharton, Christopher M; Johnston, Carol S

2016-11-15

In spite of well-documented health benefits of vegetarian diets, less is known regarding the effects of these diets on athletic performance. In this cross-sectional study, we compared elite vegetarian and omnivore adult endurance athletes for maximal oxygen uptake (VO2 max) and strength. Twenty-seven vegetarian (VEG) and 43 omnivore (OMN) athletes were evaluated using VO2 max testing on the treadmill, and strength assessment using a dynamometer to determine peak torque for leg extensions. Dietary data were assessed using detailed seven-day food logs. Although total protein intake was lower among vegetarians in comparison to omnivores, protein intake as a function of body mass did not differ by group (1.2 ± 0.3 and 1.4 ± 0.5 g/kg body mass for VEG and OMN respectively, p = 0.220). VO2 max differed for females by diet group (53.0 ± 6.9 and 47.1 ± 8.6 mL/kg/min for VEG and OMN respectively, p vegetarian endurance athletes' cardiorespiratory fitness was greater than that for their omnivorous counterparts, but that peak torque did not differ between diet groups. These data suggest that vegetarian diets do not compromise performance outcomes and may facilitate aerobic capacity in athletes.

Validation of Code ASTEC with LIVE-L1 Experimental Results

International Nuclear Information System (INIS)

Bachrata, Andrea

2008-01-01

The severe accidents with core melting are considered at the design stage of project at Generation 3+ of Nuclear Power Plants (NPP). Moreover, there is an effort to apply the severe accident management to the operated NPP. The one of main goals of severe accidents mitigation is corium localization and stabilization. The two strategies that fulfil this requirement are: the in-vessel retention (e.g. AP-600, AP- 1000) and the ex-vessel retention (e.g. EPR). To study the scenario of in-vessel retention, a large experimental program and the integrated codes have been developed. The LIVE-L1 experimental facility studied the formation of melt pools and the melt accumulation in the lower head using different cooling conditions. Nowadays, a new European computer code ASTEC is being developed jointly in France and Germany. One of the important steps in ASTEC development in the area of in-vessel retention of corium is its validation with LIVE-L1 experimental results. Details of the experiment are reported. Results of the ASTEC (module DIVA) application to the analysis of the test are presented. (author)

Coding training for medical students: How good is diagnoses coding with ICD-10 by novices?

Directory of Open Access Journals (Sweden)

Stausberg, Jürgen

2005-04-01

Full Text Available Teaching of knowledge and competence in documentation and coding is an essential part of medical education. Therefore, coding training had been placed within the course of epidemiology, medical biometry, and medical informatics. From this, we can draw conclusions about the quality of coding by novices. One hundred and eighteen students coded diagnoses from 15 nephrological cases in homework. In addition to interrater reliability, validity was calculated by comparison with a reference coding. On the level of terminal codes, 59.3% of the students' results were correct. The completeness was calculated as 58.0%. The results on the chapter level increased up to 91.5% and 87.7% respectively. For the calculation of reliability a new, simple measure was developed that leads to values of 0.46 on the level of terminal codes and 0.87 on the chapter level for interrater reliability. The figures of concordance with the reference coding are quite similar. In contrary, routine data show considerably lower results with 0.34 and 0.63 respectively. Interrater reliability and validity of coding by novices is as good as coding by experts. The missing advantage of experts could be explained by the workload of documentation and a negative attitude to coding on the one hand. On the other hand, coding in a DRG-system is handicapped by a large number of detailed coding rules, which do not end in uniform results but rather lead to wrong and random codes. Anyway, students left the course well prepared for coding.

Experimental validation for combustion analysis of GOTHIC code in 2-dimensional combustion chamber

International Nuclear Information System (INIS)

Lee, J. W.; Yang, S. Y.; Park, K. C.; Jung, S. H.

2002-01-01

In this study, the prediction capability of GOTHIC code for hydrogen combustion phenomena was validated with the results of two-dimensional premixed hydrogen combustion experiment executed by Seoul National University. The experimental chamber has about 24 liter free volume (1x0.024x1 m 3 ) and 2-dimensional rectangular shape. The test were preformed with 10% hydrogen/air gas mixture and conducted with combination of two igniter positions (top center, top corner) and two boundary conditions (bottom full open, bottom right half open). Using the lumped parameter and mechanistic combustion model in GOTHIC code, the SNU experiments were simulated under the same conditions. The GOTHIC code prediction of the hydrogen combustion phenomena did not compare well with the experimental results. In case of lumped parameter simulation, the combustion time was predicted appropriately. But any other local information related combustion phenomena could not be obtained. In case of mechanistic combustion analysis, the physical combustion phenomena of gas mixture were not matched experimental ones. In boundary open cases, the GOTHIC predicted very long combustion time and the flame front propagation could not simulate appropriately. Though GOTHIC showed flame propagation phenomenon in adiabatic calculation, the induction time of combustion was still very long compare with experimental results. Also, it was found that the combustion model of GOTHIC code had some weak points in low concentration of hydrogen combustion simulation

Linear codes associated to determinantal varieties

DEFF Research Database (Denmark)

Beelen, Peter; Ghorpade, Sudhir R.; Hasan, Sartaj Ul

2015-01-01

We consider a class of linear codes associated to projective algebraic varieties defined by the vanishing of minors of a fixed size of a generic matrix. It is seen that the resulting code has only a small number of distinct weights. The case of varieties defined by the vanishing of 2×2 minors is ...

Validations and applications of the FEAST code

Energy Technology Data Exchange (ETDEWEB)

Xu, Z.; Tayal, M.; Lau, J.H.; Evinou, D. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Jun, J.S. [Korea Atomic Energy Research Inst. (Korea, Republic of)

1999-07-01

The FEAST (Finite Element Analysis for STresses) code is part of a suite of computer codes that are used to assess the structural integrity of CANDu fuel elements and bundles. A detailed validation of the FEAST code was recently performed. The FEAST calculations are in good agreement with a variety of analytical solutions (18 cases) for stresses, strains and displacements. This consistency shows that the FEAST code correctly incorporates the fundamentals of stress analysis. Further, the calculations of the FEAST code match the variations in axial and hoop strain profiles, measured by strain gauges near the sheath-endcap weld during an out-reactor compression test. The code calculations are also consistent with photoelastic measurements in simulated endcaps. (author)

Validations and applications of the FEAST code

International Nuclear Information System (INIS)

Xu, Z.; Tayal, M.; Lau, J.H.; Evinou, D.; Jun, J.S.

1999-01-01

The FEAST (Finite Element Analysis for STresses) code is part of a suite of computer codes that are used to assess the structural integrity of CANDu fuel elements and bundles. A detailed validation of the FEAST code was recently performed. The FEAST calculations are in good agreement with a variety of analytical solutions (18 cases) for stresses, strains and displacements. This consistency shows that the FEAST code correctly incorporates the fundamentals of stress analysis. Further, the calculations of the FEAST code match the variations in axial and hoop strain profiles, measured by strain gauges near the sheath-endcap weld during an out-reactor compression test. The code calculations are also consistent with photoelastic measurements in simulated endcaps. (author)

Optimal codes as Tanner codes with cyclic component codes

DEFF Research Database (Denmark)

Høholdt, Tom; Pinero, Fernando; Zeng, Peng

2014-01-01

In this article we study a class of graph codes with cyclic code component codes as affine variety codes. Within this class of Tanner codes we find some optimal binary codes. We use a particular subgraph of the point-line incidence plane of A(2,q) as the Tanner graph, and we are able to describe ...

An IBM-1620 code for calculaton of isotopic composition of irradiated uranium (ISOCOM-1)

International Nuclear Information System (INIS)

Soliman, R.H.; Karchava, G.; Hamouda, I.

1974-01-01

The present work gives a description of an IBM-1620 code to calculate the isotopic composition during the irradiation of a nuclear fuel, which initially consists of 235 U and 238 U. The numerical results of test calculations as well as the ET-RR-1 reactor calculations are presented. The code is in operation since 1968

CH-TRU Content Codes (CH-TRUCON)

International Nuclear Information System (INIS)

2005-01-01

The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes 'shipping categories' that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the 'General Case,' which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for 'Close-Proximity Shipments' (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for 'Controlled Shipments

CH-TRU Content Codes (CH-TRUCON)

Energy Technology Data Exchange (ETDEWEB)

Washington TRU Solutions LLC

2005-10-15

The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

Integrated intra-subassembly treatment in the SASSYS-1 LMR systems analysis code

Energy Technology Data Exchange (ETDEWEB)

Dunn, F.

1992-09-01

This report discusses a hot channel treatment which has been added to the SASSYS-1 LMR systems analysis code by providing for a multiple pin treatment of each of one or more subassemblies. This is an explicit calculation of intra-subassembly effects, not a hot-channel adjustment to a calculated average channel. Thus, the code can account for effects such as transient flow redistribution, both within a subassembly and among subassemblies. The code now provides a total integrated thermal hydraulic treatment including a multiple pin treatment within subassemblies, a multi-channel treatment of the whole core, and models for the primary coolant loops, the intermediate coolant loops, the steam generators, and the balance of plant. Currently the multiple-pin option is only implemented for single-phase calculations. It is not applicable after the onset of boiling or pin disruption. The new multiple pin treatment is being verified with detailed temperature data from instrumented subassemblies in EBR-II, both steady-state and transient, with special emphasis on passive safety tests such as SHRT-45. For the SHRT-45 test, excellent agreement is obtained between code predictions and experimental measurements of coolant temperatures.

Integrated intra-subassembly treatment in the SASSYS-1 LMR systems analysis code

Energy Technology Data Exchange (ETDEWEB)

Dunn, F.

1992-01-01

This report discusses a hot channel treatment which has been added to the SASSYS-1 LMR systems analysis code by providing for a multiple pin treatment of each of one or more subassemblies. This is an explicit calculation of intra-subassembly effects, not a hot-channel adjustment to a calculated average channel. Thus, the code can account for effects such as transient flow redistribution, both within a subassembly and among subassemblies. The code now provides a total integrated thermal hydraulic treatment including a multiple pin treatment within subassemblies, a multi-channel treatment of the whole core, and models for the primary coolant loops, the intermediate coolant loops, the steam generators, and the balance of plant. Currently the multiple-pin option is only implemented for single-phase calculations. It is not applicable after the onset of boiling or pin disruption. The new multiple pin treatment is being verified with detailed temperature data from instrumented subassemblies in EBR-II, both steady-state and transient, with special emphasis on passive safety tests such as SHRT-45. For the SHRT-45 test, excellent agreement is obtained between code predictions and experimental measurements of coolant temperatures.

NARMER-1: a photon point-kernel code with build-up factors

Science.gov (United States)

Visonneau, Thierry; Pangault, Laurence; Malouch, Fadhel; Malvagi, Fausto; Dolci, Florence

2017-09-01

This paper presents an overview of NARMER-1, the new generation of photon point-kernel code developed by the Reactor Studies and Applied Mathematics Unit (SERMA) at CEA Saclay Center. After a short introduction giving some history points and the current context of development of the code, the paper exposes the principles implemented in the calculation, the physical quantities computed and surveys the generic features: programming language, computer platforms, geometry package, sources description, etc. Moreover, specific and recent features are also detailed: exclusion sphere, tetrahedral meshes, parallel operations. Then some points about verification and validation are presented. Finally we present some tools that can help the user for operations like visualization and pre-treatment.

Development of 2-d cfd code

International Nuclear Information System (INIS)

Mirza, S.A.

1999-01-01

In the present study, a two-dimensional computer code has been developed in FORTRAN using CFD technique, which is basically a numerical scheme. This computer code solves the Navier Stokes equations and continuity equation to find out the velocity and pressure fields within a given domain. This analysis has been done for the developed within a square cavity driven by the upper wall which has become a bench mark for testing and comparing the newly developed numerical schemes. Before to handle this task, different one-dimensional cases have been studied by CFD technique and their FORTRAN programs written. The cases studied are Couette flow, Poiseuille flow with and without using symmetric boundary condition. Finally a comparison between CFD results and analytical results has also been made. For the cavity flow the results from the developed code have been obtained for different Reynolds numbers which are finally presented in the form of velocity vectors. The comparison of the developed code results have been made with the results obtained from the share ware version of a commercially available code for Reynolds number of 10.0. The disagreement in the results quantitatively and qualitatively at some grid points of the calculation domain have been discussed and future recommendations in this regard have also been made. (author)

Improving coding accuracy in an academic practice.

Science.gov (United States)

Nguyen, Dana; O'Mara, Heather; Powell, Robert

2017-01-01

Practice management has become an increasingly important component of graduate medical education. This applies to every practice environment; private, academic, and military. One of the most critical aspects of practice management is documentation and coding for physician services, as they directly affect the financial success of any practice. Our quality improvement project aimed to implement a new and innovative method for teaching billing and coding in a longitudinal fashion in a family medicine residency. We hypothesized that implementation of a new teaching strategy would increase coding accuracy rates among residents and faculty. Design: single group, pretest-posttest. military family medicine residency clinic. Study populations: 7 faculty physicians and 18 resident physicians participated as learners in the project. Educational intervention: monthly structured coding learning sessions in the academic curriculum that involved learner-presented cases, small group case review, and large group discussion. overall coding accuracy (compliance) percentage and coding accuracy per year group for the subjects that were able to participate longitudinally. Statistical tests used: average coding accuracy for population; paired t test to assess improvement between 2 intervention periods, both aggregate and by year group. Overall coding accuracy rates remained stable over the course of time regardless of the modality of the educational intervention. A paired t test was conducted to compare coding accuracy rates at baseline (mean (M)=26.4%, SD=10%) to accuracy rates after all educational interventions were complete (M=26.8%, SD=12%); t24=-0.127, P=.90. Didactic teaching and small group discussion sessions did not improve overall coding accuracy in a residency practice. Future interventions could focus on educating providers at the individual level.

Game Theoretic Analysis of Cooperative Message Forwarding in Opportunistic Mobile Networks.

Science.gov (United States)

Pal, Sujata; Saha, Barun Kumar; Misra, Sudip

2017-12-01

In cooperative communication, a set of players forming a coalition ensures communal behavior among themselves by helping one another in message forwarding. Opportunistic mobile networks (OMNs) require multihop communications for transferring messages from the source to the destination nodes. However, noncooperative nodes only forward their own messages to others, and drop others' messages upon receiving them. So, the message delivery overhead increases in OMN. For minimizing the overhead and maximizing the delivery rate, we propose two coalition-based cooperative schemes: 1) simple coalition formation (SCF) and 2) overlapping coalition formation (OCF) game. In SCF, we consider the presence of a central information center, whereas OCF is a fully distributed scheme. In SCF, coalitions are disjoint, whereas in OCF, a node may be the member of multiple coalitions at the same time. All nodes in a coalition help each other cooperatively by forwarding group messages to the intermediate or destination nodes. The goal of the nodes is to achieve high success rate in delivering messages. The proposed SCF scheme is cohesive, in which disjoint coalitions always combine to form grand coalition. In OCF, a node reaches a stable grand coalition when all the nodes of the OMN are members of overlapping coalition of the node. No node gains by deviating from the grand coalition in SCF and OCF. Simulation results based on synthetic mobility model and real-life traces show that the message delivery ratio of OMNs increase by up to 67%, as compared to the noncooperative scenario. Moreover, the message overhead ratio using the proposed coalition-based schemes reduces by up to about (1/3)rd of that of the noncooperative communication scheme.

Code Switching in the Classroom: A Case Study of Economics and Management Students at the University of Sfax, Tunisia

Science.gov (United States)

Bach Baoueb, Sallouha Lamia; Toumi, Naouel

2012-01-01

This case study explores the motivations for code switching (CS) in the interactions of Tunisian students at the faculty of Economics and Management in Sfax, Tunisia. The study focuses on students' (EMSs) classroom conversations and out-of-classroom peer interactions. The analysis of the social motivations of EMSs' CS behaviour shows that…

Status of the MELTSPREAD-1 computer code for the analysis of transient spreading of core debris melts

International Nuclear Information System (INIS)

Farmer, M.T.; Sienicki, J.J.; Spencer, B.W.; Chu, C.C.

1992-01-01

A transient, one dimensional, finite difference computer code (MELTSPREAD-1) has been developed to predict spreading behavior of high temperature melts flowing over concrete and/or steel surfaces submerged in water, or without the effects of water if the surface is initially dry. This paper provides a summary overview of models and correlations currently implemented in the code, code validation activities completed thus far, LWR spreading-related safety issues for which the code has been applied, and the status of documentation for the code

An Efficient Construction of Self-Dual Codes

OpenAIRE

Lee, Yoonjin; Kim, Jon-Lark

2012-01-01

We complete the building-up construction for self-dual codes by resolving the open cases over $GF(q)$ with $q \\equiv 3 \\pmod 4$, and over $\\Z_{p^m}$ and Galois rings $\\GR(p^m,r)$ with an odd prime $p$ satisfying $p \\equiv 3 \\pmod 4$ with $r$ odd. We also extend the building-up construction for self-dual codes to finite chain rings. Our building-up construction produces many new interesting self-dual codes. In particular, we construct 945 new extremal self-dual ternary $[32,16,9]$ codes, each ...

Simulation of decay heat removal by natural convection in a pool type fast reactor model-ramona-with coupled 1D/2D thermal hydraulic code system

Energy Technology Data Exchange (ETDEWEB)

Kasinathan, N.; Rajakumar, A.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

1995-09-01

Post shutdown decay heat removal is an important safety requirement in any nuclear system. In order to improve the reliability of this function, Liquid metal (sodium) cooled fast breeder reactors (LMFBR) are equipped with redundant hot pool dipped immersion coolers connected to natural draught air cooled heat exchangers through intermediate sodium circuits. During decay heat removal, flow through the core, immersion cooler primary side and in the intermediate sodium circuits are also through natural convection. In order to establish the viability and validate computer codes used in making predictions, a 1:20 scale experimental model called RAMONA with water as coolant has been built and experimental simulation of decay heat removal situation has been performed at KfK Karlsruhe. Results of two such experiments have been compiled and published as benchmarks. This paper brings out the results of the numerical simulation of one of the benchmark case through a 1D/2D coupled code system, DHDYN-1D/THYC-2D and the salient features of the comparisons. Brief description of the formulations of the codes are also included.

Evaluation of SPACE code for simulation of inadvertent opening of spray valve in Shin Kori unit 1

International Nuclear Information System (INIS)