Comparison of different methods used in integral codes to model coagulation of aerosols

Science.gov (United States)

Beketov, A. I.; Sorokin, A. A.; Alipchenkov, V. M.; Mosunova, N. A.

2013-09-01

The methods for calculating coagulation of particles in the carrying phase that are used in the integral codes SOCRAT, ASTEC, and MELCOR, as well as the Hounslow and Jacobson methods used to model aerosol processes in the chemical industry and in atmospheric investigations are compared on test problems and against experimental results in terms of their effectiveness and accuracy. It is shown that all methods are characterized by a significant error in modeling the distribution function for micrometer particles if calculations are performed using rather "coarse" spectra of particle sizes, namely, when the ratio of the volumes of particles from neighboring fractions is equal to or greater than two. With reference to the problems considered, the Hounslow method and the method applied in the aerosol module used in the ASTEC code are the most efficient ones for carrying out calculations.

Integrated code development for studying laser driven plasmas

Energy Technology Data Exchange (ETDEWEB)

Takabe, Hideaki; Nagatomo, Hideo; Sunahara, Atsusi; Ohnishi, Naofumi; Naruo, Syuji; Mima, Kunioki [Osaka Univ., Suita (Japan). Inst. of Laser Engineering

1998-03-01

Present status and plan for developing an integrated implosion code are briefly explained by focusing on motivation, numerical scheme and issues to be developed more. Highly nonlinear stage of Rayleigh-Taylor instability of ablation front by laser irradiation has been simulated so as to be compared with model experiments. Improvement in transport and rezoning/remapping algorithms in ILESTA code is described. (author)

CBP Phase I Code Integration

International Nuclear Information System (INIS)

Smith, F.; Brown, K.; Flach, G.; Sarkar, S.

2011-01-01

The goal of the Cementitious Barriers Partnership (CBP) is to develop a reasonable and credible set of software tools to predict the structural, hydraulic, and chemical performance of cement barriers used in nuclear applications over extended time frames (greater than 100 years for operating facilities and greater than 1000 years for waste management). The simulation tools will be used to evaluate and predict the behavior of cementitious barriers used in near surface engineered waste disposal systems including waste forms, containment structures, entombments, and environmental remediation. These cementitious materials are exposed to dynamic environmental conditions that cause changes in material properties via (i) aging, (ii) chloride attack, (iii) sulfate attack, (iv) carbonation, (v) oxidation, and (vi) primary constituent leaching. A set of state-of-the-art software tools has been selected as a starting point to capture these important aging and degradation phenomena. Integration of existing software developed by the CBP partner organizations was determined to be the quickest method of meeting the CBP goal of providing a computational tool that improves the prediction of the long-term behavior of cementitious materials. These partner codes were selected based on their maturity and ability to address the problems outlined above. The GoldSim Monte Carlo simulation program (GTG 2010a, GTG 2010b) was chosen as the code integration platform (Brown and Flach 2009b). GoldSim (current Version 10.5) is a Windows based graphical object-oriented computer program that provides a flexible environment for model development (Brown and Flach 2009b). The linking of GoldSim to external codes has previously been successfully demonstrated (Eary 2007, Mattie et al. 2007). GoldSim is capable of performing deterministic and probabilistic simulations and of modeling radioactive decay and constituent transport. As part of the CBP project, a general Dynamic Link Library (DLL) interface

CBP PHASE I CODE INTEGRATION

Energy Technology Data Exchange (ETDEWEB)

Smith, F.; Brown, K.; Flach, G.; Sarkar, S.

2011-09-30

The goal of the Cementitious Barriers Partnership (CBP) is to develop a reasonable and credible set of software tools to predict the structural, hydraulic, and chemical performance of cement barriers used in nuclear applications over extended time frames (greater than 100 years for operating facilities and greater than 1000 years for waste management). The simulation tools will be used to evaluate and predict the behavior of cementitious barriers used in near surface engineered waste disposal systems including waste forms, containment structures, entombments, and environmental remediation. These cementitious materials are exposed to dynamic environmental conditions that cause changes in material properties via (i) aging, (ii) chloride attack, (iii) sulfate attack, (iv) carbonation, (v) oxidation, and (vi) primary constituent leaching. A set of state-of-the-art software tools has been selected as a starting point to capture these important aging and degradation phenomena. Integration of existing software developed by the CBP partner organizations was determined to be the quickest method of meeting the CBP goal of providing a computational tool that improves the prediction of the long-term behavior of cementitious materials. These partner codes were selected based on their maturity and ability to address the problems outlined above. The GoldSim Monte Carlo simulation program (GTG 2010a, GTG 2010b) was chosen as the code integration platform (Brown & Flach 2009b). GoldSim (current Version 10.5) is a Windows based graphical object-oriented computer program that provides a flexible environment for model development (Brown & Flach 2009b). The linking of GoldSim to external codes has previously been successfully demonstrated (Eary 2007, Mattie et al. 2007). GoldSim is capable of performing deterministic and probabilistic simulations and of modeling radioactive decay and constituent transport. As part of the CBP project, a general Dynamic Link Library (DLL) interface was

Towards Product Lining Model-Driven Development Code Generators

OpenAIRE

Roth, Alexander; Rumpe, Bernhard

2015-01-01

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

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

A Perceptual Model for Sinusoidal Audio Coding Based on Spectral Integration

NARCIS (Netherlands)

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

2005-01-01

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

A perceptual model for sinusoidal audio coding based on spectral integration

NARCIS (Netherlands)

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

2005-01-01

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

Behaviors of impurity in ITER and DEMOs using BALDUR integrated predictive modeling code

International Nuclear Information System (INIS)

Onjun, Thawatchai; Buangam, Wannapa; Wisitsorasak, Apiwat

2015-01-01

The behaviors of impurity are investigated using self-consistent modeling of 1.5D BALDUR integrated predictive modeling code, in which theory-based models are used for both core and edge region. In these simulations, a combination of NCLASS neoclassical transport and Multi-mode (MMM95) anomalous transport model is used to compute a core transport. The boundary is taken to be at the top of the pedestal, where the pedestal values are described using a theory-based pedestal model. This pedestal temperature model is based on a combination of magnetic and flow shear stabilization pedestal width scaling and an infinite-n ballooning pressure gradient model. The time evolution of plasma current, temperature and density profiles is carried out for ITER and DEMOs plasmas. As a result, the impurity behaviors such as impurity accumulation and impurity transport can be investigated. (author)

MARS CODE MANUAL VOLUME V: Models and Correlations

International Nuclear Information System (INIS)

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

2010-02-01

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

The Effectiveness of Business Codes: A Critical Examination of Existing Studies and the Development of an Integrated Research Model

OpenAIRE

Kaptein, S.P.; Schwartz, M.S.

2007-01-01

textabstractBusiness codes are a widely used management instrument. Research into the effectiveness of business codes has, however, produced conflicting results. The main reasons for the divergent findings are: varying definitions of key terms; deficiencies in the empirical data and methodologies used; and a lack of theory. In this paper, we propose an integrated research model and suggest directions for future research.

ASTEC V2 severe accident integral code main features, current V2.0 modelling status, perspectives

International Nuclear Information System (INIS)

Chatelard, P.; Reinke, N.; Arndt, S.; Belon, S.; Cantrel, L.; Carenini, L.; Chevalier-Jabet, K.; Cousin, F.; Eckel, J.; Jacq, F.; Marchetto, C.; Mun, C.; Piar, L.

2014-01-01

The severe accident integral code ASTEC, jointly developed since almost 20 years by IRSN and GRS, simulates the behaviour of a whole nuclear power plant under severe accident conditions, including severe accident management by engineering systems and procedures. Since 2004, the ASTEC code is progressively becoming the reference European severe accident integral code through in particular the intensification of research activities carried out in the frame of the SARNET European network of excellence. The first version of the new series ASTEC V2 was released in 2009 to about 30 organizations worldwide and in particular to SARNET partners. With respect to the previous V1 series, this new V2 series includes advanced core degradation models (issued from the ICARE2 IRSN mechanistic code) and necessary extensions to be applicable to Gen. III reactor designs, notably a description of the core catcher component to simulate severe accidents transients applied to the EPR reactor. Besides these two key-evolutions, most of the other physical modules have also been improved and ASTEC V2 is now coupled to the SUNSET statistical tool to make easier the uncertainty and sensitivity analyses. The ASTEC models are today at the state of the art (in particular fission product models with respect to source term evaluation), except for quenching of a severely damage core. Beyond the need to develop an adequate model for the reflooding of a degraded core, the main other mean-term objectives are to further progress on the on-going extension of the scope of application to BWR and CANDU reactors, to spent fuel pool accidents as well as to accidents in both the ITER Fusion facility and Gen. IV reactors (in priority on sodium-cooled fast reactors) while making ASTEC evolving towards a severe accident simulator constitutes the main long-term objective. This paper presents the status of the ASTEC V2 versions, focussing on the description of V2.0 models for water-cooled nuclear plants

Integrated Fuel-Coolant Interaction (IFCI 6.0) code

International Nuclear Information System (INIS)

Davis, F.J.; Young, M.F.

1994-04-01

The integrated Fuel-Coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, four-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a product of the effort to generate a stand-alone version of IFCI, IFCI 6.0. The User's Manual describes in detail the hydrodynamic method and physical models used in IFCI 6.0. Appendix A is an input manual, provided for the creation of working decks

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-01

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

Assessment of capability for modeling the core degradation in 2D geometry with ASTEC V2 integral code for VVER type of reactor

International Nuclear Information System (INIS)

Dimov, D.

2011-01-01

The ASTEC code is progressively becoming the reference European severe accident integral code through in particular the intensification of research activities carried out since 2004. The purpose of this analysis is to assess ASTEC code modelling of main phenomena arising during hypothetical severe accidents and particularly in-vessel degradation in 2D geometry. The investigation covers both early and late phase of degradation of reactor core as well as determination of corium which will enter the reactor cavity. The initial event is station back-out. In order to receive severe accident condition, failure of all active component of emergency core cooling system is apply. The analysis is focus on ICARE module of ASTEC code and particularly on so call MAGMA model. The aim of study is to determine the capability of the integral code to simulate core degradation and to determine the corium composition entering the reactor cavity. (author)

Separations and safeguards model integration.

Energy Technology Data Exchange (ETDEWEB)

Cipiti, Benjamin B.; Zinaman, Owen

2010-09-01

Research and development of advanced reprocessing plant designs can greatly benefit from the development of a reprocessing plant model capable of transient solvent extraction chemistry. This type of model can be used to optimize the operations of a plant as well as the designs for safeguards, security, and safety. Previous work has integrated a transient solvent extraction simulation module, based on the Solvent Extraction Process Having Interaction Solutes (SEPHIS) code developed at Oak Ridge National Laboratory, with the Separations and Safeguards Performance Model (SSPM) developed at Sandia National Laboratory, as a first step toward creating a more versatile design and evaluation tool. The goal of this work was to strengthen the integration by linking more variables between the two codes. The results from this integrated model show expected operational performance through plant transients. Additionally, ORIGEN source term files were integrated into the SSPM to provide concentrations, radioactivity, neutron emission rate, and thermal power data for various spent fuels. This data was used to generate measurement blocks that can determine the radioactivity, neutron emission rate, or thermal power of any stream or vessel in the plant model. This work examined how the code could be expanded to integrate other separation steps and benchmark the results to other data. Recommendations for future work will be presented.

Development of an integrated thermal-hydraulics capability incorporating RELAP5 and PANTHER neutronics code

Energy Technology Data Exchange (ETDEWEB)

Page, R.; Jones, J.R.

1997-07-01

Ensuring that safety analysis needs are met in the future is likely to lead to the development of new codes and the further development of existing codes. It is therefore advantageous to define standards for data interfaces and to develop software interfacing techniques which can readily accommodate changes when they are made. Defining interface standards is beneficial but is necessarily restricted in application if future requirements are not known in detail. Code interfacing methods are of particular relevance with the move towards automatic grid frequency response operation where the integration of plant dynamic, core follow and fault study calculation tools is considered advantageous. This paper describes the background and features of a new code TALINK (Transient Analysis code LINKage program) used to provide a flexible interface to link the RELAP5 thermal hydraulics code with the PANTHER neutron kinetics and the SIBDYM whole plant dynamic modelling codes used by Nuclear Electric. The complete package enables the codes to be executed in parallel and provides an integrated whole plant thermal-hydraulics and neutron kinetics model. In addition the paper discusses the capabilities and pedigree of the component codes used to form the integrated transient analysis package and the details of the calculation of a postulated Sizewell `B` Loss of offsite power fault transient.

Development of an integrated thermal-hydraulics capability incorporating RELAP5 and PANTHER neutronics code

International Nuclear Information System (INIS)

Page, R.; Jones, J.R.

1997-01-01

Ensuring that safety analysis needs are met in the future is likely to lead to the development of new codes and the further development of existing codes. It is therefore advantageous to define standards for data interfaces and to develop software interfacing techniques which can readily accommodate changes when they are made. Defining interface standards is beneficial but is necessarily restricted in application if future requirements are not known in detail. Code interfacing methods are of particular relevance with the move towards automatic grid frequency response operation where the integration of plant dynamic, core follow and fault study calculation tools is considered advantageous. This paper describes the background and features of a new code TALINK (Transient Analysis code LINKage program) used to provide a flexible interface to link the RELAP5 thermal hydraulics code with the PANTHER neutron kinetics and the SIBDYM whole plant dynamic modelling codes used by Nuclear Electric. The complete package enables the codes to be executed in parallel and provides an integrated whole plant thermal-hydraulics and neutron kinetics model. In addition the paper discusses the capabilities and pedigree of the component codes used to form the integrated transient analysis package and the details of the calculation of a postulated Sizewell 'B' Loss of offsite power fault transient

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

International Nuclear Information System (INIS)

Cohen, R H; Yang, T Y Brian.

1999-01-01

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

Overview of Grid Codes for Photovoltaic Integration

DEFF Research Database (Denmark)

Zheng, Qianwei; Li, Jiaming; Ai, Xiaomeng

2017-01-01

The increasing grid-connected photovoltaic (PV) power stations might threaten the safety and stability of power system. Therefore, the grid code is developed for PV power stations to ensure the security of PV integrated power systems. In this paper, requirements for PV power integration in differ......The increasing grid-connected photovoltaic (PV) power stations might threaten the safety and stability of power system. Therefore, the grid code is developed for PV power stations to ensure the security of PV integrated power systems. In this paper, requirements for PV power integration...

Overview of Grid Codes for Photovoltaic Integration

DEFF Research Database (Denmark)

Zheng, Qianwei; Li, Jiaming; Ai, Xiaomeng

2017-01-01

The increasing grid-connected photovoltaic (PV) power stations might threaten the safety and stability of power system. Therefore, the grid code is developed for PV power stations to ensure the security of PV integrated power systems. In this paper, requirements for PV power integration in differ...... in different grid codes are first investigated. On this basis, the future advocacy is concluded. Finally, several evaluation indices are proposed to quantify the grid code compliance so that the system operators can validate all these requirements by simulation....

Implementation of an enlarged model of the safety valves and relief in the plant integral model for the code RELAP/SCDAPSIM

International Nuclear Information System (INIS)

Amador G, R.; Ortiz V, J.; Castillo D, R.; Hernandez L, E. J.; Galeana R, J. C.; Gutierrez, V. H.

2013-10-01

The present work refers to the implementation of a new model on the logic of the safety valves and relief in the integral model of the Nuclear Power Plant of Laguna Verde of the thermal-hydraulic compute code RELAP/SCDAPSIM Mod. 3.4. The new model was developed with the compute package SIMULINK-MATLAB and contemplates all the operation options of the safety valves and relief, besides including the availability options of the valves in all the operation ways and of blockage in the ways of relief and low-low. The implementation means the elimination of the old model of the safety valves and to analyze the group of logical variables, of discharge and available control systems to associate them to the model of package SIMULINK-MATLAB. The implementation has been practically transparent and 27 cases corresponding to a turbine discharge were analyzed with the code RELAP/SCDAPSIM Mod. 3.4. The results were satisfactory. (Author)

Model-integrating software components engineering flexible software systems

CERN Document Server

Derakhshanmanesh, Mahdi

2015-01-01

In his study, Mahdi Derakhshanmanesh builds on the state of the art in modeling by proposing to integrate models into running software on the component-level without translating them to code. Such so-called model-integrating software exploits all advantages of models: models implicitly support a good separation of concerns, they are self-documenting and thus improve understandability and maintainability and in contrast to model-driven approaches there is no synchronization problem anymore between the models and the code generated from them. Using model-integrating components, software will be

Development of an integral computer code for simulation of heat exchangers

International Nuclear Information System (INIS)

Horvat, A.; Catton, I.

2001-01-01

Heat exchangers are one of the basic installations in power and process industries. The present guidelines provide an ad-hoc solution to certain design problems. A unified approach based on simultaneous modeling of thermal-hydraulics and structural behavior does not exist. The present paper describes the development of integral numerical code for simulation of heat exchangers. The code is based on Volume Averaging Technique (VAT) for porous media flow modeling. The calculated values of the whole-section drag and heat transfer coefficients show an excellent agreement with already published values. The matching results prove the correctness of the selected approach and verify the developed numerical code used for this calculation.(author)

Fission-product release modelling in the ASTEC integral code: the status of the ELSA module

International Nuclear Information System (INIS)

Plumecocq, W.; Kissane, M.P.; Manenc, H.; Giordano, P.

2003-01-01

Safety assessment of water-cooled nuclear reactors encompasses potential severe accidents where, in particular, the release of fission products (FPs) and actinides into the reactor coolant system (RCS) is evaluated. The ELSA module is used in the ASTEC integral code to model all releases into the RCS. A wide variety of experiments is used for validation: small-scale CRL, ORNL and VERCORS tests; large-scale Phebus-FP tests; etc. Being a tool that covers intact fuel and degraded states, ELSA is being improved maximizing the use of information from degradation modelling. Short-term improvements will include some treatment of initial FP release due to intergranular inventories and implementing models for release of additional structural materials (Sn, Fe, etc.). (author)

iTOUGH2-IFC: An Integrated Flow Code in Support of Nagra's Probabilistic Safety Assessment: User's Guide and Model Description

International Nuclear Information System (INIS)

Finsterle, Stefan A.

2009-01-01

This document describes the development and use of the Integrated Flow Code (IFC), a numerical code and related model to be used for the simulation of time-dependent, two-phase flow in the near field and geosphere of a gas-generating nuclear waste repository system located in an initially fully water-saturated claystone (Opalinus Clay) in Switzerland. The development of the code and model was supported by the Swiss National Cooperative for the Disposal of Radioactive Waste (Nagra), Wettingen, Switzerland. Gas generation (mainly H 2 , but also CH 4 and CO 2 ) may affect repository performance by (1) compromising the engineered barriers through excessive pressure build-up, (2) displacing potentially contaminated pore water, (3) releasing radioactive gases (e.g., those containing 14 C and 3 H), (4) changing hydrogeologic properties of the engineered barrier system and the host rock, and (5) altering the groundwater flow field and thus radionuclide migration paths. The IFC aims at providing water and gas flow fields as the basis for the subsequent radionuclide transport simulations, which are performed by the radionuclide transport code (RTC). The IFC, RTC and a waste-dissolution and near-field transport model (STMAN) are part of the Integrated Radionuclide Release Code (IRRC), which integrates all safety-relevant features, events, and processes (FEPs). The IRRC is embedded into a Probabilistic Safety Assessment (PSA) computational tool that (1) evaluates alternative conceptual models, scenarios, and disruptive events, and (2) performs Monte-Carlo sampling to account for parametric uncertainties. The preliminary probabilistic safety assessment concept and the role of the IFC are visualized in Figure 1. The IFC was developed based on Nagra's PSA concept. Specifically, as many phenomena as possible are to be directly simulated using a (simplified) process model, which is at the core of the IRRC model. Uncertainty evaluation (scenario uncertainty, conceptualization

ASTEC V2 severe accident integral code: Fission product modelling and validation

International Nuclear Information System (INIS)

Cantrel, L.; Cousin, F.; Bosland, L.; Chevalier-Jabet, K.; Marchetto, C.

2014-01-01

One main goal of the severe accident integral code ASTEC V2, jointly developed since almost more than 15 years by IRSN and GRS, is to simulate the overall behaviour of fission products (FP) in a damaged nuclear facility. ASTEC applications are source term determinations, level 2 Probabilistic Safety Assessment (PSA2) studies including the determination of uncertainties, accident management studies and physical analyses of FP experiments to improve the understanding of the phenomenology. ASTEC is a modular code and models of a part of the phenomenology are implemented in each module: the release of FPs and structural materials from degraded fuel in the ELSA module; the transport through the reactor coolant system approximated as a sequence of control volumes in the SOPHAEROS module; and the radiochemistry inside the containment nuclear building in the IODE module. Three other modules, CPA, ISODOP and DOSE, allow respectively computing the deposition rate of aerosols inside the containment, the activities of the isotopes as a function of time, and the gaseous dose rate which is needed to model radiochemistry in the gaseous phase. In ELSA, release models are semi-mechanistic and have been validated for a wide range of experimental data, and noticeably for VERCORS experiments. For SOPHAEROS, the models can be divided into two parts: vapour phase phenomena and aerosol phase phenomena. For IODE, iodine and ruthenium chemistry are modelled based on a semi-mechanistic approach, these FPs can form some volatile species and are particularly important in terms of potential radiological consequences. The models in these 3 modules are based on a wide experimental database, resulting for a large part from international programmes, and they are considered at the state of the art of the R and D knowledge. This paper illustrates some FPs modelling capabilities of ASTEC and computed values are compared to some experimental results, which are parts of the validation matrix

Development of system based code for integrity of FBR. Fundamental probabilistic approach, Part 1: Model calculation of creep-fatigue damage (Research report)

International Nuclear Information System (INIS)

Kawasaki, Nobuchika; Asayama, Tai

2001-09-01

Both reliability and safety have to be further improved for the successful commercialization of FBRs. At the same time, construction and operation costs need to be reduced to a same level of future LWRs. To realize compatibility among reliability, safety and, cost, the Structural Mechanics Research Group in JNC started the development of System Based Code for Integrity of FBR. This code extends the present structural design standard to include the areas of fabrication, installation, plant system design, safety design, operation and maintenance, and so on. A quantitative index is necessary to connect different partial standards in this code. Failure probability is considered as a candidate index. Therefore we decided to make a model calculation using failure probability and judge its applicability. We first investigated other probabilistic standards like ASME Code Case N-578. A probabilistic approach in the structural integrity evaluation was created based on these results, and also an evaluation flow was proposed. According to this flow, a model calculation of creep-fatigue damage was performed. This trial calculation was for a vessel in a sodium-cooled FBR. As the result of this model calculation, a crack initiation probability and a crack penetration probability were found to be effective indices. Last we discussed merits of this System Based Code, which are presented in this report. Furthermore, this report presents future development tasks. (author)

Integrated Fuel-Coolant Interaction (IFCI 7.0) Code User's Manual

International Nuclear Information System (INIS)

Young, Michael F.

1999-01-01

The integrated fuel-coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, three-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a description of IFCI 7.0. The user's manual describes the hydrodynamic method and physical models used in IFCI 7.0. Appendix A is an input manual provided for the creation of working decks

Integrated analysis of core debris interactions and their effects on containment integrity using the CONTAIN computer code

International Nuclear Information System (INIS)

Carroll, D.E.; Bergeron, K.D.; Williams, D.C.; Tills, J.L.; Valdez, G.D.

1987-01-01

The CONTAIN computer code includes a versatile system of phenomenological models for analyzing the physical, chemical and radiological conditions inside the containment building during severe reactor accidents. Important contributors to these conditions are the interactions which may occur between released corium and cavity concrete. The phenomena associated with interactions between ejected corium debris and the containment atmosphere (Direct Containment Heating or DCH) also pose a potential threat to containment integrity. In this paper, we describe recent enhancements of the CONTAIN code which allow an integrated analysis of these effects in the presence of other mitigating or aggravating physical processes. In particular, the recent inclusion of the CORCON and VANESA models is described and a calculation example presented. With this capability CONTAIN can model core-concrete interactions occurring simultaneously in multiple compartments and can couple the aerosols thereby generated to the mechanistic description of all atmospheric aerosol components. Also discussed are some recent results of modeling the phenomena involved in Direct Containment Heating. (orig.)

ANL/CANTIA code for steam generator tube integrity assessment

International Nuclear Information System (INIS)

Revankar, S.T.; Wolf, B.; Majumdar, S.; Riznic, J.R.

2009-01-01

Steam generator (SG) tubes have an important safety role in CANDU type reactors and Pressurized Water Reactors (PWR) because they constitute one of the primary barriers between the radioactive and non-radioactive sides of the nuclear plant. The SG tubes are susceptible to corrosion and damage. A failure of a single steam generator tube, or even a few tubes, would not be a serious safety-related event in a CANDU reactor. The leakage from a ruptured tube is within makeup capacity of the primary heat transport system, so that as long as the operator takes the correct actions, the off-site consequences will be negligible. A sufficient safety margin against tube rupture used to be the basis for a variety of maintenance strategies developed to maintain a suitable level of plant safety and reliability. Several through-wall flaws may remain in operation and potentially contribute to the total primary-to-secondary leak rate. Assessment of the conditional probabilities of tube failures, leak rates, and ultimately risk of exceeding licensing dose limits has been used for steam generator tube fitness-for-service assessment. The advantage of this type of analysis is that it avoids the excessive conservatism typically present in deterministic methodologies. However, it requires considerable effort and expense to develop all of the failure, leakage, probability of detection, and flaw growth distributions and models necessary to obtain meaningful results from a probabilistic model. The Canadian Nuclear Safety Commission (CNSC) recently developed the CANTIA methodology for probabilistic assessment of inspection strategies for steam generator tubes as a direct effect on the probability of tube failure and primary-to-secondary leak rate Recently Argonne National Laboratory has developed tube integrity and leak rate models under Integrated Steam Generator Tube Integrity Program (ISGTIP-2). These models have been incorporated in the ANL/CANTIA code. This paper presents the ANL

MELCOR code modeling for APR1400

Energy Technology Data Exchange (ETDEWEB)

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

2001-11-01

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

Grid Code Requirements for Wind Power Integration

DEFF Research Database (Denmark)

Wu, Qiuwei

2018-01-01

This chapter reviews the grid code requirements for integration of wind power plants (WPPs). The grid codes reviewed are from the UK, Ireland, Germany, Denmark, Spain, Sweden, the USA, and Canada. Transmission system operators (TSOs) around the world have specified requirements for WPPs under...

Gap conductance model validation in the TASS/SMR-S code

International Nuclear Information System (INIS)

Ahn, Sang-Jun; Yang, Soo-Hyung; Chung, Young-Jong; Bae, Kyoo-Hwan; Lee, Won-Jae

2011-01-01

An advanced integral pressurized water reactor, SMART (System-Integrated Modular Advanced ReacTor) has been developed by KAERI (Korea Atomic Energy Research and Institute). The purposes of the SMART are sea water desalination and an electricity generation. For the safety evaluation and performance analysis of the SMART, TASS/SMR-S (Transient And Setpoint Simulation/System-integrated Modular Reactor) code, has been developed. In this paper, the gap conductance model for the calculation of gap conductance has been validated by using another system code, MARS code, and experimental results. In the validation, the behaviors of fuel temperature and gap width are selected as the major parameters. According to the evaluation results, the TASS/SMR-S code predicts well the behaviors of fuel temperatures and gap width variation, compared to the MARS calculation results and experimental data. (author)

IM (Integrity Management) software must show flexibility to local codes

Energy Technology Data Exchange (ETDEWEB)

Brors, Markus [ROSEN Technology and Research Center GmbH (Germany); Diggory, Ian [Macaw Engineering Ltd., Northumberland (United Kingdom)

2009-07-01

There are many internationally recognized codes and standards, such as API 1160 and ASME B31.8S, which help pipeline operators to manage and maintain the integrity of their pipeline networks. However, operators in many countries still use local codes that often reflect the history of pipeline developments in their region and are based on direct experience and research on their pipelines. As pipeline companies come under increasing regulatory and financial pressures to maintain the integrity of their networks, it is important that operators using regional codes are able to benchmark their integrity management schemes against these international standards. Any comprehensive Pipeline Integrity Management System (PIMS) software package should therefore not only incorporate industry standards for pipeline integrity assessment but also be capable of implementing regional codes for comparison purposes. This paper describes the challenges and benefits of incorporating one such set of regional pipeline standards into ROSEN Asset Integrity Management Software (ROAIMS). (author)

Multiphase integral reacting flow computer code (ICOMFLO): User`s guide

Energy Technology Data Exchange (ETDEWEB)

Chang, S.L.; Lottes, S.A.; Petrick, M.

1997-11-01

A copyrighted computational fluid dynamics computer code, ICOMFLO, has been developed for the simulation of multiphase reacting flows. The code solves conservation equations for gaseous species and droplets (or solid particles) of various sizes. General conservation laws, expressed by elliptic type partial differential equations, are used in conjunction with rate equations governing the mass, momentum, enthalpy, species, turbulent kinetic energy, and turbulent dissipation. Associated phenomenological submodels of the code include integral combustion, two parameter turbulence, particle evaporation, and interfacial submodels. A newly developed integral combustion submodel replacing an Arrhenius type differential reaction submodel has been implemented to improve numerical convergence and enhance numerical stability. A two parameter turbulence submodel is modified for both gas and solid phases. An evaporation submodel treats not only droplet evaporation but size dispersion. Interfacial submodels use correlations to model interfacial momentum and energy transfer. The ICOMFLO code solves the governing equations in three steps. First, a staggered grid system is constructed in the flow domain. The staggered grid system defines gas velocity components on the surfaces of a control volume, while the other flow properties are defined at the volume center. A blocked cell technique is used to handle complex geometry. Then, the partial differential equations are integrated over each control volume and transformed into discrete difference equations. Finally, the difference equations are solved iteratively by using a modified SIMPLER algorithm. The results of the solution include gas flow properties (pressure, temperature, density, species concentration, velocity, and turbulence parameters) and particle flow properties (number density, temperature, velocity, and void fraction). The code has been used in many engineering applications, such as coal-fired combustors, air

A new 3-D integral code for computation of accelerator magnets

International Nuclear Information System (INIS)

Turner, L.R.; Kettunen, L.

1991-01-01

For computing accelerator magnets, integral codes have several advantages over finite element codes; far-field boundaries are treated automatically, and computed field in the bore region satisfy Maxwell's equations exactly. A new integral code employing edge elements rather than nodal elements has overcome the difficulties associated with earlier integral codes. By the use of field integrals (potential differences) as solution variables, the number of unknowns is reduced to one less than the number of nodes. Two examples, a hollow iron sphere and the dipole magnet of Advanced Photon Source injector synchrotron, show the capability of the code. The CPU time requirements are comparable to those of three-dimensional (3-D) finite-element codes. Experiments show that in practice it can realize much of the potential CPU time saving that parallel processing makes possible. 8 refs., 4 figs., 1 tab

A Study of Performance in Low-Power Tokamak Reactor with Integrated Predictive Modeling Code

International Nuclear Information System (INIS)

Pianroj, Y.; Onjun, T.; Suwanna, S.; Picha, R.; Poolyarat, N.

2009-07-01

Full text: A fusion hybrid or a small fusion power output with low power tokamak reactor is presented as another useful application of nuclear fusion. Such tokamak can be used for fuel breeding, high-level waste transmutation, hydrogen production at high temperature, and testing of nuclear fusion technology components. In this work, an investigation of the plasma performance in a small fusion power output design is carried out using the BALDUR predictive integrated modeling code. The simulations of the plasma performance in this design are carried out using the empirical-based Mixed Bohm/gyro Bohm (B/gB) model, whereas the pedestal temperature model is based on magnetic and flow shear (δ α ρ ζ 2 ) stabilization pedestal width scaling. The preliminary results using this core transport model show that the central ion and electron temperatures are rather pessimistic. To improve the performance, the optimization approach are carried out by varying some parameters, such as plasma current and power auxiliary heating, which results in some improvement of plasma performance

TASS/SMR Code Topical Report for SMART Plant, Vol. I: Code Structure, System Models, and Solution Methods

Energy Technology Data Exchange (ETDEWEB)

Chung, Young Jong; Kim, Soo Hyoung; Kim, See Darl (and others)

2008-10-15

The TASS/SMR code has been developed with domestic technologies for the safety analysis of the SMART plant which is an integral type pressurized water reactor. It can be applied to the analysis of design basis accidents including non-LOCA (loss of coolant accident) and LOCA of the SMART plant. The TASS/SMR code can be applied to any plant regardless of the structural characteristics of a reactor since the code solves the same governing equations for both the primary and secondary system. The code has been developed to meet the requirements of the safety analysis code. This report describes the overall structure of the TASS/SMR, input processing, and the processes of a steady state and transient calculations. In addition, basic differential equations, finite difference equations, state relationships, and constitutive models are described in the report. First, the conservation equations, a discretization process for numerical analysis, search method for state relationship are described. Then, a core power model, heat transfer models, physical models for various components, and control and trip models are explained.

An Integration of the Restructured Melcor for the Midas Computer Code

International Nuclear Information System (INIS)

Sunhee Park; Dong Ha Kim; Ko-Ryu Kim; Song-Won Cho

2006-01-01

results were compared with the restructured results for each package. Some sequences were calculated such as a steady state and SBO (Station Blackout) accident. The major variables were the same as well as the graph trends. Through out the integrating process, the base was constructed for a code improvement and an addition of new models. The integrating process proposed in this paper will be extended to the T/H and F/P packages for the MIDAS development program. (authors)

Single integrated device for optical CDMA code processing in dual-code environment.

Science.gov (United States)

Huang, Yue-Kai; Glesk, Ivan; Greiner, Christoph M; Iazkov, Dmitri; Mossberg, Thomas W; Wang, Ting; Prucnal, Paul R

2007-06-11

We report on the design, fabrication and performance of a matching integrated optical CDMA encoder-decoder pair based on holographic Bragg reflector technology. Simultaneous encoding/decoding operation of two multiple wavelength-hopping time-spreading codes was successfully demonstrated and shown to support two error-free OCDMA links at OC-24. A double-pass scheme was employed in the devices to enable the use of longer code length.

Current status of the transient integral fuel element performance code URANUS

International Nuclear Information System (INIS)

Preusser, T.; Lassmann, K.

1983-01-01

To investigate the behavior of fuel pins during normal and off-normal operation, the integral fuel rod code URANUS has been extended to include a transient version. The paper describes the current status of the program system including a presentation of newly developed models for hypothetical accident investigation. The main objective of current development work is to improve the modelling of fuel and clad material behavior during fast transients. URANUS allows detailed analysis of experiments until the onset of strong material transport phenomena. Transient fission gas analysis is carried out due to the coupling with a special version of the LANGZEIT-KURZZEIT-code (KfK). Fuel restructuring and grain growth kinetics models have been improved recently to better characterize pre-experimental steady-state operation; transient models are under development. Extensive verification of the new version has been carried out by comparison with analytical solutions, experimental evidence, and code-to-code evaluation studies. URANUS, with all these improvements, has been successfully applied to difficult fast breeder fuel rod analysis including TOP, LOF, TUCOP, local coolant blockage and specific carbide fuel experiments. Objective of further studies is the description of transient PCMI. It is expected that the results of these developments will contribute significantly to the understanding of fuel element structural behavior during severe transients. (orig.)

Status of emergency spray modelling in the integral code ASTEC

International Nuclear Information System (INIS)

Plumecocq, W.; Passalacqua, R.

2001-01-01

Containment spray systems are emergency systems that would be used in very low probability events which may lead to severe accidents in Light Water Reactors. In most cases, the primary function of the spray would be to remove heat and condense steam in order to reduce pressure and temperature in the containment building. Spray would also wash out fission products (aerosols and gaseous species) from the containment atmosphere. The efficiency of the spray system in the containment depressurization as well as in the removal of aerosols, during a severe accident, depends on the evolution of the spray droplet size distribution with the height in the containment, due to kinetic and thermal relaxation, gravitational agglomeration and mass transfer with the gas. A model has been developed taking into account all of these phenomena. This model has been implemented in the ASTEC code with a validation of the droplets relaxation against the CARAIDAS experiment (IPSN). Applications of this modelling to a PWR 900, during a severe accident, with special emphasis on the effect of spray on containment hydrogen distribution have been performed in multi-compartment configuration with the ASTEC V0.3 code. (author)

Integrated Fuel-Coolant Interaction (IFCI 7.0) Code User's Manual

Energy Technology Data Exchange (ETDEWEB)

Young, Michael F.

1999-05-01

The integrated fuel-coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, three-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a description of IFCI 7.0. The user's manual describes the hydrodynamic method and physical models used in IFCI 7.0. Appendix A is an input manual provided for the creation of working decks.

Development of integrated computer code for analysis of risk reduction strategy

International Nuclear Information System (INIS)

Kim, Dong Ha; Kim, See Darl; Kim, Hee Dong

2002-05-01

The development of the MIDAS/TH integrated severe accident code was performed in three main areas: 1) addition of new models derived from the national experimental programs and models for APR-1400 Korea next generation reactor, 2) improvement of the existing models using the recently available results, and 3) code restructuring for user friendliness. The unique MIDAS/TH models include: 1) a kinetics module for core power calculation during ATWS, 2) a gap cooling module between the molten corium pool and the reactor vessel wall, 3) a penetration tube failure module, 4) a PAR analysis module, and 5) a look-up table for the pressure and dynamic load during steam explosion. The improved models include: 1) a debris dispersal module considering the cavity geometry during DCH, 2) hydrogen burn and deflagration-to-detonation transition criteria, 3) a peak pressure estimation module for hydrogen detonation, and 4) the heat transfer module between the molten corium pool and the overlying water. The sparger and the ex-vessel heat transfer module were assessed. To enhance user friendliness, code restructuring was performed. In addition, a sample of severe accident analysis results was organized under the preliminary database structure

INTRA/Mod3.2. Manual and Code Description. Volume I - Physical Modelling

International Nuclear Information System (INIS)

Andersson, Jenny; Edlund, O.; Hermann, J.; Johansson, Lise-Lotte

1999-01-01

The INTRA Manual consists of two volumes. Volume I of the manual is a thorough description of the code INTRA, the Physical modelling of INTRA and the ruling numerical methods and volume II, the User's Manual is an input description. This document, the Physical modelling of INTRA, contains code characteristics, integration methods and applications

INTRA/Mod3.2. Manual and Code Description. Volume I - Physical Modelling

Energy Technology Data Exchange (ETDEWEB)

Andersson, Jenny; Edlund, O; Hermann, J; Johansson, Lise-Lotte

1999-01-01

The INTRA Manual consists of two volumes. Volume I of the manual is a thorough description of the code INTRA, the Physical modelling of INTRA and the ruling numerical methods and volume II, the User`s Manual is an input description. This document, the Physical modelling of INTRA, contains code characteristics, integration methods and applications

Development of a generalized integral jet model

DEFF Research Database (Denmark)

Duijm, Nijs Jan; Kessler, A.; Markert, Frank

2017-01-01

Integral type models to describe stationary plumes and jets in cross-flows (wind) have been developed since about 1970. These models are widely used for risk analysis, to describe the consequences of many different scenarios. Alternatively, CFD codes are being applied, but computational requireme......Integral type models to describe stationary plumes and jets in cross-flows (wind) have been developed since about 1970. These models are widely used for risk analysis, to describe the consequences of many different scenarios. Alternatively, CFD codes are being applied, but computational...... requirements still limit the number of scenarios that can be dealt with using CFD only. The integral models, however, are not suited to handle transient releases, such as releases from pressurized equipment, where the initially high release rate decreases rapidly with time. Further, on gas ignition, a second...... model is needed to describe the rapid combustion of the flammable part of the plume (flash fire) and a third model has to be applied for the remaining jet fire. The objective of this paper is to describe the first steps of the development of an integral-type model describing the transient development...

Development of essential system technologies for advanced reactor - Development of natural circulation analysis code for integral reactor

Energy Technology Data Exchange (ETDEWEB)

Park, Goon Cherl; Park, Ik Gyu; Kim, Jae Hak; Lee, Sang Min; Kim, Tae Wan [Seoul National University, Seoul (Korea)

1999-04-01

The objective of this study is to understand the natural circulation characteristics of integral type reactors and to develope the natural circulation analysis code for integral type reactors. This study is focused on the asymmetric 3-dimensional flow during natural circulation such as 1/4 steam generator section isolation and the inclination of the reactor systems. Natural circulation experiments were done using small-scale facilities of integral reactor SMART (System-Integrated Modular Advanced ReacTor). CFX4 code was used to investigate the flow patterns and thermal mixing phenomena in upper pressure header and downcomer. Differences between normal operation of all steam generators and the 1/4 section isolation conditions were observed and the results were used as the data 1/4 section isolation conditions were observed and the results were used as the data for RETRAN-03/INT code validation. RETRAN-03 code was modified for the development of natural circulation analysis code for integral type reactors, which was development of natural circulation analysis code for integral type reactors, which was named as RETRAN-03/INT. 3-dimensional analysis models for asymmetric flow in integral type reactors were developed using vector momentum equations in RETRAN-03. Analysis results using RETRAN-03/INT were compared with experimental and CFX4 analysis results and showed good agreements. The natural circulation characteristics obtained in this study will provide the important and fundamental design features for the future small and medium integral reactors. (author). 29 refs., 75 figs., 18 tabs.

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)

Integrated Debugging of Modelica Models

Directory of Open Access Journals (Sweden)

Adrian Pop

2014-04-01

Full Text Available The high abstraction level of equation-based object-oriented (EOO languages such as Modelica has the drawback that programming and modeling errors are often hard to find. In this paper we present integrated static and dynamic debugging methods for Modelica models and a debugger prototype that addresses several of those problems. The goal is an integrated debugging framework that combines classical debugging techniques with special techniques for equation-based languages partly based on graph visualization and interaction. To our knowledge, this is the first Modelica debugger that supports both equation-based transformational and algorithmic code debugging in an integrated fashion.

Integrated transport code system for a multicomponent plasma in a gas dynamic trap

International Nuclear Information System (INIS)

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

2000-01-01

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

Code modernization and modularization of APEX and SWAT watershed simulation models

Science.gov (United States)

SWAT (Soil and Water Assessment Tool) and APEX (Agricultural Policy / Environmental eXtender) are respectively large and small watershed simulation models derived from EPIC Environmental Policy Integrated Climate), a field-scale agroecology simulation model. All three models are coded in FORTRAN an...

Shadowfax: Moving mesh hydrodynamical integration code

Science.gov (United States)

Vandenbroucke, Bert

2016-05-01

Shadowfax simulates galaxy evolution. Written in object-oriented modular C++, it evolves a mixture of gas, subject to the laws of hydrodynamics and gravity, and any collisionless fluid only subject to gravity, such as cold dark matter or stars. For the hydrodynamical integration, it makes use of a (co-) moving Lagrangian mesh. The code has a 2D and 3D version, contains utility programs to generate initial conditions and visualize simulation snapshots, and its input/output is compatible with a number of other simulation codes, e.g. Gadget2 (ascl:0003.001) and GIZMO (ascl:1410.003).

Basic data, computer codes and integral experiments: The tools for modelling in nuclear technology

International Nuclear Information System (INIS)

Sartori, E.

2001-01-01

When studying applications in nuclear technology we need to understand and be able to predict the behavior of systems manufactured by human enterprise. First, the underlying basic physical and chemical phenomena need to be understood. We have then to predict the results from the interplay of the large number of the different basic events: i.e. the macroscopic effects. In order to be able to build confidence in our modelling capability, we need then to compare these results against measurements carried out on such systems. The different levels of modelling require the solution of different types of equations using different type of parameters. The tools required for carrying out a complete validated analysis are: - The basic nuclear or chemical data; - The computer codes, and; - The integral experiments. This article describes the role each component plays in a computational scheme designed for modelling purposes. It describes also which tools have been developed and are internationally available. The role of the OECD/NEA Data Bank, the Radiation Shielding Information Computational Center (RSICC), and the IAEA Nuclear Data Section are playing in making these elements available to the community of scientists and engineers is described. (author)

Integrating Renewable Energy Requirements Into Building Energy Codes

Energy Technology Data Exchange (ETDEWEB)

Kaufmann, John R.; Hand, James R.; Halverson, Mark A.

2011-07-01

This report evaluates how and when to best integrate renewable energy requirements into building energy codes. The basic goals were to: (1) provide a rough guide of where we’re going and how to get there; (2) identify key issues that need to be considered, including a discussion of various options with pros and cons, to help inform code deliberations; and (3) to help foster alignment among energy code-development organizations. The authors researched current approaches nationally and internationally, conducted a survey of key stakeholders to solicit input on various approaches, and evaluated the key issues related to integration of renewable energy requirements and various options to address those issues. The report concludes with recommendations and a plan to engage stakeholders. This report does not evaluate whether the use of renewable energy should be required on buildings; that question involves a political decision that is beyond the scope of this report.

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

International Nuclear Information System (INIS)

Song, Yong Mann; Park, Soo Yong

1998-03-01

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

Containment Modelling with the ASTEC Code

International Nuclear Information System (INIS)

Sadek, Sinisa; Grgic, Davor

2014-01-01

ASTEC is an integral computer code jointly developed by Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fur Anlagen-und Reaktorsicherheit (GRS, Germany) to assess the nuclear power plant behaviour during a severe accident (SA). It consists of 13 coupled modules which compute various SA phenomena in primary and secondary circuits of the nuclear power plants (NPP), and in the containment. The ASTEC code was used to model and to simulate NPP behaviour during a postulated station blackout accident in the NPP Krsko, a two-loop pressurized water reactor (PWR) plant. The primary system of the plant was modelled with 110 thermal hydraulic (TH) volumes, 113 junctions and 128 heat structures. The secondary system was modelled with 76 TH volumes, 77 junctions and 87 heat structures. The containment was modelled with 10 TH volumes by taking into account containment representation as a set of distinctive compartments, connected with 23 junctions. A total of 79 heat structures were used to simulate outer containment walls and internal steel and concrete structures. Prior to the transient calculation, a steady state analysis was performed. In order to achieve correct plant initial conditions, the operation of regulation systems was modelled. Parameters which were subjected to regulation were the pressurizer pressure, the pressurizer narrow range level and steam mass flow rates in the steam lines. The accident analysis was focused on containment behaviour, however the complete integral NPP analysis was carried out in order to provide correct boundary conditions for the containment calculation. During the accident, the containment integrity was challenged by release of reactor system coolant through degraded coolant pump seals and, later in the accident following release of the corium out of the reactor pressure vessel, by the molten corium concrete interaction and direct containment heating mechanisms. Impact of those processes on relevant

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

International Nuclear Information System (INIS)

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

1988-01-01

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

ITS - The integrated TIGER series of coupled electron/photon Monte Carlo transport codes

International Nuclear Information System (INIS)

Halbleib, J.A.; Mehlhorn, T.A.

1985-01-01

The TIGER series of time-independent coupled electron/photon Monte Carlo transport codes is a group of multimaterial, multidimensional codes designed to provide a state-of-the-art description of the production and transport of the electron/photon cascade. The codes follow both electrons and photons from 1.0 GeV down to 1.0 keV, and the user has the option of combining the collisional transport with transport in macroscopic electric and magnetic fields of arbitrary spatial dependence. Source particles can be either electrons or photons. The most important output data are (a) charge and energy deposition profiles, (b) integral and differential escape coefficients for both electrons and photons, (c) differential electron and photon flux, and (d) pulse-height distributions for selected regions of the problem geometry. The base codes of the series differ from one another primarily in their dimensionality and geometric modeling. They include (a) a one-dimensional multilayer code, (b) a code that describes the transport in two-dimensional axisymmetric cylindrical material geometries with a fully three-dimensional description of particle trajectories, and (c) a general three-dimensional transport code which employs a combinatorial geometry scheme. These base codes were designed primarily for describing radiation transport for those situations in which the detailed atomic structure of the transport medium is not important. For some applications, it is desirable to have a more detailed model of the low energy transport. The system includes three additional codes that contain a more elaborate ionization/relaxation model than the base codes. Finally, the system includes two codes that combine the collisional transport of the multidimensional base codes with transport in macroscopic electric and magnetic fields of arbitrary spatial dependence

Development of FBR integrity system code. Basic concept

International Nuclear Information System (INIS)

Asayama, Tai

2001-05-01

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

A research on verification of the CONTAIN CODE model and the uncertainty reduction method for containment integrity

Energy Technology Data Exchange (ETDEWEB)

Park, Jae-Hong; Kim, Moo-Hwan; Bae, Seong-Won; Byun, Sang-Chul [Pohang University of Science and Technology, Pohang (Korea, Republic of)

1998-03-15

The final objectives of this study are to establish the way of measuring the integrity of containment building structures and safety analysis in the period of a postuIated severe accidents and to decrease the uncertainty of these methods. For that object, the CONTAIN 1.2 codes model for analyzing the severe accidents phenomena and the heat transfer between the air inside the containment buildings and inner walls have been reviewed and analyzed. For the double containment wall provided to the next generation nuclear reactor, which is different to the previous type of containment, the temperature and pressure rising history were calculated and compared to the results of previous ones.

Parallel processing of structural integrity analysis codes

International Nuclear Information System (INIS)

Swami Prasad, P.; Dutta, B.K.; Kushwaha, H.S.

1996-01-01

Structural integrity analysis forms an important role in assessing and demonstrating the safety of nuclear reactor components. This analysis is performed using analytical tools such as Finite Element Method (FEM) with the help of digital computers. The complexity of the problems involved in nuclear engineering demands high speed computation facilities to obtain solutions in reasonable amount of time. Parallel processing systems such as ANUPAM provide an efficient platform for realising the high speed computation. The development and implementation of software on parallel processing systems is an interesting and challenging task. The data and algorithm structure of the codes plays an important role in exploiting the parallel processing system capabilities. Structural analysis codes based on FEM can be divided into two categories with respect to their implementation on parallel processing systems. The first category codes such as those used for harmonic analysis, mechanistic fuel performance codes need not require the parallelisation of individual modules of the codes. The second category of codes such as conventional FEM codes require parallelisation of individual modules. In this category, parallelisation of equation solution module poses major difficulties. Different solution schemes such as domain decomposition method (DDM), parallel active column solver and substructuring method are currently used on parallel processing systems. Two codes, FAIR and TABS belonging to each of these categories have been implemented on ANUPAM. The implementation details of these codes and the performance of different equation solvers are highlighted. (author). 5 refs., 12 figs., 1 tab

Integrated burnup calculation code system SWAT

International Nuclear Information System (INIS)

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

1997-11-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Integration of CFD codes and advanced combustion models for quantitative burnout determination

Energy Technology Data Exchange (ETDEWEB)

Javier Pallares; Inmaculada Arauzo; Alan Williams [University of Zaragoza, Zaragoza (Spain). Centre of Research for Energy Resources and Consumption (CIRCE)

2007-10-15

CFD codes and advanced kinetics combustion models are extensively used to predict coal burnout in large utility boilers. Modelling approaches based on CFD codes can accurately solve the fluid dynamics equations involved in the problem but this is usually achieved by including simple combustion models. On the other hand, advanced kinetics combustion models can give a detailed description of the coal combustion behaviour by using a simplified description of the flow field, this usually being obtained from a zone-method approach. Both approximations describe correctly general trends on coal burnout, but fail to predict quantitative values. In this paper a new methodology which takes advantage of both approximations is described. In the first instance CFD solutions were obtained of the combustion conditions in the furnace in the Lamarmora power plant (ASM Brescia, Italy) for a number of different conditions and for three coals. Then, these furnace conditions were used as inputs for a more detailed chemical combustion model to predict coal burnout. In this, devolatilization was modelled using a commercial macromolecular network pyrolysis model (FG-DVC). For char oxidation an intrinsic reactivity approach including thermal annealing, ash inhibition and maceral effects, was used. Results from the simulations were compared against plant experimental values, showing a reasonable agreement in trends and quantitative values. 28 refs., 4 figs., 4 tabs.

Direct containment heating models in the CONTAIN code

International Nuclear Information System (INIS)

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

1995-08-01

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

Direct containment heating models in the CONTAIN code

Energy Technology Data Exchange (ETDEWEB)

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

1995-08-01

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

Feasibility of the integration of CRONOS, a 3-D neutronics code, into real-time simulators

International Nuclear Information System (INIS)

Ragusa, J.C.

2001-01-01

In its effort to contribute to nuclear power plant safety, CEA proposes the integration of an engineering grade 3-D neutronics code into a real-time plant analyser. This paper describes the capabilities of the neutronics code CRONOS to achieve a fast running performance. First, we will present current core models in simulators and explain their drawbacks. Secondly, the mean features of CRONOS's spatial-kinetics methods will be reviewed. We will then present an optimum core representation with respect to mesh size, choice of finite elements (FE) basis and execution time, for accurate results as well as the multi 1-D thermal-hydraulics (T/H) model developed to take into account 3-D effects in updating the cross-sections. A Main Steam Line Break (MSLB) End-of-Life (EOL) Hot-Zero-Power (HZP) accident will be used as an example, before we conclude with the perspectives of integrating CRONOS's 3-D core model into real-time simulators. (author)

Feasibility of the integration of CRONOS, a 3-D neutronics code, into real-time simulators

Energy Technology Data Exchange (ETDEWEB)

Ragusa, J.C. [CEA Saclay, Dept. de Mecanique et de Technologie, 91 - Gif-sur-Yvette (France)

2001-07-01

In its effort to contribute to nuclear power plant safety, CEA proposes the integration of an engineering grade 3-D neutronics code into a real-time plant analyser. This paper describes the capabilities of the neutronics code CRONOS to achieve a fast running performance. First, we will present current core models in simulators and explain their drawbacks. Secondly, the mean features of CRONOS's spatial-kinetics methods will be reviewed. We will then present an optimum core representation with respect to mesh size, choice of finite elements (FE) basis and execution time, for accurate results as well as the multi 1-D thermal-hydraulics (T/H) model developed to take into account 3-D effects in updating the cross-sections. A Main Steam Line Break (MSLB) End-of-Life (EOL) Hot-Zero-Power (HZP) accident will be used as an example, before we conclude with the perspectives of integrating CRONOS's 3-D core model into real-time simulators. (author)

Integrated severe accident containment analysis with the CONTAIN computer code

International Nuclear Information System (INIS)

Bergeron, K.D.; Williams, D.C.; Rexroth, P.E.; Tills, J.L.

1985-12-01

Analysis of physical and radiological conditions iunside the containment building during a severe (core-melt) nuclear reactor accident requires quantitative evaluation of numerous highly disparate yet coupled phenomenologies. These include two-phase thermodynamics and thermal-hydraulics, aerosol physics, fission product phenomena, core-concrete interactions, the formation and combustion of flammable gases, and performance of engineered safety features. In the past, this complexity has meant that a complete containment analysis would require application of suites of separate computer codes each of which would treat only a narrower subset of these phenomena, e.g., a thermal-hydraulics code, an aerosol code, a core-concrete interaction code, etc. In this paper, we describe the development and some recent applications of the CONTAIN code, which offers an integrated treatment of the dominant containment phenomena and the interactions among them. We describe the results of a series of containment phenomenology studies, based upon realistic accident sequence analyses in actual plants. These calculations highlight various phenomenological effects that have potentially important implications for source term and/or containment loading issues, and which are difficult or impossible to treat using a less integrated code suite

Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10

International Nuclear Information System (INIS)

Lee, Y. G.; Kim, J. W.; Yoon, S. J.; Park, G. C.

2010-10-01

Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary system in the transients is confirmed. (Author)

Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-10-15

Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary system in the transients is confirmed. (Author)

Impacts of Model Building Energy Codes

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-31

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

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

International Nuclear Information System (INIS)

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

1991-06-01

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

Validating the BISON fuel performance code to integral LWR experiments

Energy Technology Data Exchange (ETDEWEB)

Williamson, R.L., E-mail: Richard.Williamson@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gamble, K.A., E-mail: Kyle.Gamble@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Perez, D.M., E-mail: Danielle.Perez@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Novascone, S.R., E-mail: Stephen.Novascone@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Pastore, G., E-mail: Giovanni.Pastore@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gardner, R.J., E-mail: Russell.Gardner@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Hales, J.D., E-mail: Jason.Hales@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Liu, W., E-mail: Wenfeng.Liu@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States); Mai, A., E-mail: Anh.Mai@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States)

2016-05-15

Highlights: • The BISON multidimensional fuel performance code is being validated to integral LWR experiments. • Code and solution verification are necessary prerequisites to validation. • Fuel centerline temperature comparisons through all phases of fuel life are very reasonable. • Accuracy in predicting fission gas release is consistent with state-of-the-art modeling and the involved uncertainties. • Rod diameter comparisons are not satisfactory and further investigation is underway. - Abstract: BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON's computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to date for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Results demonstrate that (1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, (2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and (3) comparison

A long-term, integrated impact assessment of alternative building energy code scenarios in China

International Nuclear Information System (INIS)

Yu, Sha; Eom, Jiyong; Evans, Meredydd; Clarke, Leon

2014-01-01

China is the second largest building energy user in the world, ranking first and third in residential and commercial energy consumption. Beginning in the early 1980s, the Chinese government has developed a variety of building energy codes to improve building energy efficiency and reduce total energy demand. This paper studies the impact of building energy codes on energy use and CO 2 emissions by using a detailed building energy model that represents four distinct climate zones each with three building types, nested in a long-term integrated assessment framework GCAM. An advanced building stock module, coupled with the building energy model, is developed to reflect the characteristics of future building stock and its interaction with the development of building energy codes in China. This paper also evaluates the impacts of building codes on building energy demand in the presence of economy-wide carbon policy. We find that building energy codes would reduce Chinese building energy use by 13–22% depending on building code scenarios, with a similar effect preserved even under the carbon policy. The impact of building energy codes shows regional and sectoral variation due to regionally differentiated responses of heating and cooling services to shell efficiency improvement. - Highlights: • We assessed long-term impacts of building codes and climate policy using GCAM. • Building energy codes would reduce Chinese building energy use by 13–22%. • The impacts of codes on building energy use vary by climate region and sub-sector

Influence of Modelling Options in RELAP5/SCDAPSIM and MAAP4 Computer Codes on Core Melt Progression and Reactor Pressure Vessel Integrity

Directory of Open Access Journals (Sweden)

Siniša Šadek

2010-01-01

Full Text Available RELAP5/SCDAPSIM and MAAP4 are two widely used severe accident computer codes for the integral analysis of the core and the reactor pressure vessel behaviour following the core degradation. The objective of the paper is the comparison of code results obtained by application of different modelling options and the evaluation of influence of thermal hydraulic behaviour of the plant on core damage progression. The analysed transient was postulated station blackout in NPP Krško with a leakage from reactor coolant pump seals. Two groups of calculations were performed where each group had a different break area and, thus, a different leakage rate. Analyses have shown that MAAP4 results were more sensitive to varying thermal hydraulic conditions in the primary system. User-defined parameters had to be carefully selected when the MAAP4 model was developed, in contrast to the RELAP5/SCDAPSIM model where those parameters did not have any significant impact on final results.

Development and assessment of best estimate integrated safety analysis code

International Nuclear Information System (INIS)

Chung, Bub Dong; Lee, Young Jin; Hwang, Moon Kyu

2007-03-01

Improvement of the integrated safety analysis code MARS3.0 has been carried out and a multi-D safety analysis application system has been established. Iterative matrix solver and parallel processing algorithm have been introduced, and a LINUX version has been generated to enable MARS to run in cluster PCs. MARS variables and sub-routines have been reformed and modularised to simplify code maintenance. Model uncertainty analyses have been performed for THTF, FLECHT, NEPTUN, and LOFT experiments as well as APR1400 plant. Participations in international cooperation research projects such as OECD BEMUSE, SETH, PKL, BFBT, and TMI-2 have been actively pursued as part of code assessment efforts. The assessment, evaluation and experimental data obtained through international cooperation projects have been registered and maintained in the T/H Databank. Multi-D analyses of APR1400 LBLOCA, DVI Break, SLB, and SGTR have been carried out as a part of application efforts in multi-D safety analysis. GUI based 3D input generator has been developed for user convenience. Operation of the MARS Users Group (MUG) was continued and through MUG, the technology has been transferred to 24 organisations. A set of 4 volumes of user manuals has been compiled and the correction reports for the code errors reported during MARS development have been published

Development and assessment of best estimate integrated safety analysis code

Energy Technology Data Exchange (ETDEWEB)

Chung, Bub Dong; Lee, Young Jin; Hwang, Moon Kyu (and others)

2007-03-15

Improvement of the integrated safety analysis code MARS3.0 has been carried out and a multi-D safety analysis application system has been established. Iterative matrix solver and parallel processing algorithm have been introduced, and a LINUX version has been generated to enable MARS to run in cluster PCs. MARS variables and sub-routines have been reformed and modularised to simplify code maintenance. Model uncertainty analyses have been performed for THTF, FLECHT, NEPTUN, and LOFT experiments as well as APR1400 plant. Participations in international cooperation research projects such as OECD BEMUSE, SETH, PKL, BFBT, and TMI-2 have been actively pursued as part of code assessment efforts. The assessment, evaluation and experimental data obtained through international cooperation projects have been registered and maintained in the T/H Databank. Multi-D analyses of APR1400 LBLOCA, DVI Break, SLB, and SGTR have been carried out as a part of application efforts in multi-D safety analysis. GUI based 3D input generator has been developed for user convenience. Operation of the MARS Users Group (MUG) was continued and through MUG, the technology has been transferred to 24 organisations. A set of 4 volumes of user manuals has been compiled and the correction reports for the code errors reported during MARS development have been published.

Approaches in highly parameterized inversion - PEST++, a Parameter ESTimation code optimized for large environmental models

Science.gov (United States)

Welter, David E.; Doherty, John E.; Hunt, Randall J.; Muffels, Christopher T.; Tonkin, Matthew J.; Schreuder, Willem A.

2012-01-01

An object-oriented parameter estimation code was developed to incorporate benefits of object-oriented programming techniques for solving large parameter estimation modeling problems. The code is written in C++ and is a formulation and expansion of the algorithms included in PEST, a widely used parameter estimation code written in Fortran. The new code is called PEST++ and is designed to lower the barriers of entry for users and developers while providing efficient algorithms that can accommodate large, highly parameterized problems. This effort has focused on (1) implementing the most popular features of PEST in a fashion that is easy for novice or experienced modelers to use and (2) creating a software design that is easy to extend; that is, this effort provides a documented object-oriented framework designed from the ground up to be modular and extensible. In addition, all PEST++ source code and its associated libraries, as well as the general run manager source code, have been integrated in the Microsoft Visual Studio® 2010 integrated development environment. The PEST++ code is designed to provide a foundation for an open-source development environment capable of producing robust and efficient parameter estimation tools for the environmental modeling community into the future.

Synthesis of the ASTEC integral code activities in SARNET – Focus on ASTEC V2 plant applications

International Nuclear Information System (INIS)

Chatelard, P.; Reinke, N.; Ezzidi, A.; Lombard, V.; Barnak, M.; Lajtha, G.; Slaby, J.; Constantin, M.; Majumdar, P.

2014-01-01

Highlights: • Independent assessment of the ASTEC severe accident code vs. experiments is summarised. • Main remaining modelling issues and development perspectives are identified. • Independent assessment of ASTEC code at full scale conditions is described. • Main requirements to address BWR and PHWR types of reactors are identified. - Abstract: Among the 43 organisations which joined the SARNET2 FP7 project from 2009 to 2013, 31 have been involved in the activities on the ASTEC code. This paper presents a synthesis of the main achievements that have been obtained on the ASTEC V2 integral code, jointly developed by IRSN (France) and GRS (Germany), on development, validation vs. experimental data and applications at full scale conditions for both Gen.II and Gen.III plants. As to code development, while the current V2.0 series of ASTEC versions was continuously improved (elaboration and release by IRSN and GRS of three successive V2.0 revisions), IRSN and GRS have also intensively continued in parallel the elaboration of the second ASTEC V2 major version (version V2.1) to be delivered end of 2014. Regarding code validation vs. experiments, the partners have assessed the V2.0 version and subsequent revisions vs. more than 50 experiments; this extended assessment notably confirmed that most models are today close to the State of the Art, while it also corroborated the yet known key-topics on which modelling efforts should focus in priority. As to plant applications, the comparison of ASTEC results with other codes allows concluding on a globally good agreement for in-vessel and ex-vessel severe accident progression. As to ASTEC adaptations to BWR and PHWR, significant achievements have been obtained through the elaboration and integration in the future V2.1 version of dedicated core degradation models, notably to account for multi coolant flows

Fatigue modelling according to the JCSS Probabilistic model code

NARCIS (Netherlands)

Vrouwenvelder, A.C.W.M.

2007-01-01

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

Pump Component Model in SPACE Code

International Nuclear Information System (INIS)

Kim, Byoung Jae; Kim, Kyoung Doo

2010-08-01

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

MIDAS/PK code development using point kinetics model

International Nuclear Information System (INIS)

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

1999-01-01

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

A research on the verification of models used in the computational codes and the uncertainty reduction method for the containment integrity evaluation

Energy Technology Data Exchange (ETDEWEB)

Kim, Moo Hwan; Seo, Kyoung Woo [POSTECH, Pohang (Korea, Republic of)

2001-03-15

In the probability approach, the calculated CCFPs of all the scenarios were zero, which meant that it was expected that for all the accident scenarios the maximum pressure load induced by DCH was lower than the containment failure pressure obtained from the fragility curve. Thus, it can be stated that the KSNP containment is robust to the DCH threat. And uncertainty of computer codes used to be two (deterministic and probabilistic) approaches were reduced by the sensitivity tests and the research with the verification and comparison of the DCH models in each code. So, this research was to evaluate synthetic result of DCH issue and expose accurate methodology to assess containment integrity about operating PWR in Korea.

Studies on DANESS Code Modeling

International Nuclear Information System (INIS)

Jeong, Chang Joon

2009-09-01

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

Modeling of the WWER-1000 fuel-rod behavior in steady-state condition with FRAPCONE-3 computer code

International Nuclear Information System (INIS)

Andreeva, Marina; Totev, Totju; Stoyanov, Stoyan

2008-01-01

It is presented within the paper the results of the modeling and the assessment of the integral code predictions of the WWER fuel-rod behavior in steady-state condition. The assessments in this paper have used the MASSIH and ANS 5.4 subroutine in the code. The modeling and calculations have been performed with FRAPCONE-3 computer code in Argonne National Laboratory, USA

Model-Based Least Squares Reconstruction of Coded Source Neutron Radiographs: Integrating the ORNL HFIR CG1D Source Model

Energy Technology Data Exchange (ETDEWEB)

Santos-Villalobos, Hector J [ORNL; Gregor, Jens [University of Tennessee, Knoxville (UTK); Bingham, Philip R [ORNL

2014-01-01

At the present, neutron sources cannot be fabricated small and powerful enough in order to achieve high resolution radiography while maintaining an adequate flux. One solution is to employ computational imaging techniques such as a Magnified Coded Source Imaging (CSI) system. A coded-mask is placed between the neutron source and the object. The system resolution is increased by reducing the size of the mask holes and the flux is increased by increasing the size of the coded-mask and/or the number of holes. One limitation of such system is that the resolution of current state-of-the-art scintillator-based detectors caps around 50um. To overcome this challenge, the coded-mask and object are magnified by making the distance from the coded-mask to the object much smaller than the distance from object to detector. In previous work, we have shown via synthetic experiments that our least squares method outperforms other methods in image quality and reconstruction precision because of the modeling of the CSI system components. However, the validation experiments were limited to simplistic neutron sources. In this work, we aim to model the flux distribution of a real neutron source and incorporate such a model in our least squares computational system. We provide a full description of the methodology used to characterize the neutron source and validate the method with synthetic experiments.

Theory, modeling, and integrated studies in the Arase (ERG) project

Science.gov (United States)

Seki, Kanako; Miyoshi, Yoshizumi; Ebihara, Yusuke; Katoh, Yuto; Amano, Takanobu; Saito, Shinji; Shoji, Masafumi; Nakamizo, Aoi; Keika, Kunihiro; Hori, Tomoaki; Nakano, Shin'ya; Watanabe, Shigeto; Kamiya, Kei; Takahashi, Naoko; Omura, Yoshiharu; Nose, Masahito; Fok, Mei-Ching; Tanaka, Takashi; Ieda, Akimasa; Yoshikawa, Akimasa

2018-02-01

Understanding of underlying mechanisms of drastic variations of the near-Earth space (geospace) is one of the current focuses of the magnetospheric physics. The science target of the geospace research project Exploration of energization and Radiation in Geospace (ERG) is to understand the geospace variations with a focus on the relativistic electron acceleration and loss processes. In order to achieve the goal, the ERG project consists of the three parts: the Arase (ERG) satellite, ground-based observations, and theory/modeling/integrated studies. The role of theory/modeling/integrated studies part is to promote relevant theoretical and simulation studies as well as integrated data analysis to combine different kinds of observations and modeling. Here we provide technical reports on simulation and empirical models related to the ERG project together with their roles in the integrated studies of dynamic geospace variations. The simulation and empirical models covered include the radial diffusion model of the radiation belt electrons, GEMSIS-RB and RBW models, CIMI model with global MHD simulation REPPU, GEMSIS-RC model, plasmasphere thermosphere model, self-consistent wave-particle interaction simulations (electron hybrid code and ion hybrid code), the ionospheric electric potential (GEMSIS-POT) model, and SuperDARN electric field models with data assimilation. ERG (Arase) science center tools to support integrated studies with various kinds of data are also briefly introduced.[Figure not available: see fulltext.

Improvement of MARS code reflood model

International Nuclear Information System (INIS)

Hwang, Moonkyu; Chung, Bub-Dong

2011-01-01

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

TRAC-CFD code integration and its application to containment analysis

International Nuclear Information System (INIS)

Tahara, M.; Arai, K.; Oikawa, H.

2004-01-01

Several safety systems utilizing natural driving force have been recently adopted for operating reactors, or applied to next-generation reactor design. Examples of these safety systems are the Passive Containment Cooling System (PCCS) and the Drywell Cooler (DWC) for removing decay heat, and the Passive Auto-catalytic Recombiner (PAR) for removing flammable gas in reactor containment during an accident. DWC is used in almost all Boiling Water Reactors (BWR) in service. PAR has been introduced for some reactors in Europe and will be introduced for Japanese reactors. PCCS is a safety device of next-generation BWR. The functional mechanism of these safety systems is closely related to the transient of the thermal-hydraulic condition of the containment atmosphere. The performance depends on the containment atmospheric condition, which is eventually affected by the mass and energy changes caused by the safety system. Therefore, the thermal fluid dynamics in the containment vessel should be appropriately considered in detail to properly estimate the performance of these systems. A computational fluid dynamics (CFD) code is useful for evaluating detailed thermal hydraulic behavior related to this equipment. However, it also requires a considerable amount of computational resources when it is applied to whole containment system transient analysis. The paper describes the method and structure of the integrated analysis tool, and discusses the results of its application to the start-up behavior analysis of a containment cooling system, a drywell local cooler. The integrated analysis code was developed and applied to estimate the DWC performance during a severe accident. The integrated analysis tool is composed of three codes, TRAC-PCV, CFD-DW and TRAC-CC, and analyzes the interaction of the natural convection and steam condensation of the DWC as well as analyzing the thermal hydraulic transient behavior of the containment vessel during a severe accident in detail. The

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

International Nuclear Information System (INIS)

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

1999-01-01

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

ELM triggering conditions for the integrated modeling of H-mode plasmas

International Nuclear Information System (INIS)

Pankin, A.Y.; Schnack, D.D.; Bateman, G.; Kritz, A.H.; Brennan, D.P.; Snyder, P.B.; Voitsekhovitch, I.; Kruger, S.; Janeschitz, G.; Onjun, T.; Pacher, G.W.; Pacher, H.D.

2004-01-01

Recent advances in the integrated modeling of ELMy (edge localized mode) H-mode plasmas are presented. A model for the H-mode pedestal and for the triggering of ELMs predicts the height, width, and shape of the H-mode pedestal and the frequency and width of ELMs. Formation of the pedestal and the L-H transition is the direct result of E r x B flow shear suppression of anomalous transport. The periodic ELM crashes are triggered by either the ballooning or peeling MHD instabilities. The BALOO, DCON, and ELITE ideal MHD stability codes are used to derive a new parametric expression for the peeling-ballooning threshold. The new dependence for the peeling-ballooning threshold is implemented in the ASTRA transport code. Results of integrated modeling of DIII-D like discharges are presented and compared with experimental observations. The results from the ideal MHD stability codes are compared with results from the resistive MHD stability code NIMROD. (authors)

Experimental assessment of computer codes used for safety analysis of integral reactors

Energy Technology Data Exchange (ETDEWEB)

Falkov, A.A.; Kuul, V.S.; Samoilov, O.B. [OKB Mechanical Engineering, Nizhny Novgorod (Russian Federation)

1995-09-01

Peculiarities of integral reactor thermohydraulics in accidents are associated with presence of noncondensable gas in built-in pressurizer, absence of pumped ECCS, use of guard vessel for LOCAs localisation and passive RHRS through in-reactor HX`s. These features defined the main trends in experimental investigations and verification efforts for computer codes applied. The paper reviews briefly the performed experimental investigation of thermohydraulics of AST-500, VPBER600-type integral reactors. The characteristic of UROVEN/MB-3 code for LOCAs analysis in integral reactors and results of its verification are given. The assessment of RELAP5/mod3 applicability for accident analysis in integral reactor is presented.

Study of MHD stability beta limit in LHD by hierarchy integrated simulation code

International Nuclear Information System (INIS)

Sato, M.; Watanabe, K.Y.; Nakamura, Y.

2008-10-01

The beta limit by the ideal MHD instabilities (so-called 'MHD stability beta limit') for helical plasmas is studied by a hierarchy integrated simulation code. A numerical model for the effect of the MHD instabilities is introduced such that the pressure profile is flattened around the rational surface due to the MHD instabilities. The width of the flattening of the pressure gradient is determined from the width of the eigenmode structure of the MHD instabilities. It is assumed that there is the upper limit of the mode number of the MHD instabilities which directly affect the pressure gradient. The upper limit of the mode number is determined using a recent high beta experiment in the Large Helical Device (LHD). The flattening of the pressure gradient is calculated by the transport module in a hierarchy integrated code. The achievable volume averaged beta value in the LHD is expected to be beyond 6%. (author)

Thermohydraulic modeling of nuclear thermal rockets: The KLAXON code

International Nuclear Information System (INIS)

Hall, M.L.; Rider, W.J.; Cappiello, M.W.

1992-01-01

The hydrogen flow from the storage tanks, through the reactor core, and out the nozzle of a Nuclear Thermal Rocket is an integral design consideration. To provide an analysis and design tool for this phenomenon, the KLAXON code is being developed. A shock-capturing numerical methodology is used to model the gas flow (the Harten, Lax, and van Leer method, as implemented by Einfeldt). Preliminary results of modeling the flow through the reactor core and nozzle are given in this paper

Development of seismic analysis model for HTGR core on commercial FEM code

International Nuclear Information System (INIS)

Tsuji, Nobumasa; Ohashi, Kazutaka

2015-01-01

The aftermath of the Great East Japan Earthquake prods to revise the design basis earthquake intensity severely. In aseismic design of block-type HTGR, the securement of structural integrity of core blocks and other structures which are made of graphite become more important. For the aseismic design of block-type HTGR, it is necessary to predict the motion of core blocks which are collided with adjacent blocks. Some seismic analysis codes have been developed in 1970s, but these codes are special purpose-built codes and have poor collaboration with other structural analysis code. We develop the vertical 2 dimensional analytical model on multi-purpose commercial FEM code, which take into account the multiple impacts and friction between block interfaces and rocking motion on contact with dowel pins of the HTGR core by using contact elements. This model is verified by comparison with the experimental results of 12 column vertical slice vibration test. (author)

Integrating bar-code devices with computerized MC and A systems

International Nuclear Information System (INIS)

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

1998-01-01

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

Summary Report for ASC L2 Milestone #4782: Assess Newly Emerging Programming and Memory Models for Advanced Architectures on Integrated Codes

Energy Technology Data Exchange (ETDEWEB)

Neely, J. R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hornung, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Black, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Robinson, P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2014-09-29

This document serves as a detailed companion to the powerpoint slides presented as part of the ASC L2 milestone review for Integrated Codes milestone #4782 titled “Assess Newly Emerging Programming and Memory Models for Advanced Architectures on Integrated Codes”, due on 9/30/2014, and presented for formal program review on 9/12/2014. The program review committee is represented by Mike Zika (A Program Project Lead for Kull), Brian Pudliner (B Program Project Lead for Ares), Scott Futral (DEG Group Lead in LC), and Mike Glass (Sierra Project Lead at Sandia). This document, along with the presentation materials, and a letter of completion signed by the review committee will act as proof of completion for this milestone.

Dual coding: a cognitive model for psychoanalytic research.

Science.gov (United States)

Bucci, W

1985-01-01

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

Integrative and distinctive coding of visual and conceptual object features in the ventral visual stream.

Science.gov (United States)

Martin, Chris B; Douglas, Danielle; Newsome, Rachel N; Man, Louisa Ly; Barense, Morgan D

2018-02-02

A significant body of research in cognitive neuroscience is aimed at understanding how object concepts are represented in the human brain. However, it remains unknown whether and where the visual and abstract conceptual features that define an object concept are integrated. We addressed this issue by comparing the neural pattern similarities among object-evoked fMRI responses with behavior-based models that independently captured the visual and conceptual similarities among these stimuli. Our results revealed evidence for distinctive coding of visual features in lateral occipital cortex, and conceptual features in the temporal pole and parahippocampal cortex. By contrast, we found evidence for integrative coding of visual and conceptual object features in perirhinal cortex. The neuroanatomical specificity of this effect was highlighted by results from a searchlight analysis. Taken together, our findings suggest that perirhinal cortex uniquely supports the representation of fully specified object concepts through the integration of their visual and conceptual features. © 2018, Martin et al.

Integrative and distinctive coding of visual and conceptual object features in the ventral visual stream

Science.gov (United States)

Douglas, Danielle; Newsome, Rachel N; Man, Louisa LY

2018-01-01

A significant body of research in cognitive neuroscience is aimed at understanding how object concepts are represented in the human brain. However, it remains unknown whether and where the visual and abstract conceptual features that define an object concept are integrated. We addressed this issue by comparing the neural pattern similarities among object-evoked fMRI responses with behavior-based models that independently captured the visual and conceptual similarities among these stimuli. Our results revealed evidence for distinctive coding of visual features in lateral occipital cortex, and conceptual features in the temporal pole and parahippocampal cortex. By contrast, we found evidence for integrative coding of visual and conceptual object features in perirhinal cortex. The neuroanatomical specificity of this effect was highlighted by results from a searchlight analysis. Taken together, our findings suggest that perirhinal cortex uniquely supports the representation of fully specified object concepts through the integration of their visual and conceptual features. PMID:29393853

International codes and model intercomparison for intermediate energy activation yields

International Nuclear Information System (INIS)

Rolf, M.; Nagel, P.

1997-01-01

The motivation for this intercomparison came from data needs of accelerator-based waste transmutation, energy amplification and medical therapy. The aim of this exercise is to determine the degree of reliability of current nuclear reaction models and codes when calculating activation yields in the intermediate energy range up to 5000 MeV. Emphasis has been placed for a wide range of target elements ( O, Al, Fe, Co, Zr and Au). This work is mainly based on calculation of (P,xPyN) integral cross section for incident proton. A qualitative description of some of the nuclear models and code options employed is made. The systematics of graphical presentation of the results allows a quick quantitative measure of agreement or deviation. This code intercomparison highlights the fact that modeling calculations of energy activation yields may at best have uncertainties of a factor of two. The causes of such discrepancies are multi-factorial. Problems are encountered which are connected with the calculation of nuclear masses, binding energies, Q-values, shell effects, medium energy fission and Fermi break-up. (A.C.)

Advanced tokamak research with integrated modeling in JT-60 Upgrade

International Nuclear Information System (INIS)

Hayashi, N.

2010-01-01

Researches on advanced tokamak (AT) have progressed with integrated modeling in JT-60 Upgrade [N. Oyama et al., Nucl. Fusion 49, 104007 (2009)]. Based on JT-60U experimental analyses and first principle simulations, new models were developed and integrated into core, rotation, edge/pedestal, and scrape-off-layer (SOL)/divertor codes. The integrated models clarified complex and autonomous features in AT. An integrated core model was implemented to take account of an anomalous radial transport of alpha particles caused by Alfven eigenmodes. It showed the reduction in the fusion gain by the anomalous radial transport and further escape of alpha particles. Integrated rotation model showed mechanisms of rotation driven by the magnetic-field-ripple loss of fast ions and the charge separation due to fast-ion drift. An inward pinch model of high-Z impurity due to the atomic process was developed and indicated that the pinch velocity increases with the toroidal rotation. Integrated edge/pedestal model clarified causes of collisionality dependence of energy loss due to the edge localized mode and the enhancement of energy loss by steepening a core pressure gradient just inside the pedestal top. An ideal magnetohydrodynamics stability code was developed to take account of toroidal rotation and clarified a destabilizing effect of rotation on the pedestal. Integrated SOL/divertor model clarified a mechanism of X-point multifaceted asymmetric radiation from edge. A model of the SOL flow driven by core particle orbits which partially enter the SOL was developed by introducing the ion-orbit-induced flow to fluid equations.

Pre-Service Teachers' Perception of Quick Response (QR) Code Integration in Classroom Activities

Science.gov (United States)

Ali, Nagla; Santos, Ieda M.; Areepattamannil, Shaljan

2017-01-01

Quick Response (QR) codes have been discussed in the literature as adding value to teaching and learning. Despite their potential in education, more research is needed to inform practice and advance knowledge in this field. This paper investigated the integration of the QR code in classroom activities and the perceptions of the integration by…

Code Development for Control Design Applications: Phase I: Structural Modeling

International Nuclear Information System (INIS)

Bir, G. S.; Robinson, M.

1998-01-01

The design of integrated controls for a complex system like a wind turbine relies on a system model in an explicit format, e.g., state-space format. Current wind turbine codes focus on turbine simulation and not on system characterization, which is desired for controls design as well as applications like operating turbine model analysis, optimal design, and aeroelastic stability analysis. This paper reviews structural modeling that comprises three major steps: formation of component equations, assembly into system equations, and linearization

A detailed chemistry model for transient hydrogen and carbon monoxide catalytic recombination on parallel flat Pt surfaces implemented in an integral code

International Nuclear Information System (INIS)

Jimenez, Miguel A.; Martin-Valdepenas, Juan M.; Martin-Fuertes, Francisco; Fernandez, Jose A.

2007-01-01

A detailed chemistry model has been adapted and developed for surface chemistry, heat and mass transfer between H 2 /CO/air/steam/CO 2 mixtures and vertical parallel Pt-coated surfaces. This model is based onto a simplified Deutschmann reaction scheme for methane surface combustion and the analysis by Elenbaas for buoyancy-induced heat transfer between parallel plates. Mass transfer is treated by the heat and mass transfer analogy. The proposed model is able to simulate the H 2 /CO recombination phenomena characteristic of parallel-plate Passive Autocatalytic Recombiners (PARs), which have been proposed and implemented as a promising hydrogen-control strategy in the safety of nuclear power stations or other industries. The transient model is able to approach the warm-up phase of the PAR and its shut-down as well as the dynamic changes within the surrounding atmosphere. The model has been implemented within the MELCOR code and assessed against results of the Battelle Model Containment tests of the Zx series. Results show accurate predictions and a better performance than traditional methods in integral codes, i.e. empirical correlations, which are also much case-specific. Influence of CO present in the mixture on the PAR performance is also addressed in this paper

Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) verification and validation plan. version 1.

Energy Technology Data Exchange (ETDEWEB)

Bartlett, Roscoe Ainsworth; Arguello, Jose Guadalupe, Jr.; Urbina, Angel; Bouchard, Julie F.; Edwards, Harold Carter; Freeze, Geoffrey A.; Knupp, Patrick Michael; Wang, Yifeng; Schultz, Peter Andrew; Howard, Robert (Oak Ridge National Laboratory, Oak Ridge, TN); McCornack, Marjorie Turner

2011-01-01

The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. To meet this objective, NEAMS Waste IPSC M&S capabilities will be applied to challenging spatial domains, temporal domains, multiphysics couplings, and multiscale couplings. A strategic verification and validation (V&V) goal is to establish evidence-based metrics for the level of confidence in M&S codes and capabilities. Because it is economically impractical to apply the maximum V&V rigor to each and every M&S capability, M&S capabilities will be ranked for their impact on the performance assessments of various components of the repository systems. Those M&S capabilities with greater impact will require a greater level of confidence and a correspondingly greater investment in V&V. This report includes five major components: (1) a background summary of the NEAMS Waste IPSC to emphasize M&S challenges; (2) the conceptual foundation for verification, validation, and confidence assessment of NEAMS Waste IPSC M&S capabilities; (3) specifications for the planned verification, validation, and confidence-assessment practices; (4) specifications for the planned evidence information management system; and (5) a path forward for the incremental implementation of this V&V plan.

Code package {open_quotes}SVECHA{close_quotes}: Modeling of core degradation phenomena at severe accidents

Energy Technology Data Exchange (ETDEWEB)

Veshchunov, M.S.; Kisselev, A.E.; Palagin, A.V. [Nuclear Safety Institute, Moscow (Russian Federation)] [and others

1995-09-01

The code package SVECHA for the modeling of in-vessel core degradation (CD) phenomena in severe accidents is being developed in the Nuclear Safety Institute, Russian Academy of Science (NSI RAS). The code package presents a detailed mechanistic description of the phenomenology of severe accidents in a reactor core. The modules of the package were developed and validated on separate effect test data. These modules were then successfully implemented in the ICARE2 code and validated against a wide range of integral tests. Validation results have shown good agreement with separate effect tests data and with the integral tests CORA-W1/W2, CORA-13, PHEBUS-B9+.

Modeling Guidelines for Code Generation in the Railway Signaling Context

Science.gov (United States)

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

2009-01-01

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

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.

Coding with partially hidden Markov models

DEFF Research Database (Denmark)

Forchhammer, Søren; Rissanen, J.

1995-01-01

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

Integration of CAM and CNC operation through code editing and manipulation

International Nuclear Information System (INIS)

Rosli Darmawan; Shalina Sheik Muhammad

2004-01-01

The IT technology for engineering design and manufacturing has gone through significant advancement for the last 30 years. It is widely acknowledged that IT would provide competitive advantage for engineering company in term of production cycle, productivity and efficiency. The recent development in this area is on the total system integration. While standard off-shelf CAD/CAM/CNC software and hardware packages would provide solution for system integration, more often than not users will stumble upon compatibility problems. Moreover, most of the integration deals with CAD and CAM systems. CNC integration has not been fully developed. Users always found problems in the integration of CAM and CNC machine due to the different level of technological development. CNC codes have not fundamentally progressed in the last 50 years, while CAD/CAM software packages have undergone massive evolution and improvement. This paper discusses a practical solution of CAM and CNC integration through code editing and manipulation within the CAM system in order to comply with the CNC machine requirements. (Author)

Modeling of BWR core meltdown accidents - for application in the MELRPI. MOD2 computer code

Energy Technology Data Exchange (ETDEWEB)

Koh, B R; Kim, S H; Taleyarkhan, R P; Podowski, M Z; Lahey, Jr, R T

1985-04-01

This report summarizes improvements and modifications made in the MELRPI computer code. A major difference between this new, updated version of the code, called MELRPI.MOD2, and the one reported previously, concerns the inclusion of a model for the BWR emergency core cooling systems (ECCS). This model and its computer implementation, the ECCRPI subroutine, account for various emergency injection modes, for both intact and rubblized geometries. Other changes to MELRPI deal with an improved model for canister wall oxidation, rubble bed modeling, and numerical integration of system equations. A complete documentation of the entire MELRPI.MOD2 code is also given, including an input guide, list of subroutines, sample input/output and program listing.

Tri-Lab Co-Design Milestone: In-Depth Performance Portability Analysis of Improved Integrated Codes on Advanced Architecture.

Energy Technology Data Exchange (ETDEWEB)

Hoekstra, Robert J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hammond, Simon David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Richards, David [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Bergen, Ben [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-09-01

This milestone is a tri-lab deliverable supporting ongoing Co-Design efforts impacting applications in the Integrated Codes (IC) program element Advanced Technology Development and Mitigation (ATDM) program element. In FY14, the trilabs looked at porting proxy application to technologies of interest for ATS procurements. In FY15, a milestone was completed evaluating proxy applications in multiple programming models and in FY16, a milestone was completed focusing on the migration of lessons learned back into production code development. This year, the co-design milestone focuses on extracting the knowledge gained and/or code revisions back into production applications.

ELM triggering conditions for the integrated modeling of H-mode plasmas

International Nuclear Information System (INIS)

Pankin, A.Y.; Schnack, D.D.; Bateman, G.; Kritz, A.H.; Brennan, D.P.; Snyder, P.B.; Voitsekhovitch, I.; Kruger, S.; Janeschitz, G.; Onjun, T.; Pacher, G.W.; Pacher, H.D.

2005-01-01

Recent advances in the integrated modeling of ELMy H-mode plasmas are presented. A new model for the H-mode pedestal and for the triggering of ELMs predicts the height, width, and shape of the H-mode pedestal and the frequency and width of ELMs. The model for the pedestal and ELMs is used in the ASTRA integrated transport code to follow the time evolution of tokamak discharges from L-mode through the transition from L-mode to H-mode, with the formation of the H-mode pedestal, and, subsequently, to the triggering of ELMs. Turbulence driven by the ion temperature gradient mode, resistive ballooning mode, trapped electron mode, and electron temperature gradient mode contributes to the anomalous thermal transport at the plasma edge in this model. Formation of the pedestal and the L-H transition is the direct result of E(vector) r x B(vector) flow shear suppression of anomalous transport. The periodic ELM crashes are triggered by MHD instabilities. Two mechanisms for triggering ELMs are considered: ELMs are triggered by ballooning modes if the pressure gradient exceeds the ballooning threshold or by peeling modes if the edge current density exceeds the peeling mode threshold. The BALOO, DCON, and ELITE ideal MHD stability codes are used to derive a new parametric expression for the peeling-ballooning threshold. The new dependence for the peeling-ballooning threshold is implemented in the ASTRA transport code. Results of integrated modeling of DIII-D like discharges are presented and compared with experimental observations. The results from the ideal MHD stability codes are compared with results from the resistive MHD stability code NIMROD. (author)

Status of reactor core design code system in COSINE code package

Energy Technology Data Exchange (ETDEWEB)

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

2014-07-01

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

Status of reactor core design code system in COSINE code package

International Nuclear Information System (INIS)

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

2014-01-01

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

WKB: an interactive code for solving differential equations using phase integral methods

International Nuclear Information System (INIS)

White, R.B.

1978-01-01

A small code for the analysis of ordinary differential equations interactively through the use of Phase Integral Methods (WKB) has been written for use on the DEC 10. This note is a descriptive manual for those interested in using the code

Hydrodynamic Instability, Integrated Code, Laboratory Astrophysics, and Astrophysics

Science.gov (United States)

Takabe, Hideaki

2016-10-01

This is an article for the memorial lecture of Edward Teller Medal and is presented as memorial lecture at the IFSA03 conference held on September 12th, 2003, at Monterey, CA. The author focuses on his main contributions to fusion science and its extension to astrophysics in the field of theory and computation by picking up five topics. The first one is the anomalous resisitivity to hot electrons penetrating over-dense region through the ion wave turbulence driven by the return current compensating the current flow by the hot electrons. It is concluded that almost the same value of potential as the average kinetic energy of the hot electrons is realized to prevent the penetration of the hot electrons. The second is the ablative stabilization of Rayleigh-Taylor instability at ablation front and its dispersion relation so-called Takabe formula. This formula gave a principal guideline for stable target design. The author has developed an integrated code ILESTA (ID & 2D) for analyses and design of laser produced plasma including implosion dynamics. It is also applied to design high gain targets. The third is the development of the integrated code ILESTA. The forth is on Laboratory Astrophysics with intense lasers. This consists of two parts; one is review on its historical background and the other is on how we relate laser plasma to wide-ranging astrophysics and the purposes for promoting such research. In relation to one purpose, I gave a comment on anomalous transport of relativistic electrons in Fast Ignition laser fusion scheme. Finally, I briefly summarize recent activity in relation to application of the author's experience to the development of an integrated code for studying extreme phenomena in astrophysics.

Development of integrated SOL/Divertor code and simulation study of the JT-60U/JT-60SA tokamaks

International Nuclear Information System (INIS)

Kawashima, H.; Shimizu, K.; Takizuka, T.

2007-01-01

To predict the particle and heat controllability in the divertor of tokamak reactors such as ITER and to optimize the divertor design, comprehensive simulations by integrated modelling with taking in various physical processes are indispensable. For the design study of ITER divertor, the modelling codes such as B2, UEDGE and EDGE2D have been developed, and their results have contributed to the evolution of the divertor concept. In Japan Atomic Energy Agency (JAEA), SOL/divertor codes have also been developed for the interpretation and the prediction on behaviours of plasmas, neutrals and impurities in the SOL/divertor regions. The code development is originally carried out since physics models can be verified quickly and flexibly under the circumstance of close collaboration with JT-60 team. Figure 1 shows our code system, which consists of the 2 dimensional fluid code SOLDOR, the neutral Monte Carlo (MC) code NEUT2D, and the impurity MC code IMPMC. The particle simulation code PARASOL has also been developed in order to establish the physics modelling used in fluid simulations. Integration of SOLDOR, NEUT2D and IMPMC, called the '' SONIC '' code, is being carried out to simulate self-consistently the SOL/divertor plasmas in present tokamaks and in future devices. Combination of the SOLDOR and NEUT2D was completed, which has the features such as 1) high-resolution oscillation-free scheme in solving fluid equations, 2) neutral transport calculation under the fine meshes, 3) success in reduction of MC noise, 4) optimization on the massive parallel computer, etc. The simulation reproduces the X-point MARFE in the JT-60U experiment. It is found that the chemically sputtered carbon at the dome causes the radiation peaking near the X-point. The performance of divertor pumping in JT-60U is evaluated from the particle balances. We also present the divertor designing of JT-60SA, which is the modification program of JT-60U to establish high beta steady-state operation. To

Revised SWAT. The integrated burnup calculation code system

International Nuclear Information System (INIS)

Suyama, Kenya; Mochizuki, Hiroki; Kiyosumi, Takehide

2000-07-01

SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)

Revised SWAT. The integrated burnup calculation code system

Energy Technology Data Exchange (ETDEWEB)

Suyama, Kenya; Mochizuki, Hiroki [Department of Fuel Cycle Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Kiyosumi, Takehide [The Japan Research Institute, Ltd., Tokyo (Japan)

2000-07-01

SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)

Modeling of severe accident sequences with the new modules CESAR and DIVA of ASTEC system code

International Nuclear Information System (INIS)

Pignet, Sophie; Guillard, Gaetan; Barre, Francois; Repetto, Georges

2003-01-01

Systems of computer codes, so-called 'integral' codes, are being developed to simulate the scenario of a hypothetical severe accident in a light water reactor, from the initial event until the possible radiological release of fission products out of the containment. They couple the predominant physical phenomena that occur in the different reactor zones and simulate the actuation of safety systems by procedures and by operators. In order to allow to study a great number of scenarios, a compromise must be found between precision of results and calculation time: one day of accident time should take less than one day of real time to simulate on a PC computer. This search of compromise is a real challenge for such integral codes. The development of the ASTEC integral code was initiated jointly by IRSN and GRS as an international reference code. The latest version 1.0 of ASTEC, including the new modules CESAR and DIVA which model the behaviour of the reactor cooling system and the core degradation, is presented here. Validation of the modules and one plant application are described

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

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

User's manual for a process model code

International Nuclear Information System (INIS)

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

1981-03-01

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

Thermal-hydraulic and aerosol containment phenomena modelling in ASTEC severe accident computer code

International Nuclear Information System (INIS)

Kljenak, Ivo; Dapper, Maik; Dienstbier, Jiri; Herranz, Luis E.; Koch, Marco K.; Fontanet, Joan

2010-01-01

Transients in containment systems of different scales (Phebus.FP containment, KAEVER vessel, Battelle Model Containment, LACE vessel and VVER-1000 nuclear power plant containment) involving thermal-hydraulic phenomena and aerosol behaviour, were simulated with the computer integral code ASTEC. The results of the simulations in the first four facilities were compared with experimental results, whereas the results of the simulated accident in the VVER-1000 containment were compared to results obtained with the MELCOR code. The main purpose of the simulations was the validation of the CPA module of the ASTEC code. The calculated results support the applicability of the code for predicting in-containment thermal-hydraulic and aerosol phenomena during a severe accident in a nuclear power plant.

The GNASH preequilibrium-statistical nuclear model code

International Nuclear Information System (INIS)

Arthur, E. D.

1988-01-01

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

Description of code system PLES/PTS for evaluation of pressure vessel integrity during PTS events

International Nuclear Information System (INIS)

Hirano, Masashi; Kohsaka, Atsuo.

1992-02-01

A code system PLES/PTS has been developed at the Japan Atomic Energy Research Institute (JAERI) to evaluate the integrity of the pressure vessel during plant thermal-hydraulic transients related to pressurized thermal shock (PTS) in a pressurized water reactor (PWR). The code system consists of several member codes to analyse the thermal-mixing behavior of emergency core cooling (ECC) water and primary coolant, transient stress distribution within the vessel wall, and crack growth behavior at the inner surface of the vessel. The crack growth behavior is evaluated by comparing the stress intensity factor (k I ) with the crack initiation toughness (k Ic ) and crack arrest toughness (k Ic ), taking into account the fast neutron irradiation embrittlement. This report describes the methods and models applied in PLES/PTS and the input data requirements. (author)

CSNI Integral Test Facility Matrices for Validation of Best-Estimate Thermal-Hydraulic Computer Codes

International Nuclear Information System (INIS)

Glaeser, H.

2008-01-01

Internationally agreed Integral Test Facility (ITF) matrices for validation of realistic thermal hydraulic system computer codes were established. ITF development is mainly for Pressurised Water Reactors (PWRs) and Boiling Water Reactors (BWRs). A separate activity was for Russian Pressurised Water-cooled and Water-moderated Energy Reactors (WWER). Firstly, the main physical phenomena that occur during considered accidents are identified, test types are specified, and test facilities suitable for reproducing these aspects are selected. Secondly, a list of selected experiments carried out in these facilities has been set down. The criteria to achieve the objectives are outlined. In this paper some specific examples from the ITF matrices will also be provided. The matrices will be a guide for code validation, will be a basis for comparisons of code predictions performed with different system codes, and will contribute to the quantification of the uncertainty range of code model predictions. In addition to this objective, the construction of such a matrix is an attempt to record information which has been generated around the world over the last years, so that it is more accessible to present and future workers in that field than would otherwise be the case.

The Fireball integrated code package

Energy Technology Data Exchange (ETDEWEB)

Dobranich, D.; Powers, D.A.; Harper, F.T.

1997-07-01

Many deep-space satellites contain a plutonium heat source. An explosion, during launch, of a rocket carrying such a satellite offers the potential for the release of some of the plutonium. The fireball following such an explosion exposes any released plutonium to a high-temperature chemically-reactive environment. Vaporization, condensation, and agglomeration processes can alter the distribution of plutonium-bearing particles. The Fireball code package simulates the integrated response of the physical and chemical processes occurring in a fireball and the effect these processes have on the plutonium-bearing particle distribution. This integrated treatment of multiple phenomena represents a significant improvement in the state of the art for fireball simulations. Preliminary simulations of launch-second scenarios indicate: (1) most plutonium vaporization occurs within the first second of the fireball; (2) large non-aerosol-sized particles contribute very little to plutonium vapor production; (3) vaporization and both homogeneous and heterogeneous condensation occur simultaneously; (4) homogeneous condensation transports plutonium down to the smallest-particle sizes; (5) heterogeneous condensation precludes homogeneous condensation if sufficient condensation sites are available; and (6) agglomeration produces larger-sized particles but slows rapidly as the fireball grows.

Rapid installation of numerical models in multiple parent codes

Energy Technology Data Exchange (ETDEWEB)

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

1996-10-01

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

MathModelica - An Extensible Modeling and Simulation Environment with Integrated Graphics and Literate Programming

OpenAIRE

Fritzson, Peter; Gunnarsson, Johan; Jirstrand, Mats

2002-01-01

MathModelica is an integrated interactive development environment for advanced system modeling and simulation. The environment integrates Modelica-based modeling and simulation with graphic design, advanced scripting facilities, integration of program code, test cases, graphics, documentation, mathematical type setting, and symbolic formula manipulation provided via Mathematica. The user interface consists of a graphical Model Editor and Notebooks. The Model Editor is a graphical user interfa...

A self-organized internal models architecture for coding sensory-motor schemes

Directory of Open Access Journals (Sweden)

Esaú eEscobar Juárez

2016-04-01

Full Text Available Cognitive robotics research draws inspiration from theories and models on cognition, as conceived by neuroscience or cognitive psychology, to investigate biologically plausible computational models in artificial agents. In this field, the theoretical framework of Grounded Cognition provides epistemological and methodological grounds for the computational modeling of cognition. It has been stressed in the literature that textit{simulation}, textit{prediction}, and textit{multi-modal integration} are key aspects of cognition and that computational architectures capable of putting them into play in a biologically plausible way are a necessity.Research in this direction has brought extensive empirical evidencesuggesting that textit{Internal Models} are suitable mechanisms forsensory-motor integration. However, current Internal Models architectures show several drawbacks, mainly due to the lack of a unified substrate allowing for a true sensory-motor integration space, enabling flexible and scalable ways to model cognition under the embodiment hypothesis constraints.We propose the Self-Organized Internal ModelsArchitecture (SOIMA, a computational cognitive architecture coded by means of a network of self-organized maps, implementing coupled internal models that allow modeling multi-modal sensory-motor schemes. Our approach addresses integrally the issues of current implementations of Internal Models.We discuss the design and features of the architecture, and provide empirical results on a humanoid robot that demonstrate the benefits and potentialities of the SOIMA concept for studying cognition in artificial agents.

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)

Integrated fast ignition simulation of cone-guided target with three codes

Energy Technology Data Exchange (ETDEWEB)

Sakagami, H. [Hyogo Univ., Computer Engineering, Himeji, Hyogo (Japan); Johzaki, T.; Nagatomo, H.; Mima, K. [Osaka Univ., Institute of Laser Engineering, Suita, Osaka (Japan)

2004-07-01

It was reported that the fuel core was heated up to {approx} 0.8 keV in the fast ignition experiments with cone-guided targets, but they could not theoretically explain heating mechanisms and achievement of such high temperature. Thus simulations should play an important role in estimating the scheme performance, and we must simulate each phenomenon with individual codes and integrate them under the Fast Ignition Integrated Interconnecting code project. In the previous integrated simulations, fast electrons generated by the laser-plasma interaction were too hot to efficiently heat the core and we got only a 0.096 keV temperature rise. Including the density gap at the contact surface between the cone tip and the imploded plasma, the period of core heating became longer and the core was heated by 0.162 keV, about 69% higher increment compared with ignoring the density gap effect. (authors)

Implementation of an enlarged model of the safety valves and relief in the plant integral model for the code RELAP/SCDAPSIM; Implementacion de un modelo ampliado de las valvulas de seguridad y alivio en el modelo integral de planta para el codigo RELAP/SCDAPSIM

Energy Technology Data Exchange (ETDEWEB)

Amador G, R.; Ortiz V, J.; Castillo D, R. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Hernandez L, E. J. [Instituto Tecnologico de Toluca, Av. Tecnologico s/n, Fracc. La Virgen, 52149 Metepec, Estado de Mexico (Mexico); Galeana R, J. C. [Universidad del Valle de Mexico, Campus Toluca, Av. de Las Palmas 136, 52140 Metepec, Estado de Mexico (Mexico); Gutierrez, V. H., E-mail: rodolfo.amador@inin.gob.mx [Universidad Autonoma Metropolitana, Unidad Azcapotzalco, Av. San Pablo 180, Col. Reynosa Tamaulipas, 02200 Mexico D. F. (Mexico)

2013-10-15

The present work refers to the implementation of a new model on the logic of the safety valves and relief in the integral model of the Nuclear Power Plant of Laguna Verde of the thermal-hydraulic compute code RELAP/SCDAPSIM Mod. 3.4. The new model was developed with the compute package SIMULINK-MATLAB and contemplates all the operation options of the safety valves and relief, besides including the availability options of the valves in all the operation ways and of blockage in the ways of relief and low-low. The implementation means the elimination of the old model of the safety valves and to analyze the group of logical variables, of discharge and available control systems to associate them to the model of package SIMULINK-MATLAB. The implementation has been practically transparent and 27 cases corresponding to a turbine discharge were analyzed with the code RELAP/SCDAPSIM Mod. 3.4. The results were satisfactory. (Author)

Integrated design approach of the pebble bed modular using models

International Nuclear Information System (INIS)

Venter, P.J.

2005-01-01

The Pebble Bed Modular Reactor (PBMR) is the first pebble bed reactor that will be utilised in a high temperature direct Brayton cycle configuration. This implies that there are a number of unique features in the PBMR that extend from the German experience base. One of the challenges in the design of the PBMR is managing the integrated design process between the designers, the physicists and the analysts. This integrated design process is managed through model-based development work. Three-dimensional CAD models are constructed of the components and parts in the reactor. From the CAD models, CFD models, neutronic models, shielding models, FEM models and other thermodynamic models are derived. These models range from very simple models to extremely detailed and complex models. The models are used in legacy software as well as commercial off-the-shelf software. The different models are also used in code-to-code comparisons to verify the results. This paper will briefly discuss the different models and the interaction between the models, showing the iterative design process that is used in the development of the reactor at PBMR. (author)

Experimental validation of TASS/SMR-S critical flow model for the integral reactor SMART

Energy Technology Data Exchange (ETDEWEB)

Seo, Si Won; Ra, In Sik; Kim, Kun Yeup [ACT Co., Daejeon (Korea, Republic of); Chung, Young Jong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-05-15

An advanced integral PWR, SMART (System- Integrated Modular Advanced ReacTor) is being developed in KAERI. It has a compact size and a relatively small power rating (330MWt) compared to a conventional reactor. Because new concepts are applied to SMART, an experimental and analytical validation is necessary for the safety evaluation of SMART. The analytical safety validation is being accomplished by a safety analysis code for an integral reactor, TASS/SMR-S developed by KAERI. TASS/SMR-S uses a lumped parameter one dimensional node and path modeling for the thermal hydraulic calculation and it uses point kinetics for the reactor power calculation. It has models for a general usage such as a core heat transfer model, a wall heat structure model, a critical flow model, component models, and it also has many SMART specific models such as an once through helical coiled steam generator model, and a condensate heat transfer model. To ensure that the TASS/SMR-S code has the calculation capability for the safety evaluation of SMART, the code should be validated for the specific models with the separate effect test experimental results. In this study, TASS/SMR-S critical flow model is evaluated as compared with SMD (Super Moby Dick) experiment

Steam condensation modelling in aerosol codes

International Nuclear Information System (INIS)

Dunbar, I.H.

1986-01-01

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

Recent developments in the CONTAIN-LMR code

International Nuclear Information System (INIS)

Murata, K.K.

1990-01-01

Through an international collaborative effort, a special version of the CONTAIN code is being developed for integrated mechanistic analysis of the conditions in liquid metal reactor (LMR) containments during severe accidents. The capabilities of the most recent code version, CONTAIN LMR/1B-Mod.1, are discussed. These include new models for the treatment of two condensables, sodium condensation on aerosols, chemical reactions, hygroscopic aerosols, and concrete outgassing. This code version also incorporates all of the previously released LMR model enhancements. The results of an integral demonstration calculation of a sever core-melt accident scenario are given to illustrate the features of this code version. 11 refs., 7 figs., 1 tab

Present capabilities and new developments in antenna modeling with the numerical electromagnetics code NEC

Energy Technology Data Exchange (ETDEWEB)

Burke, G.J.

1988-04-08

Computer modeling of antennas, since its start in the late 1960's, has become a powerful and widely used tool for antenna design. Computer codes have been developed based on the Method-of-Moments, Geometrical Theory of Diffraction, or integration of Maxwell's equations. Of such tools, the Numerical Electromagnetics Code-Method of Moments (NEC) has become one of the most widely used codes for modeling resonant sized antennas. There are several reasons for this including the systematic updating and extension of its capabilities, extensive user-oriented documentation and accessibility of its developers for user assistance. The result is that there are estimated to be several hundred users of various versions of NEC world wide. 23 refs., 10 figs.

SPIDERMAN: an open-source code to model phase curves and secondary eclipses

Science.gov (United States)

Louden, Tom; Kreidberg, Laura

2018-03-01

We present SPIDERMAN (Secondary eclipse and Phase curve Integrator for 2D tempERature MAppiNg), a fast code for calculating exoplanet phase curves and secondary eclipses with arbitrary surface brightness distributions in two dimensions. Using a geometrical algorithm, the code solves exactly the area of sections of the disc of the planet that are occulted by the star. The code is written in C with a user-friendly Python interface, and is optimised to run quickly, with no loss in numerical precision. Approximately 1000 models can be generated per second in typical use, making Markov Chain Monte Carlo analyses practicable. The modular nature of the code allows easy comparison of the effect of multiple different brightness distributions for the dataset. As a test case we apply the code to archival data on the phase curve of WASP-43b using a physically motivated analytical model for the two dimensional brightness map. The model provides a good fit to the data; however, it overpredicts the temperature of the nightside. We speculate that this could be due to the presence of clouds on the nightside of the planet, or additional reflected light from the dayside. When testing a simple cloud model we find that the best fitting model has a geometric albedo of 0.32 ± 0.02 and does not require a hot nightside. We also test for variation of the map parameters as a function of wavelength and find no statistically significant correlations. SPIDERMAN is available for download at https://github.com/tomlouden/spiderman.

SPIDERMAN: an open-source code to model phase curves and secondary eclipses

Science.gov (United States)

Louden, Tom; Kreidberg, Laura

2018-06-01

We present SPIDERMAN (Secondary eclipse and Phase curve Integrator for 2D tempERature MAppiNg), a fast code for calculating exoplanet phase curves and secondary eclipses with arbitrary surface brightness distributions in two dimensions. Using a geometrical algorithm, the code solves exactly the area of sections of the disc of the planet that are occulted by the star. The code is written in C with a user-friendly Python interface, and is optimized to run quickly, with no loss in numerical precision. Approximately 1000 models can be generated per second in typical use, making Markov Chain Monte Carlo analyses practicable. The modular nature of the code allows easy comparison of the effect of multiple different brightness distributions for the data set. As a test case, we apply the code to archival data on the phase curve of WASP-43b using a physically motivated analytical model for the two-dimensional brightness map. The model provides a good fit to the data; however, it overpredicts the temperature of the nightside. We speculate that this could be due to the presence of clouds on the nightside of the planet, or additional reflected light from the dayside. When testing a simple cloud model, we find that the best-fitting model has a geometric albedo of 0.32 ± 0.02 and does not require a hot nightside. We also test for variation of the map parameters as a function of wavelength and find no statistically significant correlations. SPIDERMAN is available for download at https://github.com/tomlouden/spiderman.

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

DEFF Research Database (Denmark)

Simonsen, Kent Inge

2014-01-01

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

C.A.S.H. - a transient integrated plant model for a HTR-module power plant. User manual

International Nuclear Information System (INIS)

Biesenbach, R.; Lauer, A.; Struth, S.

1997-07-01

The computer code C.A.S.H. has been developed as an integrated plant model for the HTR-Module reactor, in order to treat safety related questions about this type of power plant which require a detailed numeric simulation of the transient behaviour of the integrated plant. The present report contains the user manual for this plant model. It consists of three parts: In the first part, the code structure and functions, the course of the simulation calculations, and important code parts are described. The second part is devoted to the practical application and explains extensively the handling of the complex code system with several sample calculations. These computing cases comprise load-follow transients and the shutdown procedure of the HTR-Module and are presented and discussed with the full input data, job patterns, and numerous computer graphics. The third part contains the input manual of C.A.S.H. and is rather extensive as it includes the complete inputs of several reactor component computer codes along with the control program of the integrated plant model. (orig./DG) [de

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

Benchmarking of computer codes and approaches for modeling exposure scenarios

International Nuclear Information System (INIS)

Seitz, R.R.; Rittmann, P.D.; Wood, M.I.; Cook, J.R.

1994-08-01

The US Department of Energy Headquarters established a performance assessment task team (PATT) to integrate the activities of DOE sites that are preparing performance assessments for the disposal of newly generated low-level waste. The PATT chartered a subteam with the task of comparing computer codes and exposure scenarios used for dose calculations in performance assessments. This report documents the efforts of the subteam. Computer codes considered in the comparison include GENII, PATHRAE-EPA, MICROSHIELD, and ISOSHLD. Calculations were also conducted using spreadsheets to provide a comparison at the most fundamental level. Calculations and modeling approaches are compared for unit radionuclide concentrations in water and soil for the ingestion, inhalation, and external dose pathways. Over 30 tables comparing inputs and results are provided

Development of computer code in PNC, 3

International Nuclear Information System (INIS)

Ohtaki, Akira; Ohira, Hiroaki

1990-01-01

Super-COPD, a code which is integrated by calculation modules, has been developed in order to evaluate kinds of dynamics of LMFBR plant by improving COPD. The code involves all models and its advanced models of COPD in module structures. The code makes it possible to simulate the system dynamics of LMFBR plant of any configurations and components. (author)

Coding Model and Mapping Method of Spherical Diamond Discrete Grids Based on Icosahedron

Directory of Open Access Journals (Sweden)

LIN Bingxian

2016-12-01

Full Text Available Discrete Global Grid(DGG provides a fundamental environment for global-scale spatial data's organization and management. DGG's encoding scheme, which blocks coordinate transformation between different coordination reference frames and reduces the complexity of spatial analysis, contributes a lot to the multi-scale expression and unified modeling of spatial data. Compared with other kinds of DGGs, Diamond Discrete Global Grid(DDGG based on icosahedron is beneficial to the spherical spatial data's integration and expression for much better geometric properties. However, its structure seems more complicated than DDGG on octahedron due to its initial diamond's edges cannot fit meridian and parallel. New challenges are posed when it comes to the construction of hierarchical encoding system and mapping relationship with geographic coordinates. On this issue, this paper presents a DDGG's coding system based on the Hilbert curve and designs conversion methods between codes and geographical coordinates. The study results indicate that this encoding system based on the Hilbert curve can express space scale and location information implicitly with the similarity between DDG and planar grid put into practice, and balances efficiency and accuracy of conversion between codes and geographical coordinates in order to support global massive spatial data's modeling, integrated management and all kinds of spatial analysis.

Integrated predictive modeling simulations of the Mega-Amp Spherical Tokamak

International Nuclear Information System (INIS)

Nguyen, Canh N.; Bateman, Glenn; Kritz, Arnold H.; Akers, Robert; Byrom, Calum; Sykes, Alan

2002-01-01

Integrated predictive modeling simulations are carried out using the BALDUR transport code [Singer et al., Comput. Phys. Commun. 49, 275 (1982)] for high confinement mode (H-mode) and low confinement mode (L-mode) discharges in the Mega-Amp Spherical Tokamak (MAST) [Sykes et al., Phys. Plasmas 8, 2101 (2001)]. Simulation results, obtained using either the Multi-Mode transport model (MMM95) or, alternatively, the mixed-Bohm/gyro-Bohm transport model, are compared with experimental data. In addition to the anomalous transport, neoclassical transport is included in the simulations and the ion thermal diffusivity in the inner third of the plasma is found to be predominantly neoclassical. The sawtooth oscillations in the simulations radially spread the neutral beam injection heating profiles across a broad sawtooth mixing region. The broad sawtooth oscillations also flatten the central temperature and electron density profiles. Simulation results for the electron temperature and density profiles are compared with experimental data to test the applicability of these models and the BALDUR integrated modeling code in the limit of low aspect ratio toroidal plasmas

First step of the project for implementation of two non-symmetric cooling loops modeled by the ALMOD3 code

International Nuclear Information System (INIS)

Dominguez, L.; Camargo, C.T.M.

1984-09-01

The first step of the project for implementation of two non-symmetric cooling loops modeled by the ALMOD3 computer code is presented. This step consists of the introduction of a simplified model for simulating the steam generator. This model is the GEVAP computer code, integrant part of LOOP code, which simulates the primary coolant circuit of PWR nuclear power plants during transients. The ALMOD3 computer code has a model for the steam generator, called UTSG, which is very detailed. This model has spatial dependence, correlations for 2-phase flow, distinguished correlations for different heat transfer process. The GEVAP model has thermal equilibrium between phases (gaseous and liquid homogeneous mixture), no spatial dependence and uses only one generalized correlation to treat several heat transfer processes. (Author) [pt

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

Simulation of the containment spray system test PACOS PX2.2 with the integral code ASTEC and the containment code system COCOSYS

International Nuclear Information System (INIS)

Risken, Tobias; Koch, Marco K.

2011-01-01

The reactor safety research contains the analysis of postulated accidents in nuclear power plants (npp). These accidents may involve a loss of coolant from the nuclear plant's reactor coolant system, during which heat and pressure within the containment are increased. To handle these atmospheric conditions, containment spray systems are installed in various light water reactors (LWR) worldwide as a part of the accident management system. For the improvement and the safety ensurance in npp operation and accident management, numeric simulations of postulated accident scenarios are performed. The presented calculations regard the predictability of the containment spray system's effect with the integral code ASTEC and the containment code system COCOSYS, performed at Ruhr-Universitaet Bochum. Therefore the test PACOS Px2.2 is simulated, in which water is sprayed in the stratified containment atmosphere of the BMC (Battelle Modell-Containment). (orig.)

Time integration in the code Zgoubi and external usage of PTC's structures

International Nuclear Information System (INIS)

Forest, Etienne; Meot, F.

2006-06-01

The purpose of this note is to describe Zgoubi's integrator and to describe some pitfalls for time based integration when used in accelerators. We show why the convergence rate of an integrator can be affected by an improper treatment at the boundary when time is used as the integration variable. We also point out how the code PTC can be used as a container by other tracking engine. This work is not completed as far as incorporation of Zgoubi is concerned. (authors)

Using finite mixture models in thermal-hydraulics system code uncertainty analysis

Energy Technology Data Exchange (ETDEWEB)

Carlos, S., E-mail: scarlos@iqn.upv.es [Department d’Enginyeria Química i Nuclear, Universitat Politècnica de València, Camí de Vera s.n, 46022 València (Spain); Sánchez, A. [Department d’Estadística Aplicada i Qualitat, Universitat Politècnica de València, Camí de Vera s.n, 46022 València (Spain); Ginestar, D. [Department de Matemàtica Aplicada, Universitat Politècnica de València, Camí de Vera s.n, 46022 València (Spain); Martorell, S. [Department d’Enginyeria Química i Nuclear, Universitat Politècnica de València, Camí de Vera s.n, 46022 València (Spain)

2013-09-15

Highlights: • Best estimate codes simulation needs uncertainty quantification. • The output variables can present multimodal probability distributions. • The analysis of multimodal distribution is performed using finite mixture models. • Two methods to reconstruct output variable probability distribution are used. -- Abstract: Nuclear Power Plant safety analysis is mainly based on the use of best estimate (BE) codes that predict the plant behavior under normal or accidental conditions. As the BE codes introduce uncertainties due to uncertainty in input parameters and modeling, it is necessary to perform uncertainty assessment (UA), and eventually sensitivity analysis (SA), of the results obtained. These analyses are part of the appropriate treatment of uncertainties imposed by current regulation based on the adoption of the best estimate plus uncertainty (BEPU) approach. The most popular approach for uncertainty assessment, based on Wilks’ method, obtains a tolerance/confidence interval, but it does not completely characterize the output variable behavior, which is required for an extended UA and SA. However, the development of standard UA and SA impose high computational cost due to the large number of simulations needed. In order to obtain more information about the output variable and, at the same time, to keep computational cost as low as possible, there has been a recent shift toward developing metamodels (model of model), or surrogate models, that approximate or emulate complex computer codes. In this way, there exist different techniques to reconstruct the probability distribution using the information provided by a sample of values as, for example, the finite mixture models. In this paper, the Expectation Maximization and the k-means algorithms are used to obtain a finite mixture model that reconstructs the output variable probability distribution from data obtained with RELAP-5 simulations. Both methodologies have been applied to a separated

The nuclear reaction model code MEDICUS

International Nuclear Information System (INIS)

Ibishia, A.I.

2008-01-01

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

CMCpy: Genetic Code-Message Coevolution Models in Python

Science.gov (United States)

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

2013-01-01

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

24 CFR 200.925c - Model codes.

Science.gov (United States)

2010-04-01

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

The European Integrated Tokamak Modelling (ITM) effort: achievements and first physics results

International Nuclear Information System (INIS)

Falchetto, G.L.; Nardon, E.; Artaud, J.F.; Basiuk, V.; Huynh, Ph.; Imbeaux, F.; Coster, D.; Scott, B.D.; Coelho, R.; Alves, L.L.; Bizarro, João P.S.; Ferreira, J.; Figueiredo, A.; Figini, L.; Nowak, S.; Farina, D.; Kalupin, D.; Boulbe, C.; Faugeras, B.; Dinklage, A.

2014-01-01

A selection of achievements and first physics results are presented of the European Integrated Tokamak Modelling Task Force (EFDA ITM-TF) simulation framework, which aims to provide a standardized platform and an integrated modelling suite of validated numerical codes for the simulation and prediction of a complete plasma discharge of an arbitrary tokamak. The framework developed by the ITM-TF, based on a generic data structure including both simulated and experimental data, allows for the development of sophisticated integrated simulations (workflows) for physics application. The equilibrium reconstruction and linear magnetohydrodynamic (MHD) stability simulation chain was applied, in particular, to the analysis of the edge MHD stability of ASDEX Upgrade type-I ELMy H-mode discharges and ITER hybrid scenario, demonstrating the stabilizing effect of an increased Shafranov shift on edge modes. Interpretive simulations of a JET hybrid discharge were performed with two electromagnetic turbulence codes within ITM infrastructure showing the signature of trapped-electron assisted ITG turbulence. A successful benchmark among five EC beam/ray-tracing codes was performed in the ITM framework for an ITER inductive scenario for different launching conditions from the equatorial and upper launcher, showing good agreement of the computed absorbed power and driven current. Selected achievements and scientific workflow applications targeting key modelling topics and physics problems are also presented, showing the current status of the ITM-TF modelling suite. (paper)

Dynamic benchmarking of simulation codes

International Nuclear Information System (INIS)

Henry, R.E.; Paik, C.Y.; Hauser, G.M.

1996-01-01

Computer simulation of nuclear power plant response can be a full-scope control room simulator, an engineering simulator to represent the general behavior of the plant under normal and abnormal conditions, or the modeling of the plant response to conditions that would eventually lead to core damage. In any of these, the underlying foundation for their use in analysing situations, training of vendor/utility personnel, etc. is how well they represent what has been known from industrial experience, large integral experiments and separate effects tests. Typically, simulation codes are benchmarked with some of these; the level of agreement necessary being dependent upon the ultimate use of the simulation tool. However, these analytical models are computer codes, and as a result, the capabilities are continually enhanced, errors are corrected, new situations are imposed on the code that are outside of the original design basis, etc. Consequently, there is a continual need to assure that the benchmarks with important transients are preserved as the computer code evolves. Retention of this benchmarking capability is essential to develop trust in the computer code. Given the evolving world of computer codes, how is this retention of benchmarking capabilities accomplished? For the MAAP4 codes this capability is accomplished through a 'dynamic benchmarking' feature embedded in the source code. In particular, a set of dynamic benchmarks are included in the source code and these are exercised every time the archive codes are upgraded and distributed to the MAAP users. Three different types of dynamic benchmarks are used: plant transients; large integral experiments; and separate effects tests. Each of these is performed in a different manner. The first is accomplished by developing a parameter file for the plant modeled and an input deck to describe the sequence; i.e. the entire MAAP4 code is exercised. The pertinent plant data is included in the source code and the computer

Integrated analysis for a small break LOCA in the IRIS reactor using MELCOR and RELAP5 codes

International Nuclear Information System (INIS)

Del Nevo, A.; Manfredini, A.; Oriolo, F.; Paci, S.; Oriani, L.

2004-01-01

The pressurized light water cooled, medium power (1000 MWt) IRIS (International Reactor Innovative and Secure) has been under development for four years by an international consortium of over 21 organizations from ten countries. The plant conceptual design was completed in 2001 and the preliminary design is nearing completion. The pre-application licensing process with NRC started in October, 2002 and IRIS is one of the designs considered by US utilities as part of the ESP (Early Site Permit) process. This paper's focus is on the use of well known computer codes for integrated (reactor vessel and containment) calculations of the IRIS response to a small break loss of coolant accident (LOCA). In IRIS, large break LOCA events are eliminated by the use of a layout configuration in which the reactor vessel contains all the reactor coolant system components including the core, control rod drive mechanisms, pressurizer, steam generators, and coolant pumps. Thus the IRIS configuration has no large loop piping; also, no pipes with a diameter greater than 0.1 meters are part of the reactor coolant system boundary. For small break LOCAs, IRIS features an innovative mitigation approach that is based on maintaining coolant inventory rather than designing high and low pressure injection systems to provide makeup coolant to the reactor to maintain core cooling. The novel IRIS approach requires development of evaluation models that are different from those used for the current generation of pressurized water reactors. An analysis of small break LOCAs for IRIS is documented in two companion papers, and has been developed using a preliminary evaluation model based on the explicit coupling of the RELAP5 and GOTHIC codes. The objective of this paper is to compare the results obtained via the coupled RELAP/GOTHIC code with different computational tools. A reference case from the preliminary IRIS safety assessment was selected, and the same small break LOCA sequence is analyzed using

Interfacial and Wall Transport Models for SPACE-CAP Code

International Nuclear Information System (INIS)

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

2009-01-01

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

Interfacial and Wall Transport Models for SPACE-CAP Code

Energy Technology Data Exchange (ETDEWEB)

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

2009-10-15

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

High burnup models in computer code fair

Energy Technology Data Exchange (ETDEWEB)

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

1997-08-01

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

High burnup models in computer code fair

International Nuclear Information System (INIS)

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

1997-01-01

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

OpenSWPC: an open-source integrated parallel simulation code for modeling seismic wave propagation in 3D heterogeneous viscoelastic media

Science.gov (United States)

Maeda, Takuto; Takemura, Shunsuke; Furumura, Takashi

2017-07-01

We have developed an open-source software package, Open-source Seismic Wave Propagation Code (OpenSWPC), for parallel numerical simulations of seismic wave propagation in 3D and 2D (P-SV and SH) viscoelastic media based on the finite difference method in local-to-regional scales. This code is equipped with a frequency-independent attenuation model based on the generalized Zener body and an efficient perfectly matched layer for absorbing boundary condition. A hybrid-style programming using OpenMP and the Message Passing Interface (MPI) is adopted for efficient parallel computation. OpenSWPC has wide applicability for seismological studies and great portability to allowing excellent performance from PC clusters to supercomputers. Without modifying the code, users can conduct seismic wave propagation simulations using their own velocity structure models and the necessary source representations by specifying them in an input parameter file. The code has various modes for different types of velocity structure model input and different source representations such as single force, moment tensor and plane-wave incidence, which can easily be selected via the input parameters. Widely used binary data formats, the Network Common Data Form (NetCDF) and the Seismic Analysis Code (SAC) are adopted for the input of the heterogeneous structure model and the outputs of the simulation results, so users can easily handle the input/output datasets. All codes are written in Fortran 2003 and are available with detailed documents in a public repository.[Figure not available: see fulltext.

ITER Dynamic Tritium Inventory Modeling Code

International Nuclear Information System (INIS)

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

2005-01-01

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

Numerical computation of molecular integrals via optimized (vectorized) FORTRAN code

International Nuclear Information System (INIS)

Scott, T.C.; Grant, I.P.; Saunders, V.R.

1997-01-01

The calculation of molecular properties based on quantum mechanics is an area of fundamental research whose horizons have always been determined by the power of state-of-the-art computers. A computational bottleneck is the numerical calculation of the required molecular integrals to sufficient precision. Herein, we present a method for the rapid numerical evaluation of molecular integrals using optimized FORTRAN code generated by Maple. The method is based on the exploitation of common intermediates and the optimization can be adjusted to both serial and vectorized computations. (orig.)

Application of an integrated PC-based neutronics code system to criticality safety

International Nuclear Information System (INIS)

Briggs, J.B.; Nigg, D.W.

1991-01-01

An integrated system of neutronics and radiation transport software suitable for operation in an IBM PC-class environment has been under development at the Idaho National Engineering Laboratory (INEL) for the past four years. Four modules within the system are particularly useful for criticality safety applications. Using the neutronics portion of the integrated code system, effective neutron multiplication values (k eff values) have been calculated for a variety of benchmark critical experiments for metal systems (Plutonium and Uranium), Aqueous Systems (Plutonium and Uranium) and LWR fuel rod arrays. A description of the codes and methods used in the analysis and the results of the benchmark critical experiments are presented in this paper. In general, excellent agreement was found between calculated and experimental results. (Author)

Development of the containment transient analysis code for the passive reactor

Energy Technology Data Exchange (ETDEWEB)

Hwang, Young Dong; Kim, Young In; Bae, Yoon Young; Chang, Moon Hi [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-05-01

This study was performed to develop the analysis tools for the passively cooled steel containment and to construct the integrated code system which can analyze a thermal hydraulic behavior of the containment and reactor system during a loss of coolant accident. The computer code CONTEMPT4/MOD5/PCCS was developed by incorporating the passive containment cooling models to the containment pressure and temperature transient analysis computer code CONTEMPT4/MOD5. The integrated reactor thermal hydraulic analysis code system for passive reactor was constructed by coupling the best estimate thermal hydraulic system analysis code RELAP5/MOD3 and CONTEMPT4/MOD5/PCCS through the process control method. In addition, to evaluate the applicability of the code the CONTEMPT4/MOD5/PCCS was applied to the SMART(System-Integrated Modular Advanced Reactor). The pressure and temperature transient following the small break LOCA of SMART was analysed by modeling the safeguard vessel using both the newly added passive containment cooling model and existing pool model. (author). 16 refs., 22 figs., 7 tabs.

Alternative conceptual models and codes for unsaturated flow in fractured tuff: Preliminary assessments for GWTT-95

International Nuclear Information System (INIS)

Ho, C.K.; Altman, S.J.; Arnold, B.W.

1995-09-01

Groundwater travel time (GWTT) calculations will play an important role in addressing site-suitability criteria for the potential high-level nuclear waste repository at Yucca Mountain,Nevada. In support of these calculations, Preliminary assessments of the candidate codes and models are presented in this report. A series of benchmark studies have been designed to address important aspects of modeling flow through fractured media representative of flow at Yucca Mountain. Three codes (DUAL, FEHMN, and TOUGH 2) are compared in these benchmark studies. DUAL is a single-phase, isothermal, two-dimensional flow simulator based on the dual mixed finite element method. FEHMN is a nonisothermal, multiphase, multidimensional simulator based primarily on the finite element method. TOUGH2 is anon isothermal, multiphase, multidimensional simulator based on the integral finite difference method. Alternative conceptual models of fracture flow consisting of the equivalent continuum model (ECM) and the dual permeability (DK) model are used in the different codes

Improved choked flow model for MARS code

International Nuclear Information System (INIS)

Chung, Moon Sun; Lee, Won Jae; Ha, Kwi Seok; Hwang, Moon Kyu

2002-01-01

Choked flow calculation is improved by using a new sound speed criterion for bubbly flow that is derived by the characteristic analysis of hyperbolic two-fluid model. This model was based on the notion of surface tension for the interfacial pressure jump terms in the momentum equations. Real eigenvalues obtained as the closed-form solution of characteristic polynomial represent the sound speed in the bubbly flow regime that agrees well with the existing experimental data. The present sound speed shows more reasonable result in the extreme case than the Nguyens did. The present choked flow criterion derived by the present sound speed is employed in the MARS code and assessed by using the Marviken choked flow tests. The assessment results without any adjustment made by some discharge coefficients demonstrate more accurate predictions of choked flow rate in the bubbly flow regime than those of the earlier choked flow calculations. By calculating the Typical PWR (SBLOCA) problem, we make sure that the present model can reproduce the reasonable transients of integral reactor system

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.

Modeling report of DYMOND code (DUPIC version)

International Nuclear Information System (INIS)

Park, Joo Hwan; Yacout, Abdellatif M.

2003-04-01

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

Modeling report of DYMOND code (DUPIC version)

Energy Technology Data Exchange (ETDEWEB)

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

2003-04-01

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

Content Coding of Psychotherapy Transcripts Using Labeled Topic Models.

Science.gov (United States)

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

2017-03-01

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

Hydrogen recycle modeling in transport codes

International Nuclear Information System (INIS)

Howe, H.C.

1979-01-01

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

Applications of ASTEC integral code on a generic CANDU 6

Energy Technology Data Exchange (ETDEWEB)

Radu, Gabriela, E-mail: gabriela.radu@nuclear.ro [Institute for Nuclear Research, Campului 1, 115400 Mioveni, Arges (Romania); Prisecaru, Ilie [Power Engineering Department, University “Politehnica” of Bucharest, 313 Splaiul Independentei, Bucharest (Romania)

2015-05-15

Highlights: • Short overview of the models included in the ASTEC MCCI module. • MEDICIS/CPA coupled calculations for a generic CANDU6 reactor. • Two cases taking into account different pool/concrete interface models. - Abstract: In case of a hypothetical severe accident in a nuclear power plant, the corium consisting of the molten reactor core and internal structures may flow onto the concrete floor of containment building. This would cause an interaction between the molten corium and the concrete (MCCI), in which the heat transfer from the hot melt to the concrete would cause the decomposition and the ablation of the concrete. The potential hazard of this interaction is the loss of integrity of the containment building and the release of fission products into the environment due to the possibility of a concrete foundation melt-through or containment over-pressurization by the gases produced from the decomposition of the concrete or by the inflammation of combustible gases. In the safety assessment of nuclear power plants, it is necessary to know the consequences of such a phenomenon. The paper presents an example of application of the ASTECv2 code to a generic CANDU6 reactor. This concerns the thermal-hydraulic behaviour of the containment during molten core–concrete interaction in the reactor vault. The calculations were carried out with the help of the MEDICIS MCCI module and the CPA containment module of ASTEC code coupled through a specific prediction–correction method, which consists in describing the heat exchanges with the vault walls and partially absorbent gases. Moreover, the heat conduction inside the vault walls is described. Two cases are presented in this paper taking into account two different heat transfer models at the pool/concrete interface and siliceous concrete. The corium pool configuration corresponds to a homogeneous configuration with a detailed description of the upper crust.

A 1ST Step Integration of the Restructured MELCOR for the MIDAS Computer Code

International Nuclear Information System (INIS)

Park, S. H.; Kim, D. H.; Cho, S. W.

2006-01-01

KAERI is developing a localized severe accident code, MIDAS, based on MELCOR. MELCOR uses pointer variables for a fixed-size storage management to save the data. It passes data through two depths, its meaning is not understandable by variable itself. So it is needed to understand the methods for data passing. This method deteriorates the readability, maintainability and portability of the code. As a most important process for a localized severe accident analysis code, it is needed convenient method for data handling. So, it has been used the new features in FORTRAN90 such as a dynamic allocation for the restructuring. The restructuring of the data saving and transferring method of the existing code makes it easy to understand the code. Before an entire restructuring of the code, a restructuring for each package was developed and tested. And then integration of each restructured package was being processed one by one. In this paper, the integrating scope includes the BUR, CF, CVH, DCH, EDF, ESF, MP, SPR, TF and TP packages. As most of them use data within each package and a few packages share data with other packages. The verification was done through comparing the results before and after the restructuring

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

Energy Technology Data Exchange (ETDEWEB)

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

2018-01-12

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

The Light-Water-Reactor Version of the URANUS Integral fuel-rod code

Energy Technology Data Exchange (ETDEWEB)

Labmann, K; Moreno, A

1977-07-01

The LWR version of the URANUS code, a digital computer programme for the thermal and mechanical analysis of fuel rods, is presented. Material properties are discussed and their effect on integral fuel rod behaviour elaborated via URANUS results for some carefully selected reference experiments. The numerical results do not represent post-irradiation analyses of in-pile experiments, they illustrate rather typical and diverse URANUS capabilities. The performance test shows that URANUS is reliable and efficient, thus the code is a most valuable tool in fuel rod analysis work. K. LaBmann developed the LWR version of the URANUS code, material properties were reviewed and supplied by A. Moreno. (Author) 41 refs.

The light-water-reactor version of the Uranus integral fuel-rod code

International Nuclear Information System (INIS)

Moreno, A.; Lassmann, K.

1977-01-01

The LWR of the Uranus code, a digital computer programme for the thermal and mechanical analysis of fuel rods, is presented. Material properties are discussed and their effect on integral fuel rod behaviour elaborated via Uranus results for some carefully selected reference experiments. The numerical results do not represent post-irradiation analysis of in-pile experiments, they illustrate rather typical and diverse Uranus capabilities. The performance test shows that Uranus is reliable and efficient, thus the code is a most valuable tool in fuel fod analysis work. K. Lassmann developed the LWR version of the Uranus code, material properties were reviewed and supplied by A. Moreno. (author)

40 CFR 194.23 - Models and computer codes.

Science.gov (United States)

2010-07-01

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

Cavitation Modeling in Euler and Navier-Stokes Codes

Science.gov (United States)

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

1993-01-01

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

Research and Design in Unified Coding Architecture for Smart Grids

Directory of Open Access Journals (Sweden)

Gang Han

2013-09-01

Full Text Available Standardized and sharing information platform is the foundation of the Smart Grids. In order to improve the dispatching center information integration of the power grids and achieve efficient data exchange, sharing and interoperability, a unified coding architecture is proposed. The architecture includes coding management layer, coding generation layer, information models layer and application system layer. Hierarchical design makes the whole coding architecture to adapt to different application environments, different interfaces, loosely coupled requirements, which can realize the integration model management function of the power grids. The life cycle and evaluation method of survival of unified coding architecture is proposed. It can ensure the stability and availability of the coding architecture. Finally, the development direction of coding technology of the Smart Grids in future is prospected.

Integrated computer codes for nuclear power plant severe accident analysis

Energy Technology Data Exchange (ETDEWEB)

Jordanov, I; Khristov, Y [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

1996-12-31

This overview contains a description of the Modular Accident Analysis Program (MAAP), ICARE computer code and Source Term Code Package (STCP). STCP is used to model TMLB sample problems for Zion Unit 1 and WWER-440/V-213 reactors. Comparison is made of STCP implementation on VAX and IBM systems. In order to improve accuracy, a double precision version of MARCH-3 component of STCP is created and the overall thermal hydraulics is modelled. Results of modelling the containment pressure, debris temperature, hydrogen mass are presented. 5 refs., 10 figs., 2 tabs.

Integrated computer codes for nuclear power plant severe accident analysis

International Nuclear Information System (INIS)

Jordanov, I.; Khristov, Y.

1995-01-01

This overview contains a description of the Modular Accident Analysis Program (MAAP), ICARE computer code and Source Term Code Package (STCP). STCP is used to model TMLB sample problems for Zion Unit 1 and WWER-440/V-213 reactors. Comparison is made of STCP implementation on VAX and IBM systems. In order to improve accuracy, a double precision version of MARCH-3 component of STCP is created and the overall thermal hydraulics is modelled. Results of modelling the containment pressure, debris temperature, hydrogen mass are presented. 5 refs., 10 figs., 2 tabs

Integrating the nursing management minimum data set into the logical observation identifier names and codes system.

Science.gov (United States)

Subramanian, Amarnath; Westra, Bonnie; Matney, Susan; Wilson, Patricia S; Delaney, Connie W; Huff, Stan; Huff, Stanley M; Huber, Diane

2008-11-06

This poster describes the process used to integrate the Nursing Management Minimum Data Set (NMMDS), an instrument to measure the nursing context of care, into the Logical Observation Identifier Names and Codes (LOINC) system to facilitate contextualization of quality measures. Integration of the first three of 18 elements resulted in 48 new codes including five panels. The LOINC Clinical Committee has approved the presented mapping for their next release.

An integrative model of patient-centeredness - a systematic review and concept analysis.

Directory of Open Access Journals (Sweden)

Isabelle Scholl

Full Text Available Existing models of patient-centeredness reveal a lack of conceptual clarity. This results in a heterogeneous use of the term, unclear measurement dimensions, inconsistent results regarding the effectiveness of patient-centered interventions, and finally in difficulties in implementing patient-centered care. The aim of this systematic review was to identify the different dimensions of patient-centeredness described in the literature and to propose an integrative model of patient-centeredness based on these results.Protocol driven search in five databases, combined with a comprehensive secondary search strategy. All articles that include a definition of patient-centeredness were eligible for inclusion in the review and subject to subsequent content analysis. Two researchers independently first screened titles and abstracts, then assessed full texts for eligibility. In each article the given definition of patient-centeredness was coded independently by two researchers. We discussed codes within the research team and condensed them into an integrative model of patient-centeredness.4707 records were identified through primary and secondary search, of which 706 were retained after screening of titles and abstracts. 417 articles (59% contained a definition of patient-centeredness and were coded. 15 dimensions of patient-centeredness were identified: essential characteristics of clinician, clinician-patient relationship, clinician-patient communication, patient as unique person, biopsychosocial perspective, patient information, patient involvement in care, involvement of family and friends, patient empowerment, physical support, emotional support, integration of medical and non-medical care, teamwork and teambuilding, access to care, coordination and continuity of care. In the resulting integrative model the dimensions were mapped onto different levels of care.The proposed integrative model of patient-centeredness allows different stakeholders to speak

Tardos fingerprinting codes in the combined digit model

NARCIS (Netherlands)

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

2009-01-01

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

Models and applications of the UEDGE code

International Nuclear Information System (INIS)

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

1996-09-01

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

COMPBRN III: a computer code for modeling compartment fires

International Nuclear Information System (INIS)

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

1986-07-01

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

Chemistry models in the Victoria code

International Nuclear Information System (INIS)

Grimley, A.J. III

1988-01-01

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

Integrated Numerical Experiments (INEX) and the Free-Electron Laser Physical Process Code (FELPPC)

International Nuclear Information System (INIS)

Thode, L.E.; Chan, K.C.D.; Schmitt, M.J.; McKee, J.; Ostic, J.; Elliott, C.J.; McVey, B.D.

1990-01-01

The strong coupling of subsystem elements, such as the accelerator, wiggler, and optics, greatly complicates the understanding and design of a free electron laser (FEL), even at the conceptual level. To address the strong coupling character of the FEL the concept of an Integrated Numerical Experiment (INEX) was proposed. Unique features of the INEX approach are consistency and numerical equivalence of experimental diagnostics. The equivalent numerical diagnostics mitigates the major problem of misinterpretation that often occurs when theoretical and experimental data are compared. The INEX approach has been applied to a large number of accelerator and FEL experiments. Overall, the agreement between INEX and the experiments is very good. Despite the success of INEX, the approach is difficult to apply to trade-off and initial design studies because of the significant manpower and computational requirements. On the other hand, INEX provides a base from which realistic accelerator, wiggler, and optics models can be developed. The Free Electron Laser Physical Process Code (FELPPC) includes models developed from INEX, provides coupling between the subsystem models, and incorporates application models relevant to a specific trade-off or design study. In other words, FELPPC solves the complete physical process model using realistic physics and technology constraints. Because FELPPC provides a detailed design, a good estimate for the FEL mass, cost, and size can be made from a piece-part count of the FEL. FELPPC requires significant accelerator and FEL expertise to operate. The code can calculate complex FEL configurations including multiple accelerator and wiggler combinations

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

International Nuclear Information System (INIS)

Peak, R.D.

1983-01-01

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

A simplified model of Passive Containment Cooling System in a CFD code

International Nuclear Information System (INIS)

Jiang, X.W.; Studer, E.; Kudriakov, S.

2013-01-01

Highlights: ► We have built a condensing model using Navier–Stokes equations in CAST3M code. ► We have done a benchmark work on the condensing model using the COPAIN tests data. ► We have built an evaporating model according to Aiello's model in CAST3M code. ► We used Kang and Park's film evaporation tests data to validate the model. ► An integrated model was derived by coupling two individual models with a steel plate. -- Abstract: In this paper, we built up a simplified model of the Passive Containment Cooling System in a CFD code, including a steel plate, a condensing channel and an evaporating channel. In the inner side of the plate, the condensing channel is supposed to be the source of heat transfer into the steel plate. Along the outer side, an evaporating falling film is used to extract the heat from the steel plate. Upward flow of air is also considered along the evaporating film. In the condensing channel, a flow solver based on an asymptotic model of the Navier–Stokes equations at the low Mach number regime and two turbulence models (Buleev's model and Chien's k–ε model) are considered. The condensing channel model was used to model the COPAIN test, the computed heat flux and condensation rate were compared with the experimental data. In the evaporating channel, a simplified model developed by Aiello and Ciofalo (2009) was used, which considered the heat and mass balance between the falling film and the ascending air flow. The model was validated for two cases: a dry wall case and a completely wet wall case. In the former case, the results were compared with 2D predictions obtained by using the CFX-4 CFD code. In the latter case, the results were compared with experimental data obtained by Kang and Park. The comparison showed a satisfactory agreement on heat transfer rates, despite some overprediction depending on the air velocity. At the end, the condensing channel model and the evaporating channel model were coupled by the steel plate

Integrated design approach of the pebble BeD modular reactor using models

International Nuclear Information System (INIS)

Venter, Pieter J.; Mitchell, Mark N.

2007-01-01

The pebble bed modular reactor (PBMR) is the first pebble bed reactor that will be utilised in a high temperature direct Brayton cycle configuration. This implies that there are a number of unique features in the PBMR that extend from the German experience base. One of the challenges in the design of the PBMR is developing an understanding of the expected behaviour of the reactor through analyses and simulations and managing the integrated design process between the designers, the physicists and the analysts. This integrated design process is managed through model-based development work. Three-dimensional CAD models are constructed of the components and parts in the reactor. From the CAD models, CFD models, neutronic models, shielding models, FEM models and other thermodynamic models are derived. These models range from very simple models to extremely detailed and complex models. The models are used in legacy software as well as commercial off-the-shelf software. The different models are also used in code-to-code comparisons to verify the results. This paper will briefly discuss the different models and the interaction between the models, and how the models are used in the iterative design process that is used in the development of the reactor at PBMR

Integrative multicellular biological modeling: a case study of 3D epidermal development using GPU algorithms

Directory of Open Access Journals (Sweden)

Christley Scott

2010-08-01

Full Text Available Abstract Background Simulation of sophisticated biological models requires considerable computational power. These models typically integrate together numerous biological phenomena such as spatially-explicit heterogeneous cells, cell-cell interactions, cell-environment interactions and intracellular gene networks. The recent advent of programming for graphical processing units (GPU opens up the possibility of developing more integrative, detailed and predictive biological models while at the same time decreasing the computational cost to simulate those models. Results We construct a 3D model of epidermal development and provide a set of GPU algorithms that executes significantly faster than sequential central processing unit (CPU code. We provide a parallel implementation of the subcellular element method for individual cells residing in a lattice-free spatial environment. Each cell in our epidermal model includes an internal gene network, which integrates cellular interaction of Notch signaling together with environmental interaction of basement membrane adhesion, to specify cellular state and behaviors such as growth and division. We take a pedagogical approach to describing how modeling methods are efficiently implemented on the GPU including memory layout of data structures and functional decomposition. We discuss various programmatic issues and provide a set of design guidelines for GPU programming that are instructive to avoid common pitfalls as well as to extract performance from the GPU architecture. Conclusions We demonstrate that GPU algorithms represent a significant technological advance for the simulation of complex biological models. We further demonstrate with our epidermal model that the integration of multiple complex modeling methods for heterogeneous multicellular biological processes is both feasible and computationally tractable using this new technology. We hope that the provided algorithms and source code will be a

Effect of difference between group constants processed by codes TIMS and ETOX on integral quantities

International Nuclear Information System (INIS)

Takano, Hideki; Ishiguro, Yukio; Matsui, Yasushi.

1978-06-01

Group constants of 235 U, 238 U, 239 Pu, 240 Pu and 241 Pu have been produced with the processing code TIMS using the evaluated nuclear data of JENDL-1. The temperature and composition dependent self-shielding factors have been calculated for the two cases with and without considering mutual interference resonant nuclei. By using the group constants set produced by the TIMS code, the integral quantities, i.e. multiplication factor, Na-void reactivity effect and Doppler reactivity effect, are calculated and compared with those calculated with the use of the cross sections set produced by the ETOX code to evaluate accuracy of the approximate calculation method in ETOX. There is much difference in self-shielding factor in each energy group between the two codes. For the fast reactor assemblies under study, however, the integral quantities calculated with these two sets are in good agreement with each other, because of eventual cancelation of errors. (auth.)

Transmutation Fuel Performance Code Thermal Model Verification

Energy Technology Data Exchange (ETDEWEB)

Gregory K. Miller; Pavel G. Medvedev

2007-09-01

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

Advances in NLTE Modeling for Integrated Simulations

Energy Technology Data Exchange (ETDEWEB)

Scott, H A; Hansen, S B

2009-07-08

The last few years have seen significant progress in constructing the atomic models required for non-local thermodynamic equilibrium (NLTE) simulations. Along with this has come an increased understanding of the requirements for accurately modeling the ionization balance, energy content and radiative properties of different elements for a wide range of densities and temperatures. Much of this progress is the result of a series of workshops dedicated to comparing the results from different codes and computational approaches applied to a series of test problems. The results of these workshops emphasized the importance of atomic model completeness, especially in doubly excited states and autoionization transitions, to calculating ionization balance, and the importance of accurate, detailed atomic data to producing reliable spectra. We describe a simple screened-hydrogenic model that calculates NLTE ionization balance with surprising accuracy, at a low enough computational cost for routine use in radiation-hydrodynamics codes. The model incorporates term splitting, {Delta}n = 0 transitions, and approximate UTA widths for spectral calculations, with results comparable to those of much more detailed codes. Simulations done with this model have been increasingly successful at matching experimental data for laser-driven systems and hohlraums. Accurate and efficient atomic models are just one requirement for integrated NLTE simulations. Coupling the atomic kinetics to hydrodynamics and radiation transport constrains both discretizations and algorithms to retain energy conservation, accuracy and stability. In particular, the strong coupling between radiation and populations can require either very short timesteps or significantly modified radiation transport algorithms to account for NLTE material response. Considerations such as these continue to provide challenges for NLTE simulations.

Development of fast ignition integrated interconnecting code (FI3) for fast ignition scheme

International Nuclear Information System (INIS)

Nagatomo, H.; Johzaki, T.; Mima, K.; Sunahara, A.; Nishihara, K.; Izawa, Y.; Sakagami, H.; Nakao, Y.; Yokota, T.; Taguchi, T.

2005-01-01

The numerical simulation plays an important role in estimating the feasibility and performance of the fast ignition. There are two key issues in numerical analysis for the fast ignition. One is the controlling the implosion dynamics to form a high density core plasma in non-spherical implosion, and the other is heating core plasma efficiency by the short pulse high intense laser. From initial laser irradiation to final fusion burning, all the physics are coupling strongly in any phase, and they must be solved consistently in computational simulation. However, in general, it is impossible to simulate laser plasma interaction and radiation hydrodynamics in a single computational code, without any numerical dissipation, special assumption or conditional treatment. Recently, we have developed 'Fast Ignition Integrated Interconnecting code' (FI 3 ) which consists of collective Particle-in-Cell code, Relativistic Fokker-Planck hydro code, and 2-dimensional radiation hydrodynamics code. And those codes are connecting with each other in data-flow bases. In this paper, we will present detail feature of the FI 3 code, and numerical results of whole process of fast ignition. (author)

Integrity evaluation for stud female threads on pressure vessel according to ASME code using FEM

International Nuclear Information System (INIS)

Kim, Moon Young; Chung, Nam Yong

2003-01-01

The extension of design life among power plants is increasingly becoming a world-wide trend. Kori no.1 unit in Korea is operating two cycle. It has two man-ways for tube inspection in a steam generator which is one of the important components in a nuclear power plant. Especially, stud bolts for man-way cover have damaged by disassembly and assembly several times and degradation for bolt materials for long term operation. It should be evaluated and compared by ASME code criteria for integrity evaluation. Integrity evaluation criteria which has been made by the manufacturer is not applied on the stud bolts of nuclear pressure vessels directly because it is controlled by the yield stress of ASME code. It can apply evaluation criteria through FEM analysis to damaged female threads and to evaluated safety for helical-coil method which is used according to code case-N-496-1. From analysis results, we found that it is the same results between stress intensity which got from FEM analysis on damaged female threads over 10% by manufacture integrity criteria and 2/3 yield strength criteria on ASME code. It was also confirmed that the helical-coil repair method would be safe

ASTEC—the Aarhus STellar Evolution Code

Science.gov (United States)

Christensen-Dalsgaard, Jørgen

2008-08-01

The Aarhus code is the result of a long development, starting in 1974, and still ongoing. A novel feature is the integration of the computation of adiabatic oscillations for specified models as part of the code. It offers substantial flexibility in terms of microphysics and has been carefully tested for the computation of solar models. However, considerable development is still required in the treatment of nuclear reactions, diffusion and convective mixing.

Improving the quality of clinical coding: a comprehensive audit model

Directory of Open Access Journals (Sweden)

Hamid Moghaddasi

2014-04-01

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

Linking the plasma code EDGE2D to the neutral code NIMBUS for a self consistent transport model of the boundary

International Nuclear Information System (INIS)

De Matteis, A.

1987-01-01

This report describes the fully automatic linkage between the finite difference, two-dimensional code EDGE2D, based on the classical Braginskii partial differential equations of ion transport, and the Monte Carlo code NIMBUS, which solves the integral form of the stationary, linear Boltzmann equation for neutral transport in a plasma. The coupling has been performed for the real poloidal geometry of JET with two belt-limiters and real magnetic configurations with or without a single-null point. The new integrated system starts from the magnetic geometry computed by predictive or interpretative equilibrium codes and yields the plasma and neutrals characteristics in the edge

Conservation of concrete structures according to fib Model Code 2010

NARCIS (Netherlands)

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

2013-01-01

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

Modelling of the Gadolinium Fuel Test IFA-681 using the BISON Code

Energy Technology Data Exchange (ETDEWEB)

Pastore, Giovanni [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Williamson, Richard L [Idaho National Laboratory

2016-05-01

In this work, application of Idaho National Laboratory’s fuel performance code BISON to modelling of fuel rods from the Halden IFA-681 gadolinium fuel test is presented. First, an overview is given of BISON models, focusing on UO2/UO2-Gd2O3 fuel and Zircaloy cladding. Then, BISON analyses of selected fuel rods from the IFA-681 test are performed. For the first time in a BISON application to integral fuel rod simulations, the analysis is informed by detailed neutronics calculations in order to accurately capture the radial power profile throughout the fuel, which is strongly affected by the complex evolution of absorber Gd isotopes. In particular, radial power profiles calculated at IFE–Halden Reactor Project with the HELIOS code are used. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project. Some slide have been added as an Appendix to present the newly developed PolyPole-1 algorithm for modeling of intra-granular fission gas release.

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

Challenge problem and milestones for : Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC).

Energy Technology Data Exchange (ETDEWEB)

Freeze, Geoffrey A.; Wang, Yifeng; Howard, Robert; McNeish, Jerry A.; Schultz, Peter Andrew; Arguello, Jose Guadalupe, Jr.

2010-09-01

This report describes the specification of a challenge problem and associated challenge milestones for the Waste Integrated Performance and Safety Codes (IPSC) supporting the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The NEAMS challenge problems are designed to demonstrate proof of concept and progress towards IPSC goals. The goal of the Waste IPSC is to develop an integrated suite of modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. To demonstrate proof of concept and progress towards these goals and requirements, a Waste IPSC challenge problem is specified that includes coupled thermal-hydrologic-chemical-mechanical (THCM) processes that describe (1) the degradation of a borosilicate glass waste form and the corresponding mobilization of radionuclides (i.e., the processes that produce the radionuclide source term), (2) the associated near-field physical and chemical environment for waste emplacement within a salt formation, and (3) radionuclide transport in the near field (i.e., through the engineered components - waste form, waste package, and backfill - and the immediately adjacent salt). The initial details of a set of challenge milestones that collectively comprise the full challenge problem are also specified.

Challenge problem and milestones for: Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC)

International Nuclear Information System (INIS)

Freeze, Geoffrey A.; Wang, Yifeng; Howard, Robert; McNeish, Jerry A.; Schultz, Peter Andrew; Arguello, Jose Guadalupe Jr.

2010-01-01

This report describes the specification of a challenge problem and associated challenge milestones for the Waste Integrated Performance and Safety Codes (IPSC) supporting the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The NEAMS challenge problems are designed to demonstrate proof of concept and progress towards IPSC goals. The goal of the Waste IPSC is to develop an integrated suite of modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. To demonstrate proof of concept and progress towards these goals and requirements, a Waste IPSC challenge problem is specified that includes coupled thermal-hydrologic-chemical-mechanical (THCM) processes that describe (1) the degradation of a borosilicate glass waste form and the corresponding mobilization of radionuclides (i.e., the processes that produce the radionuclide source term), (2) the associated near-field physical and chemical environment for waste emplacement within a salt formation, and (3) radionuclide transport in the near field (i.e., through the engineered components - waste form, waste package, and backfill - and the immediately adjacent salt). The initial details of a set of challenge milestones that collectively comprise the full challenge problem are also specified.

A study on the dependency between turbulent models and mesh configurations of CFD codes

International Nuclear Information System (INIS)

Bang, Jungjin; Heo, Yujin; Jerng, Dong-Wook

2015-01-01

This paper focuses on the analysis of the behavior of hydrogen mixing and hydrogen stratification, using the GOTHIC code and the CFD code. Specifically, we examined the mesh sensitivity and how the turbulence model affects hydrogen stratification or hydrogen mixing, depending on the mesh configuration. In this work, sensitivity analyses for the meshes and the turbulence models were conducted for missing and stratification phenomena. During severe accidents in a nuclear power plants, the generation of hydrogen may occur and this will complicate the atmospheric condition of the containment by causing stratification of air, steam, and hydrogen. This could significantly impact containment integrity analyses, as hydrogen could be accumulated in local region. From this need arises the importance of research about stratification of gases in the containment. Two computation fluid dynamics code, i.e. GOTHIC and STAR-CCM+ were adopted and the computational results were benchmarked against the experimental data from PANDA facility. The main findings observed through the present work can be summarized as follows: 1) In the case of the GOTHIC code, it was observed that the aspect ratio of the mesh was found more important than the mesh size. Also, if the number of the mesh is over 3,000, the effects of the turbulence models were marginal. 2) For STAR-CCM+, the tendency is quite different from the GOTHIC code. That is, the effects of the turbulence models were small for fewer number of the mesh, however, as the number of mesh increases, the effects of the turbulence models becomes significant. Another observation is that away from the injection orifice, the role of the turbulence models tended to be important due to the nature of mixing process and inducted jet stream

A study on the dependency between turbulent models and mesh configurations of CFD codes

Energy Technology Data Exchange (ETDEWEB)

Bang, Jungjin; Heo, Yujin; Jerng, Dong-Wook [CAU, Seoul (Korea, Republic of)

2015-10-15

This paper focuses on the analysis of the behavior of hydrogen mixing and hydrogen stratification, using the GOTHIC code and the CFD code. Specifically, we examined the mesh sensitivity and how the turbulence model affects hydrogen stratification or hydrogen mixing, depending on the mesh configuration. In this work, sensitivity analyses for the meshes and the turbulence models were conducted for missing and stratification phenomena. During severe accidents in a nuclear power plants, the generation of hydrogen may occur and this will complicate the atmospheric condition of the containment by causing stratification of air, steam, and hydrogen. This could significantly impact containment integrity analyses, as hydrogen could be accumulated in local region. From this need arises the importance of research about stratification of gases in the containment. Two computation fluid dynamics code, i.e. GOTHIC and STAR-CCM+ were adopted and the computational results were benchmarked against the experimental data from PANDA facility. The main findings observed through the present work can be summarized as follows: 1) In the case of the GOTHIC code, it was observed that the aspect ratio of the mesh was found more important than the mesh size. Also, if the number of the mesh is over 3,000, the effects of the turbulence models were marginal. 2) For STAR-CCM+, the tendency is quite different from the GOTHIC code. That is, the effects of the turbulence models were small for fewer number of the mesh, however, as the number of mesh increases, the effects of the turbulence models becomes significant. Another observation is that away from the injection orifice, the role of the turbulence models tended to be important due to the nature of mixing process and inducted jet stream.

Fuel rod modelling during transients: The TOUTATIS code

International Nuclear Information System (INIS)

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

2001-01-01

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

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

International Nuclear Information System (INIS)

1995-08-01

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

Counter-part Test and Code Analysis of the Integral Test Loop, SNUF

International Nuclear Information System (INIS)

Park, Goon Cherl; Bae, B. U.; Lee, K. H.; Cho, Y. J.

2007-02-01

The thermal-hydraulic phenomena of Direct Vessel Injection (DVI) line Small-Break Loss-of-Coolant Accident (SBLOCA) in pressurized water reactor, APR1400, were investigated. The reduced-height and reduced-pressure integral test loop, SNUF (Seoul National University Facility), was constructed with scaling down the prototype. For the appropriate test conditions in the experiment of SNUF, the energy scaling methodology was suggested as scaling the coolant mass inventory and thermal power for the reduced-pressure condition. From the MARS code analysis, the energy scaling methodology was confirmed to show the reasonable transient when ideally scaled-down SNUF model was compared to the prototype model. In the experiments according to the conditions determined by energy scaling methodology, the phenomenon of downcomer seal clearing had a dominant role in decrease of the system pressure and increase of the coolant level of core. The experimental results was utilized to validate the calculation capability of MARS

Integrative structure modeling with the Integrative Modeling Platform.

Science.gov (United States)

Webb, Benjamin; Viswanath, Shruthi; Bonomi, Massimiliano; Pellarin, Riccardo; Greenberg, Charles H; Saltzberg, Daniel; Sali, Andrej

2018-01-01

Building models of a biological system that are consistent with the myriad data available is one of the key challenges in biology. Modeling the structure and dynamics of macromolecular assemblies, for example, can give insights into how biological systems work, evolved, might be controlled, and even designed. Integrative structure modeling casts the building of structural models as a computational optimization problem, for which information about the assembly is encoded into a scoring function that evaluates candidate models. Here, we describe our open source software suite for integrative structure modeling, Integrative Modeling Platform (https://integrativemodeling.org), and demonstrate its use. © 2017 The Protein Society.

MELMRK 2.0: A description of computer models and results of code testing

International Nuclear Information System (INIS)

Wittman, R.S.; Denny, V.; Mertol, A.

1992-01-01

An advanced version of the MELMRK computer code has been developed that provides detailed models for conservation of mass, momentum, and thermal energy within relocating streams of molten metallics during meltdown of Savannah River Site (SRS) reactor assemblies. In addition to a mechanistic treatment of transport phenomena within a relocating stream, MELMRK 2.0 retains the MOD1 capability for real-time coupling of the in-depth thermal response of participating assembly heat structure and, further, augments this capability with models for self-heating of relocating melt owing to steam oxidation of metallics and fission product decay power. As was the case for MELMRK 1.0, the MOD2 version offers state-of-the-art numerics for solving coupled sets of nonlinear differential equations. Principal features include application of multi-dimensional Newton-Raphson techniques to accelerate convergence behavior and direct matrix inversion to advance primitive variables from one iterate to the next. Additionally, MELMRK 2.0 provides logical event flags for managing the broad range of code options available for treating such features as (1) coexisting flow regimes, (2) dynamic transitions between flow regimes, and (3) linkages between heatup and relocation code modules. The purpose of this report is to provide a detailed description of the MELMRK 2.0 computer models for melt relocation. Also included are illustrative results for code testing, as well as an integrated calculation for meltdown of a Mark 31a assembly

Economic aspects and models for building codes

DEFF Research Database (Denmark)

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

It is the purpose of this bulletin to present an economic model for estimating the consequence of new or changed building codes. The object is to allow comparative analysis in order to improve the basis for decisions in this field. The model is applied in a case study.......It is the purpose of this bulletin to present an economic model for estimating the consequence of new or changed building codes. The object is to allow comparative analysis in order to improve the basis for decisions in this field. The model is applied in a case study....

Ultrafast method of calculating the dynamic spectral line shapes for integrated modelling of plasmas

International Nuclear Information System (INIS)

Lisitsa, V.S.

2009-01-01

An ultrafast code for spectral line shape calculations is presented to be used in the integrated modelling of plasmas. The code is based on the close analogy between two mechanisms: (i) Dicke narrowing of the Doppler-broadened spectral lines and (ii) transition from static to impact regime in the Stark broadening. The analogy makes it possible to describe the dynamic Stark broadening in terms of an analytical functional of the static line shape. A comparison of new method with the widely used Frequency Fluctuating Method (FFM) developed by the Marseille University group (B. Talin, R. Stamm, et al.) shows good agreement, with the new method being faster than the standard FFM by nearly two orders of magnitude. The method proposed may significantly simplify the radiation transport modeling and opens new possibilities for integrated modeling of the edge and divertor plasma in tokamaks. (author)

An integrative approach to predicting the functional effects of small indels in non-coding regions of the human genome.

Science.gov (United States)

Ferlaino, Michael; Rogers, Mark F; Shihab, Hashem A; Mort, Matthew; Cooper, David N; Gaunt, Tom R; Campbell, Colin

2017-10-06

Small insertions and deletions (indels) have a significant influence in human disease and, in terms of frequency, they are second only to single nucleotide variants as pathogenic mutations. As the majority of mutations associated with complex traits are located outside the exome, it is crucial to investigate the potential pathogenic impact of indels in non-coding regions of the human genome. We present FATHMM-indel, an integrative approach to predict the functional effect, pathogenic or neutral, of indels in non-coding regions of the human genome. Our method exploits various genomic annotations in addition to sequence data. When validated on benchmark data, FATHMM-indel significantly outperforms CADD and GAVIN, state of the art models in assessing the pathogenic impact of non-coding variants. FATHMM-indel is available via a web server at indels.biocompute.org.uk. FATHMM-indel can accurately predict the functional impact and prioritise small indels throughout the whole non-coding genome.

SIMIFR: A code to simulate material movement in the Integral Fast Reactor

International Nuclear Information System (INIS)

White, A.M.; Orechwa, Yuri.

1991-01-01

The SIMIFR code has been written to simulate the movement of material through a process. This code can be used to investigate inventory differences in material balances, assist in process design, and to produce operational scheduling. The particular process that is of concern to the authors is that centered around Argonne National Laboratory's Integral Fast Reactor. This is a process which involves the irradiation of fissile material for power production, and the recycling of the irradiated reactor fuel pins into fresh fuel elements. To adequately simulate this process it is necessary to allow for locations which can contain either discrete items or homogeneous mixtures. It is also necessary to allow for a very flexible process control algorithm. Further, the code must have the capability of transmuting isotopic compositions and computing internally the fraction of material from a process ending up in a given location. The SIMIFR code has been developed to perform all of these tasks. In addition to simulating the process, the code is capable of generating random measurement values and sampling errors for all locations, and of producing a restart deck so that terminated problems may be continued. In this paper the authors first familiarize the reader with the IFR fuel cycle. The different capabilities of the SIMIFR code are described. Finally, the simulation of the IFR fuel cycle using the SIMIFR code is discussed. 4 figs

Saint: a lightweight integration environment for model annotation.

Science.gov (United States)

Lister, Allyson L; Pocock, Matthew; Taschuk, Morgan; Wipat, Anil

2009-11-15

Saint is a web application which provides a lightweight annotation integration environment for quantitative biological models. The system enables modellers to rapidly mark up models with biological information derived from a range of data sources. Saint is freely available for use on the web at http://www.cisban.ac.uk/saint. The web application is implemented in Google Web Toolkit and Tomcat, with all major browsers supported. The Java source code is freely available for download at http://saint-annotate.sourceforge.net. The Saint web server requires an installation of libSBML and has been tested on Linux (32-bit Ubuntu 8.10 and 9.04).

Latest improvements on TRACPWR six-equations thermohydraulic code

International Nuclear Information System (INIS)

Rivero, N.; Batuecas, T.; Martinez, R.; Munoz, J.; Lenhardt, G.; Serrano, P.

1999-01-01

The paper presents the latest improvements on TRACPWR aimed at adapting the code to present trends on computer platforms, architectures and training requirements as well as extending the scope of the code itself and its applicability to other technologies different from Westinghouse PWR one. Firstly major features of TRACPWR as best estimate and real time simulation code are summed, then the areas where TRACPWR is being improved are presented. These areas comprising: (1) Architecture: integrating TRACPWR and RELAP5 codes, (2) Code scope enhancement: modelling the Mid-Loop operation, (3) Code speed-up: applying parallelization techniques, (4) Code platform downswing: porting to Windows N1 platform, (5) On-line performance: allowing simulation initialisation from a Plant Process Computer, and (6) Code scope extension: using the code for modelling VVER and PHWR technology. (author)

Data-driven modelling of structured populations a practical guide to the integral projection model

CERN Document Server

Ellner, Stephen P; Rees, Mark

2016-01-01

This book is a “How To” guide for modeling population dynamics using Integral Projection Models (IPM) starting from observational data. It is written by a leading research team in this area and includes code in the R language (in the text and online) to carry out all computations. The intended audience are ecologists, evolutionary biologists, and mathematical biologists interested in developing data-driven models for animal and plant populations. IPMs may seem hard as they involve integrals. The aim of this book is to demystify IPMs, so they become the model of choice for populations structured by size or other continuously varying traits. The book uses real examples of increasing complexity to show how the life-cycle of the study organism naturally leads to the appropriate statistical analysis, which leads directly to the IPM itself. A wide range of model types and analyses are presented, including model construction, computational methods, and the underlying theory, with the more technical material in B...

Fuel behavior modeling using the MARS computer code

International Nuclear Information System (INIS)

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

1983-01-01

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

Continuous integration in a social-coding world : empirical evidence from GitHub

NARCIS (Netherlands)

Vasilescu, B.N.; van Schuylenburg, S.B.; Wulms, Jules; Serebrenik, A.; Brand, van den M.G.J.

2014-01-01

Continuous integration is a software engineering practice of frequently merging all developer working copies with a shared main branch, e.g., several times a day. With the advent of GitHub, a platform well known for its "social coding" features that aid collaboration and sharing, and currently the

Development and validation of corium oxidation model for the VAPEX code

International Nuclear Information System (INIS)

Blinkov, V.N.; Melikhov, V.I.; Davydov, M.V.; Melikhov, O.I.; Borovkova, E.M.

2011-01-01

In light water reactor core melt accidents, the molten fuel (corium) can be brought into contact with coolant water in the course of the melt relocation in-vessel and ex-vessel as well as in an accident mitigation action of water addition. Mechanical energy release from such an interaction is of interest in evaluating the structural integrity of the reactor vessel as well as of the containment. Usually, the source for the energy release is considered to be the rapid transfer of heat from the molten fuel to the water ('vapor explosion'). When the fuel contains a chemically reactive metal component, there could be an additional source for the energy release, which is the heat release and hydrogen production due to the metal-water chemical reaction. In Electrogorsk Research and Engineering Center the computer code VAPEX (VAPor EXplosion) has been developed for analysis of the molten fuel coolant interaction. Multifield approach is used for modeling of dynamics of following phases: water, steam, melt jet, melt droplets, debris. The VAPEX code was successfully validated on FARO experimental data. Hydrogen generation was observed in FARO tests even though corium didn't contain metal component. The reason for hydrogen generation was not clear, so, simplified empirical model of hydrogen generation was implemented in the VAPEX code to take into account input of hydrogen into pressure increase. This paper describes new more detailed model of hydrogen generation due to the metal-water chemical reaction and results of its validation on ZREX experiments. (orig.)

Genetic coding and gene expression - new Quadruplet genetic coding model

Science.gov (United States)

Shankar Singh, Rama

2012-07-01

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

Development and Application of a Plant Code to the Analysis of Transients in Integrated Reactors

International Nuclear Information System (INIS)

Rabiti, A.; Gimenez, M.; Delmastro, D.; Zanocco, P.

2003-01-01

In this work, a secondary system model for a CAREM-25 type nuclear power plant was developed.A two-phase flow homogenous model was used and found adequate for the scope of the present work.A finite difference scheme was used for the numerical implementation of the model.This model was coupled to the HUARPE code, a primary circuit code, in order to obtain a plant code.This plant code was used to analyze the inherent response of the system, without control feedback loops, for a transient of steam generator feed-water mass flow reduction.The results obtained are satisfactory, but a validation against other plant codes is still necessary

Numerical modelling of the long-term evolution of EDZ. Development of material models, implementation in finite-element codes, and validation

International Nuclear Information System (INIS)

Pudewills, A.

2005-11-01

Construction of deep underground structures disturbs the initial stress field in the surrounding rock. This effect can generate microcracks and alter the hydromechanical properties of the rock salt around the excavations. For the long-term performance of an underground repository in rock salt, the evolution of the 'Excavation Disturbed Zone' (EDZ) and the hydromechanical behaviour of this zone represent important issues with respect to the integrity of the geological and technical barriers. Within the framework of the NF-PRO project, WP 4.4, attention focuses on the mathematical modelling of the development and evolution of the EDZ in the rock near a disposal drift due to its relevance on the integrity of the geological and technical barriers. To perform this task, finite-element codes containing a set of time- and temperature-dependent constitutive models have been improved. A new viscoplastic constitutive model for rock salt that can describe the damage of the rock has been implemented in the finite-element codes available. The model parameters were evaluated based on experimental results. Additionally, the long-term evolution of the EDZ around a gallery in a salt mine at about 700 m below the surface was analysed and the numerical results were compared with in-situ measurements. The calculated room closure, stress distribution and the increase of rock permeability in the EDZ were compared with in-situ data, thus providing confidence in the model used. (orig.)

Improvement of a combustion model in MELCOR code

International Nuclear Information System (INIS)

Ogino, Masao; Hashimoto, Takashi

1999-01-01

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

Implementation of non-condensable gases condensation suppression model into the WCOBRA/TRAC-TF2 LOCA safety evaluation code

Energy Technology Data Exchange (ETDEWEB)

Liao, J.; Cao, L.; Ohkawa, K.; Frepoli, C. [LOCA Integrated Services I, Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

The non-condensable gases condensation suppression model is important for a realistic LOCA safety analysis code. A condensation suppression model for direct contact condensation was previously developed by Westinghouse using first principles. The model is believed to be an accurate description of the direct contact condensation process in the presence of non-condensable gases. The Westinghouse condensation suppression model is further revised by applying a more physical model. The revised condensation suppression model is thus implemented into the WCOBRA/TRAC-TF2 LOCA safety evaluation code for both 3-D module (COBRA-TF) and 1-D module (TRAC-PF1). Parametric study using the revised Westinghouse condensation suppression model is conducted. Additionally, the performance of non-condensable gases condensation suppression model is examined in the ACHILLES (ISP-25) separate effects test and LOFT L2-5 (ISP-13) integral effects test. (authors)

LIANA Model Integration System - architecture, user interface design and application in MOIRA DSS

Directory of Open Access Journals (Sweden)

D. Hofman

2005-01-01

Full Text Available The LIANA Model Integration System is the shell application supporting model integration and user interface functionality required for the rapid construction and run-time support of the environmental decision support systems (EDSS. Internally it is constructed as the framework of C++ classes and functions covering most common tasks performed by the EDSS (such as managing of and alternative strategies, running of the chain of the models, supporting visualisation of the data with tables and graphs, keeping ranges and default values for input parameters etc.. EDSS is constructed by integration of LIANA system with the models or other applications such as GIS or MAA software. The basic requirements to the model or other application to be integrated is minimal - it should be a Windows or DOS .exe file and receive input and provide output as text files. For the user the EDSS is represented as the number of data sets describing scenario or giving results of evaluation of scenario via modelling. Internally data sets correspond to the I/O files of the models. During the integration the parameters included in each the data sets as well as specifications necessary to present the data set in GUI and export or import it to/from text file are provided with MIL_LIANA language. Visual C++ version of LIANA has been developed in the frame of MOIRA project and is used as the basis for the MOIRA Software Framework - the shell and user interface component of the MOIRA Decision Support System. At present, the usage of LIANA for the creation of a new EDSS requires changes to be made in its C++ code. The possibility to use LIANA for the new EDSS construction without extending the source code is achieved by substituting MIL_LIANA with the object-oriented LIANA language.

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-09-15

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

ETR/ITER systems code

Energy Technology Data Exchange (ETDEWEB)

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

1988-04-01

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

ETR/ITER systems code

International Nuclear Information System (INIS)

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

1988-04-01

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

CSNI Integral test facility validation matrix for the assessment of thermal-hydraulic codes for LWR LOCA and transients

International Nuclear Information System (INIS)

1996-07-01

This report deals with an internationally agreed integral test facility (ITF) matrix for the validation of best estimate thermal-hydraulic computer codes. Firstly, the main physical phenomena that occur during the considered accidents are identified, test types are specified, and test facilities suitable for reproducing these aspects are selected. Secondly, a life of selected experiments carried out in these facilities has been set down. The criteria to achieve the objectives are outlined. The construction of such a matrix is an attempt to collect together in a systematic way the best sets of openly available test data for code validation, assessment and improvement, including quantitative assessment of uncertainties in the modelling of phenomena by the codes. In addition to this objective, it is an attempt to record information which has been generated around the world over the last 20 years so that it is more accessible to present and future workers in that field than would otherwise be the case

Integrated Modelling - the next steps (Invited)

Science.gov (United States)

Moore, R. V.

2010-12-01

Integrated modelling (IM) has made considerable advances over the past decade but it has not yet been taken up as an operational tool in the way that its proponents had hoped. The reasons why will be discussed in Session U17. This talk will propose topics for a research and development programme and suggest an institutional structure which, together, could overcome the present obstacles. Their combined aim would be first to make IM into an operational tool useable by competent public authorities and commercial companies and, in time, to see it evolve into the modelling equivalent of Google Maps, something accessible and useable by anyone with a PC or an iphone and an internet connection. In a recent study, a number of government agencies, water authorities and utilities applied integrated modelling to operational problems. While the project demonstrated that IM could be used in an operational setting and had benefit, it also highlighted the advances that would be required for its widespread uptake. These were: greatly improving the ease with which models could be a) made linkable, b) linked and c) run; developing a methodology for applying integrated modelling; developing practical options for calibrating and validating linked models; addressing the science issues that arise when models are linked; extending the range of modelling concepts that can be linked; enabling interface standards to pass uncertainty information; making the interface standards platform independent; extending the range of platforms to include those for high performance computing; developing the concept of modelling components as web services; separating simulation code from the model’s GUI, so that all the results from the linked models can be viewed through a single GUI; developing scenario management systems so that that there is an audit trail of the version of each model and dataset used in each linked model run. In addition to the above, there is a need to build a set of integrated

Development of a computer code for the calculation of stellar evolution, with applications to solar models of low neutrino flux

International Nuclear Information System (INIS)

Newman, M.J.

1975-01-01

A general purpose computer code has been developed to allow the detailed calculation of evolutionary sequences of hydrostatic stellar models under many circumstances of astrophysical interest. Solution of the structure equations is by the relaxation technique throughout the star without explicit integration and fitting for the outer envelope. A new matrix method of algebraic solution of the finite difference equations is employed, together with a modification of that method for the treatment of the central boundary condition. The method is easily adapted to an integration technique for the construction of initial models. It is demonstrated how the matrix technique allows determination of the derivatives of the matching condition in a single integration. The modification of the code for the purpose of detailed evolutionary calculation of a portion of a star is presented through the modification of the boundary conditions to represent in simple fashion the remainder of the star. Stability and convergence problems encountered in earlier versions of the code are discussed, as well as the techniques used to overcome them. The structure of the code is highly modular, so as to easily accommodate changes in input physics. Following the ad hoc suggestion of Clayton (1974), the calculations were repeated with the high energy tail of the Maxwell distribution of relative ion velocities depleted by various amounts. As an example of the technique of evolving a portion of a star a second application to the solar neutrino problem is made

Modeling peripheral olfactory coding in Drosophila larvae.

Directory of Open Access Journals (Sweden)

Derek J Hoare

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

Improved Intra-coding Methods for H.264/AVC

Directory of Open Access Journals (Sweden)

Li Song

2009-01-01

Full Text Available The H.264/AVC design adopts a multidirectional spatial prediction model to reduce spatial redundancy, where neighboring pixels are used as a prediction for the samples in a data block to be encoded. In this paper, a recursive prediction scheme and an enhanced (block-matching algorithm BMA prediction scheme are designed and integrated into the state-of-the-art H.264/AVC framework to provide a new intra coding model. Extensive experiments demonstrate that the coding efficiency can be on average increased by 0.27 dB with comparison to the performance of the conventional H.264 coding model.

Data model description for the DESCARTES and CIDER codes

International Nuclear Information System (INIS)

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

1993-01-01

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

LMFBR models for the ORIGEN2 computer code

International Nuclear Information System (INIS)

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

1981-10-01

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

MARS CODE MANAUAL VOLUME IV - Developmental Assessment Report

International Nuclear Information System (INIS)

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

2010-02-01

Korea Advanced Energy Research Institute (KAERI) conceived and started the development of MARS code with the main objective of producing a state-of-the-art realistic thermal hydraulic systems analysis code with multi-dimensional analysis capability. MARS achieves this objective by very tightly integrating the one dimensional RELAP5/MOD3 with the multi-dimensional COBRA-TF codes. The method of integration of the two codes is based on the dynamic link library techniques, and the system pressure equation matrices of both codes are implicitly integrated and solved simultaneously. In addition, the Equation-Of-State (EOS) for the light water was unified by replacing the EOS of COBRA-TF by that of the RELAP5. This assessment manual provides a complete list of code assessment results of the MARS code for various conceptual problem, separate effect test and integral effect test. From these validation procedures, the soundness and accuracy of the MARS code has been confirmed. The overall structure of the input is modeled on the structure of the RELAP5 and as such the layout of the manual is very similar to that of the RELAP. This similitude to RELAP5 input is intentional as this input scheme will allow minimum modification between the inputs of RELAP5 and MARS3.1. MARS3.1 development team would like to express its appreciation to the RELAP5 Development Team and the USNRC for making this manual possible

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

Science.gov (United States)

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

1978-01-01

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

Repairing business process models as retrieved from source code

NARCIS (Netherlands)

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

2013-01-01

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

Research Integrity and Research Ethics in Professional Codes of Ethics: Survey of Terminology Used by Professional Organizations across Research Disciplines.

Science.gov (United States)

Komić, Dubravka; Marušić, Stjepan Ljudevit; Marušić, Ana

2015-01-01

Professional codes of ethics are social contracts among members of a professional group, which aim to instigate, encourage and nurture ethical behaviour and prevent professional misconduct, including research and publication. Despite the existence of codes of ethics, research misconduct remains a serious problem. A survey of codes of ethics from 795 professional organizations from the Illinois Institute of Technology's Codes of Ethics Collection showed that 182 of them (23%) used research integrity and research ethics terminology in their codes, with differences across disciplines: while the terminology was common in professional organizations in social sciences (82%), mental health (71%), sciences (61%), other organizations had no statements (construction trades, fraternal social organizations, real estate) or a few of them (management, media, engineering). A subsample of 158 professional organizations we judged to be directly involved in research significantly more often had statements on research integrity/ethics terminology than the whole sample: an average of 10.4% of organizations with a statement (95% CI = 10.4-23-5%) on any of the 27 research integrity/ethics terms compared to 3.3% (95% CI = 2.1-4.6%), respectively (Porganizations should define research integrity and research ethics issues in their ethics codes and collaborate within and across disciplines to adequately address responsible conduct of research and meet contemporary needs of their communities.

WDEC: A Code for Modeling White Dwarf Structure and Pulsations

Science.gov (United States)

Bischoff-Kim, Agnès; Montgomery, Michael H.

2018-05-01

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

Model-Driven Engineering of Machine Executable Code

Science.gov (United States)

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

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

Integration of anatomical and external response mappings explains crossing effects in tactile localization: A probabilistic modeling approach.

Science.gov (United States)

Badde, Stephanie; Heed, Tobias; Röder, Brigitte

2016-04-01

To act upon a tactile stimulus its original skin-based, anatomical spatial code has to be transformed into an external, posture-dependent reference frame, a process known as tactile remapping. When the limbs are crossed, anatomical and external location codes are in conflict, leading to a decline in tactile localization accuracy. It is unknown whether this impairment originates from the integration of the resulting external localization response with the original, anatomical one or from a failure of tactile remapping in crossed postures. We fitted probabilistic models based on these diverging accounts to the data from three tactile localization experiments. Hand crossing disturbed tactile left-right location choices in all experiments. Furthermore, the size of these crossing effects was modulated by stimulus configuration and task instructions. The best model accounted for these results by integration of the external response mapping with the original, anatomical one, while applying identical integration weights for uncrossed and crossed postures. Thus, the model explained the data without assuming failures of remapping. Moreover, performance differences across tasks were accounted for by non-individual parameter adjustments, indicating that individual participants' task adaptation results from one common functional mechanism. These results suggest that remapping is an automatic and accurate process, and that the observed localization impairments in touch result from a cognitively controlled integration process that combines anatomically and externally coded responses.

Adaptation of multidimensional group particle tracking and particle wall-boundary condition model to the FDNS code

Science.gov (United States)

Chen, Y. S.; Farmer, R. C.

1992-01-01

A particulate two-phase flow CFD model was developed based on the FDNS code which is a pressure based predictor plus multi-corrector Navier-Stokes flow solver. Turbulence models with compressibility correction and the wall function models were employed as submodels. A finite-rate chemistry model was used for reacting flow simulation. For particulate two-phase flow simulations, a Eulerian-Lagrangian solution method using an efficient implicit particle trajectory integration scheme was developed in this study. Effects of particle-gas reaction and particle size change to agglomeration or fragmentation were not considered in this investigation. At the onset of the present study, a two-dimensional version of FDNS which had been modified to treat Lagrangian tracking of particles (FDNS-2DEL) had already been written and was operational. The FDNS-2DEL code was too slow for practical use, mainly because it had not been written in a form amenable to vectorization on the Cray, nor was the full three-dimensional form of FDNS utilized. The specific objective of this study was to reorder to calculations into long single arrays for automatic vectorization on the Cray and to implement the full three-dimensional version of FDNS to produce the FDNS-3DEL code. Since the FDNS-2DEL code was slow, a very limited number of test cases had been run with it. This study was also intended to increase the number of cases simulated to verify and improve, as necessary, the particle tracking methodology coded in FDNS.

Development and assessment of the COBRA/RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Jeong, Jae Jun; Ha, Kwi Seok; Sim, Seok Ku

1997-04-01

The COBRA/RELAP5 code, a merged version of the COBRA-TF and RELAP5/MOD3.2 codes, has been developed to combine the realistic three-dimensional reactor vessel model of COBRA-TF with RELAP5/MOD3, thus to produce an advanced system analysis code with a multidimensional thermal-hydraulic module. This report provides the integration scheme of the two codes and the results of developmental assessments. These includes single channel tests, manometric flow oscillation problem, THTF Test 105, and LOFT L2-3 large-break loss-of-coolant experiment. From the single channel tests the integration scheme and its implementation were proven to be valid. Other simulation results showed good agreement with the experiments. The computational speed was also satisfactory. So it is confirmed that COBRA/RELAP5 can be a promising tool for analysis of complicated, multidimensional two-phase flow transients. The area of further improvements in the code integration are also identified. This report also serves as a user`s manual for the COBRA/RELAP5 code. (author). 6 tabs., 20 figs., 20 refs.

An Analysis of Countries which have Integrated Coding into their Curricula and the Content Analysis of Academic Studies on Coding Training in Turkey

Directory of Open Access Journals (Sweden)

Hüseyin Uzunboylu

2017-11-01

Full Text Available The first aim is to conduct a general analysis of countries which have integrated coding training into their curricula, and the second aim is to conduct a content analysis of studies on coding training in Turkey. It was identified that there are only a few academic studies on coding training in Turkey, and that the majority of them were published in 2016, the intended population was mainly “undergraduate students” and that the majority of these students were Computer Education and Instructional Technology undergraduates. It was determined that the studies mainly focused on the subjects of “programming” and “Scratch”, the terms programming and coding were used as synonyms, most of the studies were carried out using quantitative methods and data was obtained mostly by literature review and scale/survey interval techniques.

Computer code development plant for SMART design

International Nuclear Information System (INIS)

Bae, Kyoo Hwan; Choi, S.; Cho, B.H.; Kim, K.K.; Lee, J.C.; Kim, J.P.; Kim, J.H.; Chung, M.; Kang, D.J.; Chang, M.H.

1999-03-01

In accordance with the localization plan for the nuclear reactor design driven since the middle of 1980s, various computer codes have been transferred into the korea nuclear industry through the technical transfer program from the worldwide major pressurized water reactor supplier or through the international code development program. These computer codes have been successfully utilized in reactor and reload core design works. As the results, design- related technologies have been satisfactorily accumulated. However, the activities for the native code development activities to substitute the some important computer codes of which usages are limited by the original technique owners have been carried out rather poorly. Thus, it is most preferentially required to secure the native techniques on the computer code package and analysis methodology in order to establish the capability required for the independent design of our own model of reactor. Moreover, differently from the large capacity loop-type commercial reactors, SMART (SYSTEM-integrated Modular Advanced ReacTor) design adopts a single reactor pressure vessel containing the major primary components and has peculiar design characteristics such as self-controlled gas pressurizer, helical steam generator, passive residual heat removal system, etc. Considering those peculiar design characteristics for SMART, part of design can be performed with the computer codes used for the loop-type commercial reactor design. However, most of those computer codes are not directly applicable to the design of an integral reactor such as SMART. Thus, they should be modified to deal with the peculiar design characteristics of SMART. In addition to the modification efforts, various codes should be developed in several design area. Furthermore, modified or newly developed codes should be verified their reliability through the benchmarking or the test for the object design. Thus, it is necessary to proceed the design according to the

Computer code development plant for SMART design

Energy Technology Data Exchange (ETDEWEB)

Bae, Kyoo Hwan; Choi, S.; Cho, B.H.; Kim, K.K.; Lee, J.C.; Kim, J.P.; Kim, J.H.; Chung, M.; Kang, D.J.; Chang, M.H

1999-03-01

In accordance with the localization plan for the nuclear reactor design driven since the middle of 1980s, various computer codes have been transferred into the korea nuclear industry through the technical transfer program from the worldwide major pressurized water reactor supplier or through the international code development program. These computer codes have been successfully utilized in reactor and reload core design works. As the results, design- related technologies have been satisfactorily accumulated. However, the activities for the native code development activities to substitute the some important computer codes of which usages are limited by the original technique owners have been carried out rather poorly. Thus, it is most preferentially required to secure the native techniques on the computer code package and analysis methodology in order to establish the capability required for the independent design of our own model of reactor. Moreover, differently from the large capacity loop-type commercial reactors, SMART (SYSTEM-integrated Modular Advanced ReacTor) design adopts a single reactor pressure vessel containing the major primary components and has peculiar design characteristics such as self-controlled gas pressurizer, helical steam generator, passive residual heat removal system, etc. Considering those peculiar design characteristics for SMART, part of design can be performed with the computer codes used for the loop-type commercial reactor design. However, most of those computer codes are not directly applicable to the design of an integral reactor such as SMART. Thus, they should be modified to deal with the peculiar design characteristics of SMART. In addition to the modification efforts, various codes should be developed in several design area. Furthermore, modified or newly developed codes should be verified their reliability through the benchmarking or the test for the object design. Thus, it is necessary to proceed the design according to the

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-01-15

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

Implementation of wall film condensation model to two-fluid model in component thermal hydraulic analysis code CUPID - 15237

International Nuclear Information System (INIS)

Lee, J.H.; Park, G.C.; Cho, H.K.

2015-01-01

In the containment of a nuclear reactor, the wall condensation occurs when containment cooling system and structures remove the mass and energy release and this phenomenon is of great importance to ensure containment integrity. If the phenomenon occurs in the presence of non-condensable gases, their accumulation near the condensate film leads to significant reduction in heat transfer during the condensation. This study aims at simulating the wall film condensation in the presence of non-condensable gas using CUPID, a computational multi-fluid dynamics code, which is developed by the Korea Atomic Energy Research Institute (KAERI) for the analysis of transient two-phase flows in nuclear reactor components. In order to simulate the wall film condensation in containment, the code requires a proper wall condensation model and liquid film model applicable to the analysis of the large scale system. In the present study, the liquid film model and wall film condensation model were implemented in the two-fluid model of CUPID. For the condensation simulation, a wall function approach with heat and mass transfer analogy was applied in order to save computational time without considerable refinement for the boundary layer. This paper presents the implemented wall film condensation model and then, introduces the simulation result using CUPID with the model for a conceptual condensation problem in a large system. (authors)

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.

Development of an advanced PFM code for the integrity evaluation of nuclear piping system under combined aging mechanisms

International Nuclear Information System (INIS)

Datta, Debashis

2010-02-01

A nuclear piping system is composed of several straight pipes and elbows joined by welding. These weld sections are usually the most susceptible failure parts susceptible to various degradation mechanisms. Whereas a specific location of a reactor piping system might fail by a combination of different aging mechanisms, e.g. fatigue and/or stress corrosion cracking, the majority of the piping probabilistic fracture mechanics (PFM) codes can only consider a single aging mechanism at a time. So, a probabilistic fracture mechanics computer code capable of considering multiple aging mechanisms was developed for an accurate failure analysis of each specific component of a nuclear piping section. The newly proposed crack morphology based probabilistic leak flow rate module is introduced in this code to separately treat fatigue and SCC type cracks. Improved models e.g. stressors models, elbow failure model, SIFs model, local seismic occurrence probability model, performance based crack detection models, etc., are also included in this code. Recent probabilistic fatigue (S-N) and SCC crack initiation (S-T) and subsequent crack growth rate models are coded. An integrated probabilistic risk assessment and probabilistic fracture mechanics methodology is proposed. A complete flow chart regarding the combined aging mechanism model is presented. The combined aging mechanism based module can significantly reduce simulation efforts and time. Two NUREG benchmark problems, e.g. reactor pressure vessel outlet nozzle section and a surge line elbow located just below the pressurizer are reinvestigated by this code. The results showed that, contribution of pre-existing cracks in addition to initiating cracks, can significantly increase the overall failure probability. Inconel weld location of reactor pressure vessel outlet nozzle section showed the weakest point in terms of relative through-wall leak failure probability in the order of about 10 -2 at the 40-year plant life. Considering

Development of Adiabatic Doppler Feedback Model in 3D space time analysis Code ARCH

International Nuclear Information System (INIS)

Dwivedi, D.K.; Gupta, Anurag

2015-01-01

Integrated 3D space-time neutron kinetics with thermal-hydraulic feedback code system is being developed for transient analysis of Compact High Temperature Reactor (CHTR) and Advanced Heavy Water Reactor (AHWR). ARCH (code for Analysis of Reactor transients in Cartesian and Hexagon geometries) has been developed with IQS module for efficient 3D space time analysis. Recently, an adiabatic Doppler (fuel temperature) feedback module has been incorporated in this ARCH-IQS version of tile code. In the adiabatic model of fuel temperature feedback, the transfer of the excess heat from the fuel to the coolant during transient is neglected. The viability of Doppler feedback in ARCH-IQS with adiabatic heating has been checked with AER benchmark (Dyn002). Analyses of anticipated transient without scram (ATWS) case in CHTR as well as in AHWR have been performed with adiabatic fuel temperature feedback. The methodology and results have been presented in this paper. (author)

Results from the Metis code participation to the Hydrocoin exercise

International Nuclear Information System (INIS)

Raimbault, P.

1987-04-01

The METIS code, developed at the ENSMP is a 2D finite element radionuclide transport and groundwater flow model based on the hypothesis of an equivalent porous medium with an explicit description of the main fractures. It is integrated in the global risk assessment code MELODIE for nuclear waste repositories in geological formations. The participation of the METIS code to the HYDROCOIN exercise is of prime importance for its development and its incorporation in the performance assessment procedure in France. Results from HYDROCOIN cases show that the code can handle correctly fractured media, high permeability contrast formations and buoyancy effects. A 3D version of the code has been developed for carrying comparisons of field experiments and groundwater flow models in HYDROCOIN level 2. In order to carry out the exercise, several pre and post-processing programs were developed and integrated in a conversational module. They include: contour plots, velocity field representations, interpolations, particule tracking routines and uncertainty and sensitivity analysis modules

Structure of fuel performance audit code for SFR metal fuel

Energy Technology Data Exchange (ETDEWEB)

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

2012-10-15

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

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

International Nuclear Information System (INIS)

Kim, See Darl; Kim, Dong Ha

1998-04-01

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

Two-phase flow characteristics analysis code: MINCS

International Nuclear Information System (INIS)

Watanabe, Tadashi; Hirano, Masashi; Akimoto, Masayuki; Tanabe, Fumiya; Kohsaka, Atsuo.

1992-03-01

Two-phase flow characteristics analysis code: MINCS (Modularized and INtegrated Code System) has been developed to provide a computational tool for analyzing two-phase flow phenomena in one-dimensional ducts. In MINCS, nine types of two-phase flow models-from a basic two-fluid nonequilibrium (2V2T) model to a simple homogeneous equilibrium (1V1T) model-can be used under the same numerical solution method. The numerical technique is based on the implicit finite difference method to enhance the numerical stability. The code structure is highly modularized, so that new constitutive relations and correlations can be easily implemented into the code and hence evaluated. A flow pattern can be fixed regardless of flow conditions, and state equations or steam tables can be selected. It is, therefore, easy to calculate physical or numerical benchmark problems. (author)

Research Integrity and Research Ethics in Professional Codes of Ethics: Survey of Terminology Used by Professional Organizations across Research Disciplines

Science.gov (United States)

Komić, Dubravka; Marušić, Stjepan Ljudevit; Marušić, Ana

2015-01-01

Professional codes of ethics are social contracts among members of a professional group, which aim to instigate, encourage and nurture ethical behaviour and prevent professional misconduct, including research and publication. Despite the existence of codes of ethics, research misconduct remains a serious problem. A survey of codes of ethics from 795 professional organizations from the Illinois Institute of Technology’s Codes of Ethics Collection showed that 182 of them (23%) used research integrity and research ethics terminology in their codes, with differences across disciplines: while the terminology was common in professional organizations in social sciences (82%), mental health (71%), sciences (61%), other organizations had no statements (construction trades, fraternal social organizations, real estate) or a few of them (management, media, engineering). A subsample of 158 professional organizations we judged to be directly involved in research significantly more often had statements on research integrity/ethics terminology than the whole sample: an average of 10.4% of organizations with a statement (95% CI = 10.4-23-5%) on any of the 27 research integrity/ethics terms compared to 3.3% (95% CI = 2.1–4.6%), respectively (Pethics concepts used prescriptive language in describing the standard of practice. Professional organizations should define research integrity and research ethics issues in their ethics codes and collaborate within and across disciplines to adequately address responsible conduct of research and meet contemporary needs of their communities. PMID:26192805

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

DEFF Research Database (Denmark)

Huang, Xin; Forchhammer, Søren

2012-01-01

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

Model integration and a theory of models

OpenAIRE

Dolk, Daniel R.; Kottemann, Jeffrey E.

1993-01-01

Model integration extends the scope of model management to include the dimension of manipulation as well. This invariably leads to comparisons with database theory. Model integration is viewed from four perspectives: Organizational, definitional, procedural, and implementational. Strategic modeling is discussed as the organizational motivation for model integration. Schema and process integration are examined as the logical and manipulation counterparts of model integr...

Coupling a Basin Modeling and a Seismic Code using MOAB

KAUST Repository

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

2012-01-01

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

Coupling a Basin Modeling and a Seismic Code using MOAB

KAUST Repository

Yan, Mi

2012-06-02

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

Development of plant dynamic analysis code for integrated self-pressurized water reactor (ISPDYN), and comparative study of pressure control methods

International Nuclear Information System (INIS)

Kusunoki, Tsuyoshi; Yokomura, Takeyoshi; Nabeshima, Kunihiko; Shimazaki, Junya; Shinohara, Yoshikuni.

1988-01-01

This report describes the development of plant dynamic analysis code (ISPDYN) for integrated self-pressurized water reactor, and comparative study of pressure control methods with this code. ISPDYN is developed for integrated self-pressurized water reactor, one of the trial design by JAERI. In the transient responses, the calculated results by ISPDYN are in good agreement with the DRUCK calculations. In addition, this report presents some sensitivity studies for selected cases. Computing time of this code is very short so as about one fifth of real time. The comparative study of self-pressurized system with forced-pressurized system by this code, for rapid load decrease and increase cases, has provided useful informations. (author)

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

International Nuclear Information System (INIS)

1995-08-01

The RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents and operational transients such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. RELAP5/MOD3 code documentation is divided into seven volumes: Volume I presents modeling theory and associated numerical schemes; Volume II details instructions for code application and input data preparation; Volume III presents the results of developmental assessment cases that demonstrate and verify the models used in the code; Volume IV discusses in detail RELAP5 models and correlations; Volume V presents guidelines that have evolved over the past several years through the use of the RELAP5 code; Volume VI discusses the numerical scheme used in RELAP5; and Volume VII presents a collection of independent assessment calculations

A vectorized Monte Carlo code for modeling photon transport in SPECT

International Nuclear Information System (INIS)

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

1993-01-01

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

The analysis of thermal-hydraulic models in MELCOR code

Energy Technology Data Exchange (ETDEWEB)

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

1996-07-15

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

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

International Nuclear Information System (INIS)

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

1987-05-01

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

MCB. A continuous energy Monte Carlo burnup simulation code

International Nuclear Information System (INIS)

Cetnar, J.; Wallenius, J.; Gudowski, W.

1999-01-01

A code for integrated simulation of neutrinos and burnup based upon continuous energy Monte Carlo techniques and transmutation trajectory analysis has been developed. Being especially well suited for studies of nuclear waste transmutation systems, the code is an extension of the well validated MCNP transport program of Los Alamos National Laboratory. Among the advantages of the code (named MCB) is a fully integrated data treatment combined with a time-stepping routine that automatically corrects for burnup dependent changes in reaction rates, neutron multiplication, material composition and self-shielding. Fission product yields are treated as continuous functions of incident neutron energy, using a non-equilibrium thermodynamical model of the fission process. In the present paper a brief description of the code and applied methods are given. (author)

ICARE/CATHARE and ASTEC code development trends

International Nuclear Information System (INIS)

Chatelard, P.; Dorsselaere, J.-P. van

2000-01-01

Regarding the computer code development for simulation of LWR severe accidents, IPSN developed a two-tier approach based on detailed codes such as ICARE/CATHARE and simplified models to be assembled in the ASTEC integral code. The ICARE/CATHARE code results from the coupling between the ICARE2 code modelling the core degradation phenomena and the thermalhydraulics code CATHARE2. It allows to calculate PWR and VVER severe accident sequences in the whole RCS. The modelling of the early degradation phase can be considered as rather complete in the ICARE/CATHARE V1 mod1 version (to be released by mid-2000) whereas some models are still missing for the late phase. The main future developments (ICARE/CATHARE V2) will concern the multi-dimensional thermalhydraulics, the quenching of partially damaged cores (mechanical and chemical effects), the debris bed two-phase thermalhydraulics (including reflooding) and the corium behaviour in the lower head. The main other physical improvements should concern the behaviour of boron carbide control rods, the processes governing the core loss of geometry (transition phase) and the oxidation of relocated melts. The ASTEC (Accident Source Term Evaluation Code) integral code, commonly developed by IPSN and GRS, aims to predict an entire LWR (PWR, VVER and BWR) severe accident sequence from the initiating event through to FP release out of the containment, for source term, PSA level 2, or accident management studies. The version ASTEC VO.3 to be released by mid-2000 can be considered now as robust and fast-running enough (between 2 and 12 hours for a one day accident) and allows to perform, with a containment multi-compartment configuration, any scenario accident study accounting for the main safety systems and operator procedures (spray, recombiner, etc.). The next version ASTEC V1, to be released beginning of 2002, will include the frontend simulation and improve modelling of in-vessel core degradation. A large validation activity will

MARS CODE MANUAL VOLUME III - Programmer's Manual

International Nuclear Information System (INIS)

Chung, Bub Dong; Hwang, Moon Kyu; Jeong, Jae Jun; Kim, Kyung Doo; Bae, Sung Won; Lee, Young Jin; Lee, Won Jae

2010-02-01

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

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

CSIR Research Space (South Africa)

Grobler, Jan H

2010-01-01

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

Universal Regularizers For Robust Sparse Coding and Modeling

OpenAIRE

Ramirez, Ignacio; Sapiro, Guillermo

2010-01-01

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

Phenomenological optical potentials and optical model computer codes

International Nuclear Information System (INIS)

Prince, A.

1980-01-01

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

Predictive Coding and Multisensory Integration: An Attentional Account of the Multisensory Mind

Directory of Open Access Journals (Sweden)

Durk eTalsma

2015-03-01

Full Text Available Multisensory integration involves a host of different cognitive processes, occurring at different stages of sensory processing. Here I argue that, despite recent insights suggesting that multisensory interactions can occur at very early latencies, the actual integration of individual sensory traces into an internally consistent mental representation is dependent on both top-down and bottom-up processes. Moreover, I argue that this integration is not limited to just sensory inputs, but that internal cognitive processes also shape the resulting mental representation. Studies showing that memory recall is affected by the initial multisensory context in which the stimuli were presented will be discussed, as well as several studies showing that mental imagery can affect multisensory illusions. This empirical evidence will be discussed from a predictive coding perspective, in which a central top-down attentional process is proposed to play a central role in coordinating the integration of all these inputs into a coherent mental representation.

Anthropomorphic Coding of Speech and Audio: A Model Inversion Approach

Directory of Open Access Journals (Sweden)

W. Bastiaan Kleijn

2005-06-01

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

Validation of TGLF in C-Mod and DIII-D using machine learning and integrated modeling tools

Science.gov (United States)

Rodriguez-Fernandez, P.; White, Ae; Cao, Nm; Creely, Aj; Greenwald, Mj; Grierson, Ba; Howard, Nt; Meneghini, O.; Petty, Cc; Rice, Je; Sciortino, F.; Yuan, X.

2017-10-01

Predictive models for steady-state and perturbative transport are necessary to support burning plasma operations. A combination of machine learning algorithms and integrated modeling tools is used to validate TGLF in C-Mod and DIII-D. First, a new code suite, VITALS, is used to compare SAT1 and SAT0 models in C-Mod. VITALS exploits machine learning and optimization algorithms for the validation of transport codes. Unlike SAT0, the SAT1 saturation rule contains a model to capture cross-scale turbulence coupling. Results show that SAT1 agrees better with experiments, further confirming that multi-scale effects are needed to model heat transport in C-Mod L-modes. VITALS will next be used to analyze past data from DIII-D: L-mode ``Shortfall'' plasma and ECH swing experiments. A second code suite, PRIMA, allows for integrated modeling of the plasma response to Laser Blow-Off cold pulses. Preliminary results show that SAT1 qualitatively reproduces the propagation of cold pulses after LBO injections and SAT0 does not, indicating that cross-scale coupling effects play a role in the plasma response. PRIMA will be used to ``predict-first'' cold pulse experiments using the new LBO system at DIII-D, and analyze existing ECH heat pulse data. Work supported by DE-FC02-99ER54512, DE-FC02-04ER54698.

A Simplified Baseband Prefilter Model with Adaptive Kalman Filter for Ultra-Tight COMPASS/INS Integration

Science.gov (United States)

Luo, Yong; Wu, Wenqi; Babu, Ravindra; Tang, Kanghua; Luo, Bing

2012-01-01

COMPASS is an indigenously developed Chinese global navigation satellite system and will share many features in common with GPS (Global Positioning System). Since the ultra-tight GPS/INS (Inertial Navigation System) integration shows its advantage over independent GPS receivers in many scenarios, the federated ultra-tight COMPASS/INS integration has been investigated in this paper, particularly, by proposing a simplified prefilter model. Compared with a traditional prefilter model, the state space of this simplified system contains only carrier phase, carrier frequency and carrier frequency rate tracking errors. A two-quadrant arctangent discriminator output is used as a measurement. Since the code tracking error related parameters were excluded from the state space of traditional prefilter models, the code/carrier divergence would destroy the carrier tracking process, and therefore an adaptive Kalman filter algorithm tuning process noise covariance matrix based on state correction sequence was incorporated to compensate for the divergence. The federated ultra-tight COMPASS/INS integration was implemented with a hardware COMPASS intermediate frequency (IF), and INS's accelerometers and gyroscopes signal sampling system. Field and simulation test results showed almost similar tracking and navigation performances for both the traditional prefilter model and the proposed system; however, the latter largely decreased the computational load. PMID:23012564

A Simplified Baseband Prefilter Model with Adaptive Kalman Filter for Ultra-Tight COMPASS/INS Integration

Directory of Open Access Journals (Sweden)

Bing Luo

2012-07-01

Full Text Available COMPASS is an indigenously developed Chinese global navigation satellite system and will share many features in common with GPS (Global Positioning System. Since the ultra-tight GPS/INS (Inertial Navigation System integration shows its advantage over independent GPS receivers in many scenarios, the federated ultra-tight COMPASS/INS integration has been investigated in this paper, particularly, by proposing a simplified prefilter model. Compared with a traditional prefilter model, the state space of this simplified system contains only carrier phase, carrier frequency and carrier frequency rate tracking errors. A two-quadrant arctangent discriminator output is used as a measurement. Since the code tracking error related parameters were excluded from the state space of traditional prefilter models, the code/carrier divergence would destroy the carrier tracking process, and therefore an adaptive Kalman filter algorithm tuning process noise covariance matrix based on state correction sequence was incorporated to compensate for the divergence. The federated ultra-tight COMPASS/INS integration was implemented with a hardware COMPASS intermediate frequency (IF, and INS’s accelerometers and gyroscopes signal sampling system. Field and simulation test results showed almost similar tracking and navigation performances for both the traditional prefilter model and the proposed system; however, the latter largely decreased the computational load.

Fission product release from nuclear fuel I. Physical modelling in the ASTEC code

International Nuclear Information System (INIS)

Brillant, G.; Marchetto, C.; Plumecocq, W.

2013-01-01

Highlights: • Physical modeling of FP and SM release in ASTEC is presented. • The release is described as solid state diffusion within fuel for high volatile FP. • The release is described as FP vaporisation for semi volatile FP. • The release is described as fuel vaporisation for low volatile FP. • ASTEC validation is presented in the second paper. - Abstract: This article is the first of a series of two articles dedicated to the mechanisms of fission product release from a degraded core as they are modelled in the ASTEC code. The ASTEC code aims at simulating severe accidents in nuclear reactors from the initiating event up to the radiological consequences on the environment. This code is used for several applications such as nuclear plant safety evaluation including probabilistic studies and emergency preparedness. To cope with the requirements of robustness and low calculation time, the code is based on a semi-empirical approach and only the main limiting phenomena that govern the release from intact rods and from debris beds are considered. For solid fuel, fission products are classified into three groups, depending on their degree of volatility. The kinetics of volatile fission products release depend on the rate-limiting process of solid-state diffusion through fuel grains. For semi-volatile fission products, the release from the open fuel porosities is assumed to be governed by vaporisation and mass transfer processes. The key phenomenon for the release of low volatile fission products is supposed to be fuel volatilisation. A similar approach is used for the release of fission products from a rubble bed. An in-depth validation of the code including both analytical and integral experiments is the subject of the second article

An improved Zircaloy-steam reaction model for use with the March 2 (Meltdown Accident Response Characteristics) code

International Nuclear Information System (INIS)

Manahan, M.P.

1983-01-01

An improved Zircaloy-steam oxidation reaction model has been incorporated into the MARCH 2 code which includes: (1) improved physical modeling for solid-state process oxidation, (2) improved geometric modeling for gaseous diffusion oxidation, (3) chemisorption/dissociation retardation due to high hydrogen partial pressures, and (4) laminar and turbulent flow conditions. Several accident sequences have been analyzed using the model, and for the sequences considered, the results indicate that the integrated and averaged variables are not significantly altered for the current level of fuel modeling, however, the localized variables such as nodal temperature and oxide thickness are affected

InP monolithically integrated label swapper device for spectral amplitude coded optical packet networks

NARCIS (Netherlands)

Muñoz, P.; García-Olcina, R.; Doménech, J.D.; Rius, M.; Sancho, J.C.; Capmany, J.; Chen, L.R.; Habib, C.; Leijtens, X.J.M.; Vries, de T.; Heck, M.J.R.; Augustin, L.M.; Nötzel, R.; Robbins, D.J.

2010-01-01

In this paper a label swapping device, for spectral amplitude coded optical packet networks, fully integrated using InP technology is presented. Compared to previous demonstrations using discrete component assembly, the device footprint is reduced by a factor of 105 and the operation speed is

Constitutive model development needs for reactor safety thermal-hydraulic codes

International Nuclear Information System (INIS)

Kelly, J.M.

1998-01-01

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

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

Science.gov (United States)

American Inst. of Architects, Washington, DC.

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-08-01

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

Development of thermal hydraulic models for the reliable regulatory auditing code

Energy Technology Data Exchange (ETDEWEB)

Chung, B. D.; Song, C. H.; Lee, Y. J.; Kwon, T. S.; Lee, S. W. [Korea Automic Energy Research Institute, Taejon (Korea, Republic of)

2004-02-15

The objective of this project is to develop thermal hydraulic models for use in improving the reliability of the regulatory auditing codes. The current year fall under the second step of the 3 year project, and the main researches were focused on the development of downcorner boiling model. During the current year, the bubble stream model of downcorner has been developed and installed in he auditing code. The model sensitivity analysis has been performed for APR1400 LBLOCA scenario using the modified code. The preliminary calculation has been performed for the experimental test facility using FLUENT and MARS code. The facility for air bubble experiment has been installed. The thermal hydraulic phenomena for VHTR and super critical reactor have been identified for the future application and model development.

Coupling methods for parallel running RELAPSim codes in nuclear power plant simulation

Energy Technology Data Exchange (ETDEWEB)

Li, Yankai; Lin, Meng, E-mail: linmeng@sjtu.edu.cn; Yang, Yanhua

2016-02-15

When the plant is modeled detailedly for high precision, it is hard to achieve real-time calculation for one single RELAP5 in a large-scale simulation. To improve the speed and ensure the precision of simulation at the same time, coupling methods for parallel running RELAPSim codes were proposed in this study. Explicit coupling method via coupling boundaries was realized based on a data-exchange and procedure-control environment. Compromise of synchronization frequency was well considered to improve the precision of simulation and guarantee the real-time simulation at the same time. The coupling methods were assessed using both single-phase flow models and two-phase flow models and good agreements were obtained between the splitting–coupling models and the integrated model. The mitigation of SGTR was performed as an integral application of the coupling models. A large-scope NPP simulator was developed adopting six splitting–coupling models of RELAPSim and other simulation codes. The coupling models could improve the speed of simulation significantly and make it possible for real-time calculation. In this paper, the coupling of the models in the engineering simulator is taken as an example to expound the coupling methods, i.e., coupling between parallel running RELAPSim codes, and coupling between RELAPSim code and other types of simulation codes. However, the coupling methods are also referable in other simulator, for example, a simulator employing ATHLETE instead of RELAP5, other logic code instead of SIMULINK. It is believed the coupling method is commonly used for NPP simulator regardless of the specific codes chosen in this paper.

SWAT4.0 - The integrated burnup code system driving continuous energy Monte Carlo codes MVP, MCNP and deterministic calculation code SRAC

International Nuclear Information System (INIS)

Kashima, Takao; Suyama, Kenya; Takada, Tomoyuki

2015-03-01

There have been two versions of SWAT depending on details of its development history: the revised SWAT that uses the deterministic calculation code SRAC as a neutron transportation solver, and the SWAT3.1 that uses the continuous energy Monte Carlo code MVP or MCNP5 for the same purpose. It takes several hours, however, to execute one calculation by the continuous energy Monte Carlo code even on the super computer of the Japan Atomic Energy Agency. Moreover, two-dimensional burnup calculation is not practical using the revised SWAT because it has problems on production of effective cross section data and applying them to arbitrary fuel geometry when a calculation model has multiple burnup zones. Therefore, SWAT4.0 has been developed by adding, to SWAT3.1, a function to utilize the deterministic code SARC2006, which has shorter calculation time, as an outer module of neutron transportation solver for burnup calculation. SWAT4.0 has been enabled to execute two-dimensional burnup calculation by providing an input data template of SRAC2006 to SWAT4.0 input data, and updating atomic number densities of burnup zones in each burnup step. This report describes outline, input data instruction, and examples of calculations of SWAT4.0. (author)

CATHARE code development and assessment methodologies

International Nuclear Information System (INIS)

Micaelli, J.C.; Barre, F.; Bestion, D.

1995-01-01

The CATHARE thermal-hydraulic code has been developed jointly by Commissariat a l'Energie Atomique (CEA), Electricite de France (EdF), and Framatorne for safety analysis. Since the beginning of the project (September 1979), development and assessment activities have followed a methodology supported by two series of experimental tests: separate effects tests and integral effects tests. The purpose of this paper is to describe this methodology, the code assessment status, and the evolution to take into account two new components of this program: the modeling of three-dimensional phenomena and the requirements of code uncertainty evaluation

The TE coupled RELAP5/PANTHER/COBRA code package and methodology for integrated PWR accident analysis

International Nuclear Information System (INIS)

Schneidesch, Christophe R.; Zhang, Jinzhao; Ammirabile, Luca; Dalleur, Jean-Paul

2006-01-01

At Tractebel Engineering (TE), a dynamic coupling has been developed between the best estimate thermal hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel core thermal-hydraulic analysis code COBRA 3C has been established for on-line calculation of the Departure from Nucleate Boiling Ratio (DNBR). In addition to the standard RELAP5-PANTHER coupling, the fully dynamic coupling of the RELAP5/PANTHER/COBRA3C-TE code package can be activated for evaluation purposes in which the PANTHER close-channel thermal-hydraulics module is replaced by the COBRA3C-TE with cross flow modelling and extended T-H flow conditions capabilities. The qualification of the RELAP5-PANTHER coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been approved by the Belgian Safety Authorities and is used at TE for analyzing asymmetric PWR accidents with strong core-system interactions. In particular, the TE coupled code package was first used to develop a main steam line break in hot shutdown conditions (SLBHZP) accident analysis methodology based on the TE deterministic bounding approach. This methodology has been reviewed and accepted by the Belgian Safety Authorities for specific applications. Those specific applications are related to the power up-rate and steam generator replacement project of the Doel 2 plant or to the Tihange-3 SLB accident re-analysis. A coupled feedwater line break (FLB) accident analysis methodology is currently being reviewed for application approval. The results of coupled thermal-hydraulic and neutronic analysis of SLB and FLB show that there exist important margins in the traditional final safety analysis report (FSAR) accident analysis. Those margins can be used to increase the operational flexibility of the plants. Moreover, the

The TE coupled RELAP5/PANTHER/COBRA code package and methodology for integrated PWR accident analysis

Energy Technology Data Exchange (ETDEWEB)

Schneidesch, Christophe R.; Zhang, Jinzhao; Ammirabile, Luca; Dalleur, Jean-Paul [Suez-Tractebel Engineering, Avenue Ariane 7, B-1200 Brussels (Belgium)

2006-07-01

At Tractebel Engineering (TE), a dynamic coupling has been developed between the best estimate thermal hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel core thermal-hydraulic analysis code COBRA 3C has been established for on-line calculation of the Departure from Nucleate Boiling Ratio (DNBR). In addition to the standard RELAP5-PANTHER coupling, the fully dynamic coupling of the RELAP5/PANTHER/COBRA3C-TE code package can be activated for evaluation purposes in which the PANTHER close-channel thermal-hydraulics module is replaced by the COBRA3C-TE with cross flow modelling and extended T-H flow conditions capabilities. The qualification of the RELAP5-PANTHER coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been approved by the Belgian Safety Authorities and is used at TE for analyzing asymmetric PWR accidents with strong core-system interactions. In particular, the TE coupled code package was first used to develop a main steam line break in hot shutdown conditions (SLBHZP) accident analysis methodology based on the TE deterministic bounding approach. This methodology has been reviewed and accepted by the Belgian Safety Authorities for specific applications. Those specific applications are related to the power up-rate and steam generator replacement project of the Doel 2 plant or to the Tihange-3 SLB accident re-analysis. A coupled feedwater line break (FLB) accident analysis methodology is currently being reviewed for application approval. The results of coupled thermal-hydraulic and neutronic analysis of SLB and FLB show that there exist important margins in the traditional final safety analysis report (FSAR) accident analysis. Those margins can be used to increase the operational flexibility of the plants. Moreover, the

ARC integration into the NEAMS Workbench

Energy Technology Data Exchange (ETDEWEB)

Stauff, N. [Argonne National Lab. (ANL), Argonne, IL (United States); Gaughan, N. [Argonne National Lab. (ANL), Argonne, IL (United States); Kim, T. [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-01-01

One of the objectives of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Integration Product Line (IPL) is to facilitate the deployment of the high-fidelity codes developed within the program. The Workbench initiative was launched in FY-2017 by the IPL to facilitate the transition from conventional tools to high fidelity tools. The Workbench provides a common user interface for model creation, real-time validation, execution, output processing, and visualization for integrated codes.

Comparative evaluation of structural integrity for ITER blanket shield block based on SDC-IC and ASME code

Energy Technology Data Exchange (ETDEWEB)

Shim, Hee-Jin [ITER Korea, National Fusion Research Institute, 169-148 Gwahak-Ro, Yuseong-Gu, Daejeon (Korea, Republic of); Ha, Min-Su, E-mail: msha12@nfri.re.kr [ITER Korea, National Fusion Research Institute, 169-148 Gwahak-Ro, Yuseong-Gu, Daejeon (Korea, Republic of); Kim, Sa-Woong; Jung, Hun-Chea [ITER Korea, National Fusion Research Institute, 169-148 Gwahak-Ro, Yuseong-Gu, Daejeon (Korea, Republic of); Kim, Duck-Hoi [ITER Organization, Route de Vinon sur Verdon - CS 90046, 13067 Sant Paul Lez Durance (France)

2016-11-01

Highlights: • The procedure of structural integrity and fatigue assessment was described. • Case studies were performed according to both SDC-IC and ASME Sec. • III codes The conservatism of the ASME code was demonstrated. • The study only covers the specifically comparable case about fatigue usage factor. - Abstract: The ITER blanket Shield Block is a bulk structure to absorb radiation and to provide thermal shielding to vacuum vessel and external vessel components, therefore the most significant load for Shield Block is the thermal load. In the previous study, the thermo-mechanical analysis has been performed under the inductive operation as representative loading condition. And the fatigue evaluations were conducted to assure structural integrity for Shield Block according to Structural Design Criteria for In-vessel Components (SDC-IC) which provided by ITER Organization (IO) based on the code of RCC-MR. Generally, ASME code (especially, B&PV Sec. III) is widely applied for design of nuclear components, and is usually well known as more conservative than other specific codes. For the view point of the fatigue assessment, ASME code is very conservative compared with SDC-IC in terms of the reflected K{sub e} factor, design fatigue curve and other factors. Therefore, an accurate fatigue assessment comparison is needed to measure of conservatism. The purpose of this study is to provide the fatigue usage comparison resulting from the specified operating conditions shall be evaluated for Shield Block based on both SDC-IC and ASME code, and to discuss the conservatism of the results.

Comparative evaluation of structural integrity for ITER blanket shield block based on SDC-IC and ASME code

International Nuclear Information System (INIS)

Shim, Hee-Jin; Ha, Min-Su; Kim, Sa-Woong; Jung, Hun-Chea; Kim, Duck-Hoi

2016-01-01

Highlights: • The procedure of structural integrity and fatigue assessment was described. • Case studies were performed according to both SDC-IC and ASME Sec. • III codes The conservatism of the ASME code was demonstrated. • The study only covers the specifically comparable case about fatigue usage factor. - Abstract: The ITER blanket Shield Block is a bulk structure to absorb radiation and to provide thermal shielding to vacuum vessel and external vessel components, therefore the most significant load for Shield Block is the thermal load. In the previous study, the thermo-mechanical analysis has been performed under the inductive operation as representative loading condition. And the fatigue evaluations were conducted to assure structural integrity for Shield Block according to Structural Design Criteria for In-vessel Components (SDC-IC) which provided by ITER Organization (IO) based on the code of RCC-MR. Generally, ASME code (especially, B&PV Sec. III) is widely applied for design of nuclear components, and is usually well known as more conservative than other specific codes. For the view point of the fatigue assessment, ASME code is very conservative compared with SDC-IC in terms of the reflected K_e factor, design fatigue curve and other factors. Therefore, an accurate fatigue assessment comparison is needed to measure of conservatism. The purpose of this study is to provide the fatigue usage comparison resulting from the specified operating conditions shall be evaluated for Shield Block based on both SDC-IC and ASME code, and to discuss the conservatism of the results.

Trading speed and accuracy by coding time: a coupled-circuit cortical model.

Directory of Open Access Journals (Sweden)

Dominic Standage

2013-04-01

Full Text Available Our actions take place in space and time, but despite the role of time in decision theory and the growing acknowledgement that the encoding of time is crucial to behaviour, few studies have considered the interactions between neural codes for objects in space and for elapsed time during perceptual decisions. The speed-accuracy trade-off (SAT provides a window into spatiotemporal interactions. Our hypothesis is that temporal coding determines the rate at which spatial evidence is integrated, controlling the SAT by gain modulation. Here, we propose that local cortical circuits are inherently suited to the relevant spatial and temporal coding. In simulations of an interval estimation task, we use a generic local-circuit model to encode time by 'climbing' activity, seen in cortex during tasks with a timing requirement. The model is a network of simulated pyramidal cells and inhibitory interneurons, connected by conductance synapses. A simple learning rule enables the network to quickly produce new interval estimates, which show signature characteristics of estimates by experimental subjects. Analysis of network dynamics formally characterizes this generic, local-circuit timing mechanism. In simulations of a perceptual decision task, we couple two such networks. Network function is determined only by spatial selectivity and NMDA receptor conductance strength; all other parameters are identical. To trade speed and accuracy, the timing network simply learns longer or shorter intervals, driving the rate of downstream decision processing by spatially non-selective input, an established form of gain modulation. Like the timing network's interval estimates, decision times show signature characteristics of those by experimental subjects. Overall, we propose, demonstrate and analyse a generic mechanism for timing, a generic mechanism for modulation of decision processing by temporal codes, and we make predictions for experimental verification.

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

Development of our laser fusion integration simulation

International Nuclear Information System (INIS)

Li, J.; Zhai, C.; Li, S.; Li, X.; Zheng, W.; Yong, H.; Zeng, Q.; Hang, X.; Qi, J.; Yang, R.; Cheng, J.; Song, P.; Gu, P.; Zhang, A.; An, H.; Xu, X.; Guo, H.; Cao, X.; Mo, Z.; Pei, W.; Jiang, S.; Zhu, S. P.

2013-01-01

In the target design of the Inertial Confinement Fusion (ICF) program, it is common practice to apply radiation hydrodynamics code to study the key physical processes happening in ICF process, such as hohlraum physics, radiation drive symmetry, capsule implosion physics in the radiation-drive approach of ICF. Recently, many efforts have been done to develop our 2D integrated simulation capability of laser fusion with a variety of optional physical models and numerical methods. In order to effectively integrate the existing codes and to facilitate the development of new codes, we are developing an object-oriented structured-mesh parallel code-supporting infrastructure, called JASMIN. Based on two-dimensional three-temperature hohlraum physics code LARED-H and two-dimensional multi-group radiative transfer code LARED-R, we develop a new generation two-dimensional laser fusion code under the JASMIN infrastructure, which enable us to simulate the whole process of laser fusion from the laser beams' entrance into the hohlraum to the end of implosion. In this paper, we will give a brief description of our new-generation two-dimensional laser fusion code, named LARED-Integration, especially in its physical models, and present some simulation results of holhraum. (authors)

Noise Residual Learning for Noise Modeling in Distributed Video Coding

DEFF Research Database (Denmark)

Luong, Huynh Van; Forchhammer, Søren

2012-01-01

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

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

Science.gov (United States)

Subotin, Michael; Davis, Anthony R

2016-09-01

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

SCANAIR a transient fuel performance code Part two: Assessment of modelling capabilities

Energy Technology Data Exchange (ETDEWEB)

Georgenthum, Vincent, E-mail: vincent.georgenthum@irsn.fr; Moal, Alain; Marchand, Olivier

2014-12-15

Highlights: • The SCANAIR code is devoted to the study of irradiated fuel rod behaviour during RIA. • The paper deals with the status of the code validation for PWR rods. • During the PCMI stage there is a good agreement between calculations and experiments. • The boiling crisis occurrence is rather well predicted. • The code assessment during the boiling crisis has still to be improved. - Abstract: In the frame of their research programmes on fuel safety, the French Institut de Radioprotection et de Sûreté Nucléaire develops the SCANAIR code devoted to the study of irradiated fuel rod behaviour during reactivity initiated accident. A first paper was focused on detailed modellings and code description. This second paper deals with the status of the code validation for pressurised water reactor rods performed thanks to the available experimental results. About 60 integral tests carried out in CABRI and NSRR experimental reactors and 24 separated tests performed in the PATRICIA facility (devoted to the thermal-hydraulics study) have been recalculated and compared to experimental data. During the first stage of the transient, the pellet clad mechanical interaction phase, there is a good agreement between calculations and experiments: the clad residual elongation and hoop strain of non failed tests but also the failure occurrence and failure enthalpy of failed tests are correctly calculated. After this first stage, the increase of cladding temperature can lead to the Departure from Nucleate Boiling. During the film boiling regime, the clad temperature can reach a very high temperature (>700 °C). If the boiling crisis occurrence is rather well predicted, the calculation of the clad temperature and the clad hoop strain during this stage have still to be improved.

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

DEFF Research Database (Denmark)

Huang, Xin; Forchhammer, Søren

2009-01-01

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

Development of the three dimensional flow model in the SPACE code

International Nuclear Information System (INIS)

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

2014-01-01

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

The drift flux model in the ASSERT subchannel code

International Nuclear Information System (INIS)

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

1987-01-01

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

Development of fast reactor containment safety analysis code, CONTAIN-LMR. (3) Improvement of sodium-concrete reaction model

International Nuclear Information System (INIS)

Kawaguchi, Munemichi; Doi, Daisuke; Seino, Hiroshi; Miyahara, Shinya

2015-01-01

A computer code, CONTAIN-LMR, is an integrated analysis tool to predict the consequence of severe accident in a liquid metal fast reactor. Because a sodium-concrete reaction behavior is one of the most important phenomena in the accident, a Sodium-Limestone Concrete Ablation Model (SLAM) has been developed and installed into the original CONTAIN code at Sandia National Laboratories (SNL) in the U.S. The SLAM treats chemical reaction kinetics between the sodium and the concrete compositions mechanistically using a three-region model, containing a pool (sodium and reaction debris) region, a dry (boundary layer (B/L) and dehydrated concrete) region, and a wet (hydrated concrete) region, the application is limited to the reaction between sodium and limestone concrete. In order to apply SLAM to the reaction between sodium and siliceous concrete which is an ordinary structural concrete in Japan, the chemical reaction kinetics model has been improved to consider the new chemical reactions between sodium and silicon dioxide. The improved model was validated to analyze a series of sodium-concrete experiments which were conducted in Japan Atomic Energy Agency (JAEA). It has been found that relatively good agreement between calculation and experimental results is obtained and the CONTAIN-LMR code has been validated with regard to the sodium-concrete reaction phenomena. (author)

MC21 v.6.0 - A continuous-energy Monte Carlo particle transport code with integrated reactor feedback capabilities

International Nuclear Information System (INIS)

Grieshemer, D.P.; Gill, D.F.; Nease, B.R.; Carpenter, D.C.; Joo, H.; Millman, D.L.; Sutton, T.M.; Stedry, M.H.; Dobreff, P.S.; Trumbull, T.H.; Caro, E.

2013-01-01

MC21 is a continuous-energy Monte Carlo radiation transport code for the calculation of the steady-state spatial distributions of reaction rates in three-dimensional models. The code supports neutron and photon transport in fixed source problems, as well as iterated-fission-source (eigenvalue) neutron transport problems. MC21 has been designed and optimized to support large-scale problems in reactor physics, shielding, and criticality analysis applications. The code also supports many in-line reactor feedback effects, including depletion, thermal feedback, xenon feedback, eigenvalue search, and neutron and photon heating. MC21 uses continuous-energy neutron/nucleus interaction physics over the range from 10 -5 eV to 20 MeV. The code treats all common neutron scattering mechanisms, including fast-range elastic and non-elastic scattering, and thermal- and epithermal-range scattering from molecules and crystalline materials. For photon transport, MC21 uses continuous-energy interaction physics over the energy range from 1 keV to 100 GeV. The code treats all common photon interaction mechanisms, including Compton scattering, pair production, and photoelectric interactions. All of the nuclear data required by MC21 is provided by the NDEX system of codes, which extracts and processes data from EPDL-, ENDF-, and ACE-formatted source files. For geometry representation, MC21 employs a flexible constructive solid geometry system that allows users to create spatial cells from first- and second-order surfaces. The system also allows models to be built up as hierarchical collections of previously defined spatial cells, with interior detail provided by grids and template overlays. Results are collected by a generalized tally capability which allows users to edit integral flux and reaction rate information. Results can be collected over the entire problem or within specific regions of interest through the use of phase filters that control which particles are allowed to score each

Development of Integrated Code for Risk Assessment (INCORIA) for Physical Protection System

International Nuclear Information System (INIS)

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

2010-01-01

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

Integration of Environmental Issues in a Physics Course: 'Physics by Inquiry' High School Teachers' Integration Models and Challenges

Science.gov (United States)

Kimori, David Abiya

As we approach the second quarter of the twenty-first century, one may predict that the environment will be among the dominant themes in the political and educational discourse. Over the past three decades, particular perspectives regarding the environment have begun to emerge: (i) realization by human beings that we not only live on earth and use its resources at an increasingly high rate but we also actually belong to the earth and the total ecology of all living systems, (ii) there are strong interactions among different components of the large and complex systems that make up our environment, and (iii) the rising human population and its impact on the environment is a great concern (Hughes & Mason, 2014). Studies have revealed that although the students do not have a deep understanding of environmental issues and lack environmental awareness and attitudes necessary for protecting the environment, they have great concern for the environment (Chapman & Sharma, 2001; Fien, Yencken, & Sykes, 2002). However, addressing environmental issues in the classroom and other disciplines has never been an easy job for teachers (Pennock & Bardwell, 1994; Edelson, 2007). Using multiple case studies, this study investigated how three purposefully selected physics teachers teaching a 'Physics by Inquiry' course integrated environmental topics and issues in their classroom. Particularly this study looked at what integration models and practices the three physics teachers employed in integrating environmental topics and issues in their classroom and what challenges the teachers faced while integrating environmental topics in their classrooms. Data collection methods including field notes taken from observations, teachers' interviews and a collection of artifacts and documents were used. The data were coded analyzed and organized into codes and categories guided by Fogarty (1991) models of curriculum integration and Ham and Sewing (1988) four categories of barriers to environmental

Progress in integrated energy-economy-environment model system development

International Nuclear Information System (INIS)

Yasukawa, Shigeru; Mankin, Shuichi; Sato, Osamu; Tadokoro, Yoshihiro; Nakano, Yasuyuki; Nagano, Takao

1987-11-01

The Integrated Energy-Economy-Environment Model System has been developed for providing analytical tools for the system analysis and technology assessments in the field of nuclear research and development. This model system consists of the following four model groups. The first model block installs 5 models and can serve to analyze and generate long-term scenarios on economy-energy-environment evolution. The second model block installs 2 models and can serve to analyze the structural transition phenomena in energy-economy-environment interactions. The third model block installs 2 models and can handle power reactor installation strategy problem and long-term fuel cycle analysis. The fourth model block installs 5 models and codes and can treats cost-benefit-risk analysis and assessments. This report describes mainly the progress and the outlines of application of the model system in these years after the first report on the research and development of the model system (JAERI-M 84 - 139). (author)

Code system to compute radiation dose in human phantoms

International Nuclear Information System (INIS)

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

1986-01-01

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

Using models in Integrated Ecosystem Assessment of coastal areas

Science.gov (United States)

Solidoro, Cosimo; Bandelj, Vinko; Cossarini, Gianpiero; Melaku Canu, Donata; Libralato, Simone

2014-05-01

Numerical Models can greatly contribute to integrated ecological assessment of coastal and marine systems. Indeed, models can: i) assist in the identification of efficient sampling strategy; ii) provide space interpolation and time extrapolation of experiemtanl data which are based on the knowedge on processes dynamics and causal realtionships which is coded within the model, iii) provide estimates of hardly measurable indicators. Furthermore model can provide indication on potential effects of implementation of alternative management policies. Finally, by providing a synthetic representation of an ideal system, based on its essential dynamic, model return a picture of ideal behaviour of a system in the absence of external perturbation, alteration, noise, which might help in the identification of reference behaivuor. As an important example, model based reanalyses of biogeochemical and ecological properties are an urgent need for the estimate of the environmental status and the assessment of efficacy of conservation and environmental policies, also with reference to the enforcement of the European MSFD. However, the use of numerical models, and particularly of ecological models, in modeling and in environmental management still is far from be the rule, possibly because of a lack in realizing the benefits which a full integration of modeling and montoring systems might provide, possibly because of a lack of trust in modeling results, or because many problems still exists in the development, validation and implementation of models. For istance, assessing the validity of model results is a complex process that requires the definition of appropriate indicators, metrics, methodologies and faces with the scarcity of real-time in-situ biogeochemical data. Furthermore, biogeochemical models typically consider dozens of variables which are heavily undersampled. Here we show how the integration of mathematical model and monitoring data can support integrated ecosystem

Field-based tests of geochemical modeling codes: New Zealand hydrothermal systems

International Nuclear Information System (INIS)

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

1993-12-01

Hydrothermal systems in the Taupo Volcanic Zone, North Island, New Zealand are being used as field-based modeling exercises for the EQ3/6 geochemical modeling code package. Comparisons of the observed state and evolution of the hydrothermal systems with predictions of fluid-solid equilibria made using geochemical modeling codes will determine how the codes can be used to predict the chemical and mineralogical response of the environment to nuclear waste emplacement. Field-based exercises allow us to test the models on time scales unattainable in the laboratory. Preliminary predictions of mineral assemblages in equilibrium with fluids sampled from wells in the Wairakei and Kawerau geothermal field suggest that affinity-temperature diagrams must be used in conjunction with EQ6 to minimize the effect of uncertainties in thermodynamic and kinetic data on code predictions

RCS modeling with the TSAR FDTD code

Energy Technology Data Exchange (ETDEWEB)

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

1992-03-01

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

Using Coding Apps to Support Literacy Instruction and Develop Coding Literacy

Science.gov (United States)

Hutchison, Amy; Nadolny, Larysa; Estapa, Anne

2016-01-01

In this article the authors present the concept of Coding Literacy and describe the ways in which coding apps can support the development of Coding Literacy and disciplinary and digital literacy skills. Through detailed examples, we describe how coding apps can be integrated into literacy instruction to support learning of the Common Core English…

GASFLOW computer code (physical models and input data)

International Nuclear Information System (INIS)

Muehlbauer, Petr

2007-11-01

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

Integrated modelling of material migration and target plate power handling at JET

International Nuclear Information System (INIS)

Coster, D.P.; Bonnin, X.; Chankin, A.

2005-01-01

The complexity of the tokamak edge and scrape-off layer (SOL) region is such that extrapolation to ITER requires modelling to be pursued through the integration of a number of edge codes, each of which must be thoroughly tested against results from present day machines. This contribution demonstrates how the edge modelling effort at JET is focused on such an approach by considering two examples, target power loading and material erosion and migration, the understanding of which are crucial issues for ITER. (author)

Psychometric properties of the Motivational Interviewing Treatment Integrity coding system 4.2 with jail inmates.

Science.gov (United States)

Owens, Mandy D; Rowell, Lauren N; Moyers, Theresa

2017-10-01

Motivational Interviewing (MI) is an evidence-based approach shown to be helpful for a variety of behaviors across many populations. Treatment fidelity is an important tool for understanding how and with whom MI may be most helpful. The Motivational Interviewing Treatment Integrity coding system was recently updated to incorporate new developments in the research and theory of MI, including the relational and technical hypotheses of MI (MITI 4.2). To date, no studies have examined the MITI 4.2 with forensic populations. In this project, twenty-two brief MI interventions with jail inmates were evaluated to test the reliability of the MITI 4.2. Validity of the instrument was explored using regression models to examine the associations between global scores (Empathy, Partnership, Cultivating Change Talk and Softening Sustain Talk) and outcomes. Reliability of this coding system with these data was strong. We found that therapists had lower ratings of Empathy with participants who had more extensive criminal histories. Both Relational and Technical global scores were associated with criminal histories as well as post-intervention ratings of motivation to decrease drug use. Findings indicate that the MITI 4.2 was reliable for coding sessions with jail inmates. Additionally, results provided information related to the relational and technical hypotheses of MI. Future studies can use the MITI 4.2 to better understand the mechanisms behind how MI works with this high-risk group. Published by Elsevier Ltd.

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

Science.gov (United States)

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

2018-07-01

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

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

International Nuclear Information System (INIS)

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

1992-06-01

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

JAPC Compact Simulator evolution to latest integration

International Nuclear Information System (INIS)

Nabeta, T.; Nakayama, Y.

1999-01-01

This paper describes the evolution of JAPC compact simulator from the first installation in 1988 until recent integration with SIMULATE-3 engineering code core model and extended simulation for Mid-loop operation and severe accidents. JAPC Compact Simulator has an advanced super compact rotating panel design. Three plants, Tokai 2 (GE BWR 5), Tsuruga 1 (GE BWR 2), Tsuruga 2 (MHI PWR 4-Loop) are simulated. The simulator has been used for training of operator and engineering personnel, and has continuously been upgraded to follow normal plant modifications as well as development in modeling and computer technology. The integration of SIMULATE-3 core model is, to our knowledge, the first integration of a real design code into a training simulator. SIMULATE-3 has been successfully integrated into the simulator and run in real time, without compromising the accuracy of SIMULATE-3. The code has been modified to also handle mid-loop operation and severe accidents. (author)

Comparison of experimental pulse-height distributions in germanium detectors with integrated-tiger-series-code predictions

International Nuclear Information System (INIS)

Beutler, D.E.; Halbleib, J.A.; Knott, D.P.

1989-01-01

This paper reports pulse-height distributions in two different types of Ge detectors measured for a variety of medium-energy x-ray bremsstrahlung spectra. These measurements have been compared to predictions using the integrated tiger series (ITS) Monte Carlo electron/photon transport code. In general, the authors find excellent agreement between experiments and predictions using no free parameters. These results demonstrate that the ITS codes can predict the combined bremsstrahlung production and energy deposition with good precision (within measurement uncertainties). The one region of disagreement observed occurs for low-energy (<50 keV) photons using low-energy bremsstrahlung spectra. In this case the ITS codes appear to underestimate the produced and/or absorbed radiation by almost an order of magnitude

Description of codes and models to be used in risk assessment

International Nuclear Information System (INIS)

1991-09-01

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

MoGIRE: A Model for Integrated Water Management

Science.gov (United States)

Reynaud, A.; Leenhardt, D.

2008-12-01

optimizes at each date (10 days step) the allocation of water across agricultural and urban water demands in order to maximize the social surplus derived from water consumption given the constraints imposed by the water network. An application of the model is proposed for the Neste system located in South-West of France. 67 regions competing for water allocation have been identified in the Neste system. Those regions are characterized by specific cropping systems, specific climate and soil characteristics and by their connections to the water network. The model, including the nodal representation of the water network, has been coded using the algebraic modeling language GAMS. We are currently analyzing the robustness of the approach through scenario testing. Keywords : Integrated water management, optimization-simulation model, agronomic-economic modeling, river basin.

Development of an Auto-Validation Program for MARS Code Assessments

International Nuclear Information System (INIS)

Lee, Young Jin; Chung, Bub Dong

2006-01-01

MARS (Multi-dimensional Analysis of Reactor Safety) code is a best-estimate thermal hydraulic system analysis code developed at KAERI. It is important for a thermal hydraulic computer code to be assessed against theoretical and experimental data to verify and validate the performance and the integrity of the structure, models and correlations of the code. The code assessment efforts for complex thermal hydraulics code such as MARS code can be tedious, time-consuming and require large amount of human intervention in data transfer to see the results in graphic forms. Code developers produce many versions of a code during development and each version need to be verified for integrity. Thus, for MARS code developers, it is desirable to have an automatic way of carrying out the code assessment calculations. In the present work, an Auto-Validation program that carries out the code assessment efforts has been developed. The program uses the user supplied configuration file (with '.vv' extension) which contain commands to read input file, to execute the user selected MARS program, and to generate result graphs. The program can be useful if a same set of code assessments is repeated with different versions of the code. The program is written with the Delphi program language. The program runs under the Microsoft Windows environment

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)

Embedded Systems Hardware Integration and Code Development for Maraia Capsule and E-MIST

Science.gov (United States)

Carretero, Emmanuel S.

2015-01-01

The cost of sending large spacecraft to orbit makes them undesirable for carrying out smaller scientific missions. Small spacecraft are more economical and can be tailored for missions where specific tasks need to be carried out, the Maraia capsule is such a spacecraft. Maraia will allow for samples of experiments conducted on the International Space Station to be returned to earth. The use of balloons to conduct experiments at the edge of space is a practical approach to reducing the large expense of using rockets. E-MIST is a payload designed to fly on a high altitude balloon. It can maintain science experiments in a controlled manner at the edge of space. The work covered here entails the integration of hardware onto each of the mentioned systems and the code associated with such work. In particular, the resistance temperature detector, pressure transducers, cameras, and thrusters for Maraia are discussed. The integration of the resistance temperature detectors and motor controllers to E-MIST is described. Several issues associated with sensor accuracy, code lock-up, and in-flight reset issues are mentioned. The solutions and proposed solutions to these issues are explained.

RFQ simulation code

International Nuclear Information System (INIS)

Lysenko, W.P.

1984-04-01

We have developed the RFQLIB simulation system to provide a means to systematically generate the new versions of radio-frequency quadrupole (RFQ) linac simulation codes that are required by the constantly changing needs of a research environment. This integrated system simplifies keeping track of the various versions of the simulation code and makes it practical to maintain complete and up-to-date documentation. In this scheme, there is a certain standard version of the simulation code that forms a library upon which new versions are built. To generate a new version of the simulation code, the routines to be modified or added are appended to a standard command file, which contains the commands to compile the new routines and link them to the routines in the library. The library itself is rarely changed. Whenever the library is modified, however, this modification is seen by all versions of the simulation code, which actually exist as different versions of the command file. All code is written according to the rules of structured programming. Modularity is enforced by not using COMMON statements, simplifying the relation of the data flow to a hierarchy diagram. Simulation results are similar to those of the PARMTEQ code, as expected, because of the similar physical model. Different capabilities, such as those for generating beams matched in detail to the structure, are available in the new code for help in testing new ideas in designing RFQ linacs

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

International Nuclear Information System (INIS)

Nelson, Hal T.

2012-01-01

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