Economic levels of thermal resistance for house envelopes: Considerations for a national energy code
International Nuclear Information System (INIS)
Swinton, M.C.; Sander, D.M.
1992-01-01
A code for energy efficiency in new buildings is being developed by the Standing Committee on Energy Conservation in Buildings. The precursor to the new code used national average energy rates and construction costs to determine economic optimum levels of insulation, and it is believed that this resulted in prescription of sub-optimum insulation levels in any region of Canada where energy or construction costs differ significantly from the average. A new approach for determining optimum levels of thermal insulation is proposed. The analytic techniques use month-by-month energy balances of heat loss and gain; use gain load ratio correlation (GLR) for predicting the fraction of useable free heat; increase confidence in the savings predictions for above grade envelopes; can take into account solar effects on windows; and are compatible with below-grade heat loss analysis techniques in use. A sensitivity analysis was performed to determine whether reasonable variations in house characteristics would cause significant differences in savings predicted. The life cycle costing technique developed will allow the selection of thermal resistances that are commonly met by industry. Environmental energy cost multipliers can be used with the proposed methodology, which could have a minor role in encouraging the next higher level of energy efficiency. 11 refs., 6 figs., 2 tabs
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.
International Nuclear Information System (INIS)
Akimoto, Hajime; Kukita; Ohnuki, Akira
1997-01-01
The Japan Atomic Energy Research Institute (JAERI) is conducting several research programs related to thermal-hydraulic and neutronic behavior of light water reactors (LWRs). These include LWR safety research projects, which are conducted in accordance with the Nuclear Safety Commission's research plan, and reactor engineering projects for the development of innovative reactor designs or core/fuel designs. Thermal-hydraulic and neutronic codes are used for various purposes including experimental analysis, nuclear power plant (NPP) safety analysis, and design assessment
Energy Technology Data Exchange (ETDEWEB)
Akimoto, Hajime; Kukita; Ohnuki, Akira [Japan Atomic Energy Research Institute, Ibaraki (Japan)
1997-07-01
The Japan Atomic Energy Research Institute (JAERI) is conducting several research programs related to thermal-hydraulic and neutronic behavior of light water reactors (LWRs). These include LWR safety research projects, which are conducted in accordance with the Nuclear Safety Commission`s research plan, and reactor engineering projects for the development of innovative reactor designs or core/fuel designs. Thermal-hydraulic and neutronic codes are used for various purposes including experimental analysis, nuclear power plant (NPP) safety analysis, and design assessment.
An improved thermal model for the computer code NAIAD
International Nuclear Information System (INIS)
Rainbow, M.T.
1982-12-01
An improved thermal model, based on the concept of heat slabs, has been incorporated as an option into the thermal hydraulic computer code NAIAD. The heat slabs are one-dimensional thermal conduction models with temperature independent thermal properties which may be internal and/or external to the fluid. Thermal energy may be added to or removed from the fluid via heat slabs and passed across the external boundary of external heat slabs at a rate which is a linear function of the external surface temperatures. The code input for the new option has been restructured to simplify data preparation. A full description of current input requirements is presented
Thermal-hydraulic code selection for modular high temperature gas-cooled reactors
Energy Technology Data Exchange (ETDEWEB)
Komen, E M.J.; Bogaard, J.P.A. van den
1995-06-01
In order to study the transient thermal-hydraulic system behaviour of modular high temperature gas-cooled reactors, the thermal-hydraulic computer codes RELAP5, MELCOR, THATCH, MORECA, and VSOP are considered at the Netherlands Energy Research Foundation ECN. This report presents the selection of the most appropriate codes. To cover the range of relevant accidents, a suite of three codes is recommended for analyses of HTR-M and MHTGR reactors. (orig.).
Current lead thermal analysis code 'CURRENT'
International Nuclear Information System (INIS)
Yamaguchi, Masahito; Tada, Eisuke; Shimamoto, Susumu; Hata, Kenichiro.
1985-08-01
Large gas-cooled current lead with the capacity more than 30 kA and 22 kV is required for superconducting toroidal and poloidal coils for fusion application. The current lead is used to carry electrical current from the power supply system at room temperature to the superconducting coil at 4 K. Accordingly, the thermal performance of the current lead is significantly important to determine the heat load requirements of the coil system at 4 K. Japan Atomic Energy Research Institute (JAERI) has being developed the large gas-cooled current leads with the optimum condition in which the heat load is around 1 W per 1 kA at 4 K. In order to design the current lead with the optimum thermal performances, JAERI developed thermal analysis code named as ''CURRENT'' which can theoretically calculate the optimum geometric shape and cooling conditions of the current lead. The basic equations and the instruction manual of the analysis code are described in this report. (author)
Monte Carlo simulation of a coded-aperture thermal neutron camera
International Nuclear Information System (INIS)
Dioszegi, I.; Salwen, C.; Forman, L.
2011-01-01
We employed the MCNPX Monte Carlo code to simulate image formation in a coded-aperture thermal-neutron camera. The camera, developed at Brookhaven National Laboratory (BNL), consists of a 20 x 17 cm"2 active area "3He-filled position-sensitive wire chamber in a cadmium enclosure box. The front of the box is a coded-aperture cadmium mask (at present with three different resolutions). We tested the detector experimentally with various arrangements of moderated point-neutron sources. The purpose of using the Monte Carlo modeling was to develop an easily modifiable model of the device to predict the detector's behavior using different mask patterns, and also to generate images of extended-area sources or large numbers (up to ten) of them, that is important for nonproliferation and arms-control verification, but difficult to achieve experimentally. In the model, we utilized the advanced geometry capabilities of the MCNPX code to simulate the coded aperture mask. Furthermore, the code simulated the production of thermal neutrons from fission sources surrounded by a thermalizer. With this code we also determined the thermal-neutron shadow cast by the cadmium mask; the calculations encompassed fast- and epithermal-neutrons penetrating into the detector through the mask. Since the process of signal production in "3He-filled position-sensitive wire chambers is well known, we omitted this part from our modeling. Simplified efficiency values were used for the three (thermal, epithermal, and fast) neutron-energy regions. Electronic noise and the room's background were included as a uniform irradiation component. We processed the experimental- and simulated-images using identical LabVIEW virtual instruments. (author)
User's manual for computer code SOLTES-1 (simulator of large thermal energy systems)
International Nuclear Information System (INIS)
Fewell, M.E.; Grandjean, N.R.; Dunn, J.C.; Edenburn, M.W.
1978-09-01
SOLTES simulates the steady-state response of thermal energy systems to time-varying data such as weather and loads. Thermal energy system models of both simple and complex systems can easily be modularly constructed from a library of routines. These routines mathematically model solar collectors, pumps, switches, thermal energy storage, thermal boilers, auxiliary boilers, heat exchangers, extraction turbines, extraction turbine/generators, condensers, regenerative heaters, air conditioners, heating and cooling of buildings, process vapor, etc.; SOLTES also allows user-supplied routines. The analyst need only specify fluid names to obtain readout of property data for heat-transfer fluids and constants that characterize power-cycle working fluids from a fluid property data bank. A load management capability allows SOLTES to simulate total energy systems that simultaneously follow heat and power loads and demands. Generalized energy accounting is available, and values for system performance parameters may be automatically determined by SOLTES. Because of its modularity and flexibility, SOLTES can be used to simulate a wide variety of thermal energy systems such as solar power/total energy, fossil fuel power plants/total energy, nuclear power plants/total energy, solar energy heating and cooling, geothermal energy, and solar hot water heaters
Regulatory requirements to the thermal-hydraulic and thermal-mechanical computer codes
International Nuclear Information System (INIS)
Vitkova, M.; Kalchev, B.; Stefanova, S.
2006-01-01
The paper presents an overview of the regulatory requirements to the thermal-hydraulic and thermal-mechanical computer codes, which are used for safety assessment of the fuel design and the fuel utilization. Some requirements to the model development, verification and validation of the codes and analysis of code uncertainties are also define. Questions concerning Quality Assurance during development and implementation of the codes as well as preparation of a detailed verification and validation plan are briefly discussed
COBRA-SFS [Spent Fuel Storage]: A thermal-hydraulic analysis computer code: Volume 2, User's manual
International Nuclear Information System (INIS)
Rector, D.R.; Cuta, J.M.; Lombardo, N.J.; Michener, T.E.; Wheeler, C.L.
1986-11-01
COBRA-SFS (Spent Fuel Storage) is a general thermal-hydraulic analysis computer code used to predict temperatures and velocities in a wide variety of systems. The code was refined and specialized for spent fuel storage system analyses for the US Department of Energy's Commercial Spent Fuel Management Program. The finite-volume equations governing mass, momentum, and energy conservation are written for an incompressible, single-phase fluid. The flow equations model a wide range of conditions including natural circulation. The energy equations include the effects of solid and fluid conduction, natural convection, and thermal radiation. The COBRA-SFS code is structured to perform both steady-state and transient calculations; however, the transient capability has not yet been validated. This volume contains the input instructions for COBRA-SFS and an auxiliary radiation exchange factor code, RADX-1. It is intended to aid the user in becoming familiar with the capabilities and modeling conventions of the code
Validation of thermal hydraulic computer codes for advanced light water reactor
International Nuclear Information System (INIS)
Macek, J.
2001-01-01
The Czech Republic operates 4 WWER-440 units, two WWER-1000 units are being finalised (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppressure system are modelled with RALOC and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems. An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. The paper provides a concise information on these activities of the NRI and its Thermal-hydraulics Department. A detailed example of the system code validation and the consequent utilisation of the results for a real NPP purposes is included. (author)
Performance testing of thermal analysis codes for nuclear fuel casks
International Nuclear Information System (INIS)
Sanchez, L.C.
1987-01-01
In 1982 Sandia National Laboratories held the First Industry/Government Joint Thermal and Structural Codes Information Exchange and presented the initial stages of an investigation of thermal analysis computer codes for use in the design of nuclear fuel shipping casks. The objective of the investigation was to (1) document publicly available computer codes, (2) assess code capabilities as determined from their user's manuals, and (3) assess code performance on cask-like model problems. Computer codes are required to handle the thermal phenomena of conduction, convection and radiation. Several of the available thermal computer codes were tested on a set of model problems to assess performance on cask-like problems. Solutions obtained with the computer codes for steady-state thermal analysis were in good agreement and the solutions for transient thermal analysis differed slightly among the computer codes due to modeling differences
JAERI thermal reactor standard code system for reactor design and analysis SRAC
International Nuclear Information System (INIS)
Tsuchihashi, Keichiro
1985-01-01
SRAC, JAERI thermal reactor standard code system for reactor design and analysis, developed in Japan Atomic Energy Research Institute, is for all types of thermal neutron nuclear design and analysis. The code system has undergone extensive verifications to confirm its functions, and has been used in core modification of the research reactor, detailed design of the multi-purpose high temperature gas reactor and analysis of the experiment with a critical assembly. In nuclear calculation with the code system, multi-group lattice calculation is first made with the libraries. Then, with the resultant homogeneous equivalent group constants, reactor core calculation is made. Described are the following: purpose and development of the code system, functions of the SRAC system, bench mark tests and usage state and future development. (Mori, K.)
Thermal-hydraulic interfacing code modules for CANDU reactors
Energy Technology Data Exchange (ETDEWEB)
Liu, W.S.; Gold, M.; Sills, H. [Ontario Hydro Nuclear, Toronto (Canada)] [and others
1997-07-01
The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.
Thermal-hydraulic interfacing code modules for CANDU reactors
International Nuclear Information System (INIS)
Liu, W.S.; Gold, M.; Sills, H.
1997-01-01
The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis
International Nuclear Information System (INIS)
Shamasundar, B.I.; Fehrenbach, M.E.
1981-05-01
The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations
An Improved Thermal Blooming Model for the Laser Performance Code Anchor
2016-06-01
over which a laser beam can maintain transverse coherence throughout its propagation distance. Typical values of ro are on the order of a few...G. Gebhardt, “Twenty-five years of thermal blooming: An overview,” in Proceedings of SPIE 1221 Propagation of High-Energy Laser Beams Through the...TERMS thermal blooming, atmospheric propagation , laser , scaling code, Strehl ratio, ANCHOR, COAMPS, NAVSLaM, LEEDR 15. NUMBER OF PAGES 77 16
International Nuclear Information System (INIS)
Kotegawa, Hiroshi; Sasamoto, Nobuo; Tanaka, Shun-ichi
1987-02-01
Both ''measured radioactive inventory due to neutron activation in the shield concrete of JPDR'' and ''measured intermediate and low energy neutron spectra penetrating through a graphite sphere'' are analyzed using a continuous energy model Monte Carlo code MCNP so as to estimate calculational accuracy of the code for neutron transport in thermal and epithermal energy regions. Analyses reveal that MCNP calculates thermal neutron spectra fairly accurately, while it apparently over-estimates epithermal neutron spectra (of approximate 1/E distribution) as compared with the measurements. (author)
Validation of containment thermal hydraulic computer codes for VVER reactor
Energy Technology Data Exchange (ETDEWEB)
Jiri Macek; Lubomir Denk [Nuclear Research Institute Rez plc Thermal-Hydraulic Analyses Department CZ 250 68 Husinec-Rez (Czech Republic)
2005-07-01
Full text of publication follows: The Czech Republic operates 4 VVER-440 units, two VVER-1000 units are being finalized (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppression system are modelled with COCOSYS and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems.An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. One of the important components of the VVER 440/213 NPP is its containment with pressure suppression system (bubble condenser). For safety analyses of this system, computer codes of the type MELCOR and COCOSYS are used in the Czech Republic. These codes were developed for containments of classic PWRs or BWRs. In order to apply these codes for VVER 440 systems, their validation on experimental facilities must be performed.The paper provides concise information on these activities of the NRI and its Thermal-hydraulics Department. The containment system of the VVER 440/213, its functions and approaches to solution of its safety is described with definition of acceptance criteria. A detailed example of the containment code validation on EREC Test facility (LOCA and MSLB) and the consequent utilisation of the results for a real NPP purposes is included. An approach to
Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD
Energy Technology Data Exchange (ETDEWEB)
Trambauer, K. [GRS, Garching (Germany)
1997-07-01
The system code ATHLET-CD is being developed by GRS in cooperation with IKE and IPSN. Its field of application comprises the whole spectrum of leaks and large breaks, as well as operational and abnormal transients for LWRs and VVERs. At present the analyses cover the in-vessel thermal-hydraulics, the early phases of core degradation, as well as fission products and aerosol release from the core and their transport in the Reactor Coolant System. The aim of the code development is to extend the simulation of core degradation up to failure of the reactor pressure vessel and to cover all physically reasonable accident sequences for western and eastern LWRs including RMBKs. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development. The code consists of four general modules to describe the reactor coolant system thermal-hydraulics, the core degradation, the fission product core release, and fission product and aerosol transport. Each general module consists of some basic modules which correspond to the process to be simulated or to its specific purpose. Besides the code structure based on the physical modelling, the code follows four strictly separated steps during the course of a calculation: (1) input of structure, geometrical data, initial and boundary condition, (2) initialization of derived quantities, (3) steady state calculation or input of restart data, and (4) transient calculation. In this paper, the transient solution method is briefly presented and the coupling methods are discussed. Three aspects have to be considered for the coupling of different modules in one code system. First is the conservation of masses and energy in the different subsystems as there are fluid, structures, and fission products and aerosols. Second is the convergence of the numerical solution and stability of the calculation. The third aspect is related to the code performance, and running time.
An engineering code to analyze hypersonic thermal management systems
Vangriethuysen, Valerie J.; Wallace, Clark E.
1993-01-01
Thermal loads on current and future aircraft are increasing and as a result are stressing the energy collection, control, and dissipation capabilities of current thermal management systems and technology. The thermal loads for hypersonic vehicles will be no exception. In fact, with their projected high heat loads and fluxes, hypersonic vehicles are a prime example of systems that will require thermal management systems (TMS) that have been optimized and integrated with the entire vehicle to the maximum extent possible during the initial design stages. This will not only be to meet operational requirements, but also to fulfill weight and performance constraints in order for the vehicle to takeoff and complete its mission successfully. To meet this challenge, the TMS can no longer be two or more entirely independent systems, nor can thermal management be an after thought in the design process, the typical pervasive approach in the past. Instead, a TMS that was integrated throughout the entire vehicle and subsequently optimized will be required. To accomplish this, a method that iteratively optimizes the TMS throughout the vehicle will not only be highly desirable, but advantageous in order to reduce the manhours normally required to conduct the necessary tradeoff studies and comparisons. A thermal management engineering computer code that is under development and being managed at Wright Laboratory, Wright-Patterson AFB, is discussed. The primary goal of the code is to aid in the development of a hypersonic vehicle TMS that has been optimized and integrated on a total vehicle basis.
Light-water-reactor coupled neutronic and thermal-hydraulic codes
International Nuclear Information System (INIS)
Diamond, D.J.
1982-01-01
An overview is presented of computer codes that model light water reactor cores with coupled neutronics and thermal-hydraulics. This includes codes for transient analysis and codes for steady state analysis which include fuel depletion and fission product buildup. Applications in nuclear design, reactor operations and safety analysis are given and the major codes in use in the USA are identified. The neutronic and thermal-hydraulic methodologies and other code features are outlined for three steady state codes (PDQ7, NODE-P/B and SIMULATE) and four dynamic codes (BNL-TWIGL, MEKIN, RAMONA-3B, RETRAN-02). Speculation as to future trends with such codes is also presented
Development of thermal hydraulic evaluation code for CANDU reactors
International Nuclear Information System (INIS)
Kim, Man Woong; Yu, Seon Oh; Choi, Yong Seog; Shin, Chull; Hwang, Soo Hyun
2004-02-01
To enhance the safety of operating CANDU reactors, the establishment of the safety analysis codes system for CANDU reactors is in progress. As for the development of thermal-hydraulic analysis code for CANDU system, the studies for improvement of evaluation model inside RELAP/CANDU code and the development of safety assessment methodology for GSI (Generic Safety Issues) are in progress as a part of establishment of CANDU safety assessment system. To develop the 3-D thermal-hydraulic analysis code for moderator system, the CFD models for analyzing the CANDU-6 moderator circulation are developed. One model uses a structured grid system with the porous media approach for the 380 Calandria tubes in the core region. The other uses a unstructured grid system on the real geometry of 380 Calandria tubes, so that the detailed fluid flow between the Calandria tubes can be observed. As to the development of thermal-hydraulic analysis code for containment, the study on the applicability of CONTAIN 2.0 code to a CANDU containment was conducted and a simulation of the thermal-hydraulic phenomena during the accident was performed. Besides, the model comparison of ESFs (Engineered Safety Features) inside CONTAIN 2.0 code and PRESCON code has also conducted
International Nuclear Information System (INIS)
Oliveira Panao, Marta J.N.; Camelo, Susana M.L.; Goncalves, Helder J.P.
2011-01-01
In this paper, cooling energy needs are calculated by the steady-state methodology of the Portuguese building thermal code. After the first period of building code implementation, re-evaluation according to EN ISO 13790 is recommended in order to compare results with the dynamic simulation results. From these analyses, a newly revised methodology arises including a few corrections in procedure. This iterative result is sufficiently accurate to calculate the building's cooling energy needs. Secondly, results show that the required conditions are insufficient to prevent overheating. The use of the gain utilization factor as an overheating risk index is suggested, according to an adaptive comfort protocol, and is integrated in the method used to calculate the maximum value for cooling energy needs. This proposed streamlined method depends on reference values: window-to-floor area ratio, window shading g-value, integrated solar radiation and gain utilization factor, which leads to threshold values significantly below the ones currently used. These revised requirements are more restrictive and, therefore, will act to improve a building's thermal performance during summer. As a rule of thumb applied for Portuguese climates, the reference gain utilization factor should assume a minimum value of 0.8 for a latitude angle range of 40-41 o N, 0.6 for 38-39 o N and 0.5 for 37 o N. -- Highlights: → A newly revised methodology for Portuguese building thermal code. → The use of the gain utilization factor as an overheating risk index is suggested. → The proposed streamlined method depends on reference values. → Threshold maximum values are significantly below the ones currently used.
Current and anticipated uses of thermal hydraulic codes in Korea
Energy Technology Data Exchange (ETDEWEB)
Kim, Kyung-Doo; Chang, Won-Pyo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1997-07-01
In Korea, the current uses of thermal hydraulic codes are categorized into 3 areas. The first application is in designing both nuclear fuel and NSSS. The codes have usually been introduced based on the technology transfer programs agreed between KAERI and the foreign vendors. Another area is in the supporting of the plant operations and licensing by the utility. The third category is research purposes. In this area assessments and some applications to the safety issue resolutions are major activities using the best estimate thermal hydraulic codes such as RELAP5/MOD3 and CATHARE2. Recently KEPCO plans to couple thermal hydraulic codes with a neutronics code for the design of the evolutionary type reactor by 2004. KAERI also plans to develop its own best estimate thermal hydraulic code, however, application range is different from KEPCO developing code. Considering these activities, it is anticipated that use of the best estimate hydraulic analysis code developed in Korea may be possible in the area of safety evaluation within 10 years.
A flexible coupling scheme for Monte Carlo and thermal-hydraulics codes
Energy Technology Data Exchange (ETDEWEB)
Hoogenboom, J. Eduard, E-mail: J.E.Hoogenboom@tudelft.nl [Delft University of Technology (Netherlands); Ivanov, Aleksandar; Sanchez, Victor, E-mail: Aleksandar.Ivanov@kit.edu, E-mail: Victor.Sanchez@kit.edu [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology, Eggenstein-Leopoldshafen (Germany); Diop, Cheikh, E-mail: Cheikh.Diop@cea.fr [CEA/DEN/DANS/DM2S/SERMA, Commissariat a l' Energie Atomique, Gif-sur-Yvette (France)
2011-07-01
A coupling scheme between a Monte Carlo code and a thermal-hydraulics code is being developed within the European NURISP project for comprehensive and validated reactor analysis. The scheme is flexible as it allows different Monte Carlo codes and different thermal-hydraulics codes to be used. At present the MCNP and TRIPOLI4 Monte Carlo codes can be used and the FLICA4 and SubChanFlow thermal-hydraulics codes. For all these codes only an original executable is necessary. A Python script drives the iterations between Monte Carlo and thermal-hydraulics calculations. It also calls a conversion program to merge a master input file for the Monte Carlo code with the appropriate temperature and coolant density data from the thermal-hydraulics calculation. Likewise it calls another conversion program to merge a master input file for the thermal-hydraulics code with the power distribution data from the Monte Carlo calculation. Special attention is given to the neutron cross section data for the various required temperatures in the Monte Carlo calculation. Results are shown for an infinite lattice of PWR fuel pin cells and a 3 x 3 fuel BWR pin cell cluster. Various possibilities for further improvement and optimization of the coupling system are discussed. (author)
A flexible coupling scheme for Monte Carlo and thermal-hydraulics codes
International Nuclear Information System (INIS)
Hoogenboom, J. Eduard; Ivanov, Aleksandar; Sanchez, Victor; Diop, Cheikh
2011-01-01
A coupling scheme between a Monte Carlo code and a thermal-hydraulics code is being developed within the European NURISP project for comprehensive and validated reactor analysis. The scheme is flexible as it allows different Monte Carlo codes and different thermal-hydraulics codes to be used. At present the MCNP and TRIPOLI4 Monte Carlo codes can be used and the FLICA4 and SubChanFlow thermal-hydraulics codes. For all these codes only an original executable is necessary. A Python script drives the iterations between Monte Carlo and thermal-hydraulics calculations. It also calls a conversion program to merge a master input file for the Monte Carlo code with the appropriate temperature and coolant density data from the thermal-hydraulics calculation. Likewise it calls another conversion program to merge a master input file for the thermal-hydraulics code with the power distribution data from the Monte Carlo calculation. Special attention is given to the neutron cross section data for the various required temperatures in the Monte Carlo calculation. Results are shown for an infinite lattice of PWR fuel pin cells and a 3 x 3 fuel BWR pin cell cluster. Various possibilities for further improvement and optimization of the coupling system are discussed. (author)
Comparative study of Thermal Hydraulic Analysis Codes for Pressurized Water Reactors
Energy Technology Data Exchange (ETDEWEB)
Kim, Yang Hoon; Jang, Mi Suk; Han, Kee Soo [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)
2015-05-15
Various codes are used for the thermal hydraulic analysis of nuclear reactors. The use of some codes among these is limited by user and some codes are not even open to general person. Thus, the use of alternative code is considered for some analysis. In this study, simple thermal hydraulic behaviors are analyzed using three codes to show that alternative codes are possible for the analysis of nuclear reactors. We established three models of the simple u-tube manometer using three different codes. RELAP5 (Reactor Excursion and Leak Analysis Program), SPACE (Safety and Performance Analysis CodE for nuclear power Plants), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are selected for this analysis. RELAP5 is widely used codes for the analysis of system behavior of PWRs. SPACE has been developed based on RELAP5 for the analysis of system behavior of PWRs and licensing of the code is in progress. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. The internal behavior of u-tube manometer was analyzed by RELAP5, SPACE and GOTHIC codes. The general transient behavior was similar among 3 codes. However, the stabilized status of the transient period analyzed by REPAP5 was different from the other codes. It would be resulted from the different physical models used in the other codes, which is specialized for the multi-phase thermal hydraulic behavior analysis.
Energy Technology Data Exchange (ETDEWEB)
Ahn, S.H., E-mail: k175ash@kins.re.kr [Korea Institute of Nuclear Safety (KINS) (Korea, Republic of); Aksan, N., E-mail: nusr.aksan@gmail.com [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Austregesilo, H., E-mail: henrique.austregesilo@grs.de [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) (Germany); Bestion, D., E-mail: dominique.bestion@cea.fr [Commissariat à l’énergie atomique et aux énergies alternatives (CEA) (France); Chung, B.D., E-mail: bdchung@kaeri.re.kr [Korea Atomic Energy Research Institute (KAERI) (Korea, Republic of); D’Auria, F., E-mail: f.dauria@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Emonot, P., E-mail: philippe.emonot@cea.fr [Commissariat à l’énergie atomique et aux énergies alternatives (CEA) (France); Gandrille, J.L., E-mail: jeanluc.gandrille@areva.com [AREVA NP (France); Hanninen, M., E-mail: markku.hanninen@vtt.fi [VTT Technical Research Centre of Finland (VTT) (Finland); Horvatović, I., E-mail: i.horvatovic@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Kim, K.D., E-mail: kdkim@kaeri.re.kr [Korea Atomic Energy Research Institute (KAERI) (Korea, Republic of); Kovtonyuk, A., E-mail: a.kovtonyuk@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Petruzzi, A., E-mail: a.petruzzi@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy)
2015-01-15
Highlights: • We briefly presented the project called Forum and Network of System Thermal-Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics (FONESYS). • We presented FONESYS participants and their codes. • We explained FONESYS projects motivation, its main targets and working modalities. • We presented FONESYS position about projects topics and subtopics. - Abstract: The purpose of this article is to present briefly the project called Forum and Network of System Thermal-Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics (FONESYS), its participants, the motivation for the project, its main targets and working modalities. System Thermal-Hydraulics (SYS-TH) codes, also as part of the Best Estimate Plus Uncertainty (BEPU) approaches, are expected to achieve a more-and-more relevant role in nuclear reactor technology, safety and design. Namely, the number of code-users can easily be predicted to increase in the countries where nuclear technology is exploited. Thus, the idea of establishing a forum and a network among the code developers and with possible extension to code users has started to have major importance and value. In this framework the FONESYS initiative has been created. The main targets of FONESYS are: • To promote the use of SYS-TH Codes and the application of the BEPU approaches. • To establish acceptable and recognized procedures and thresholds for Verification and Validation (V and V). • To create a common ground for discussing envisaged improvements in various areas, including user-interface, and the connection with other numerical tools, including Computational Fluid Dynamics (CFD) Codes.
Current and anticipated uses of thermal-hydraulic codes in NFI
Energy Technology Data Exchange (ETDEWEB)
Tsuda, K. [Nuclear Fuel Industries, Ltd., Tokyo (Japan); Takayasu, M. [Nuclear Fuel Industries, Ltd., Sennann-gun (Japan)
1997-07-01
This paper presents the thermal-hydraulic codes currently used in NFI for the LWR fuel development and licensing application including transient and design basis accident analyses of LWR plants. The current status of the codes are described in the context of code capability, modeling feature, and experience of code application related to the fuel development and licensing. Finally, the anticipated use of the future thermal-hydraulic code in NFI is briefly given.
Energy Technology Data Exchange (ETDEWEB)
Vitruk, S.G.; Korsun, A.S. [Moscow Engineering Physics Institute (Russian Federation); Ushakov, P.A. [Institute of Physics and Power Engineering, Obninsk (R)] [and others
1995-09-01
The multilevel mathematical model of neutron thermal hydrodynamic processes in a passive safety core without assemblies duct walls and appropriate computer code SKETCH, consisted of thermal hydrodynamic module THEHYCO-3DT and neutron one, are described. A new effective discretization technique for energy, momentum and mass conservation equations is applied in hexagonal - z geometry. The model adequacy and applicability are presented. The results of the calculations show that the model and the computer code could be used in conceptual design of advanced reactors.
International Nuclear Information System (INIS)
Vitruk, S.G.; Korsun, A.S.; Ushakov, P.A.
1995-01-01
The multilevel mathematical model of neutron thermal hydrodynamic processes in a passive safety core without assemblies duct walls and appropriate computer code SKETCH, consisted of thermal hydrodynamic module THEHYCO-3DT and neutron one, are described. A new effective discretization technique for energy, momentum and mass conservation equations is applied in hexagonal - z geometry. The model adequacy and applicability are presented. The results of the calculations show that the model and the computer code could be used in conceptual design of advanced reactors
How good are thermal-hydraulics codes for analyses of plant transients
International Nuclear Information System (INIS)
Fabic, S.
1996-01-01
In the early seventies, all thermal-hydraulics codes were based on the Homogeneous Equilibrium Model (HEM), represented by three conservation equations: mixture mass, momentum and energy. Various means were utilized to solve the resulting system of equations: finite differences in FLASH, SATAN, RELAP3 and RELAP4, method of characteristics in BLOWDWN2, loop momentum method in RAMONA and NORCOOL, and others. As the result the world came to regard HEM as too restrictive and the Two-Fluid model came into fashion, first featuring a six and later, a seven-equation model. New codes like KACHINA, TRAC and RELAP5 were developed also. Experience and comparisons with test data have recently forced us to wonder whether the ability to 'compute' while considering great many complexities, ran ahead of the ability to competently define various interactions between fluid phases and components that such complex codes require. The long running times are also a problem that needs to be resolved. More recent trends in the treatment of thermal-hydraulics in Power Plant Simulators and in Plant Analyzers will also be discussed
Neutron fluence rate and energy spectrum in SPRR-300 reactor thermal column
International Nuclear Information System (INIS)
Dou Haifeng; Dai Junlong
2006-01-01
In order to modify the simple one-dimension model, the neutron fluence rate distribution calculated with ANISN code ws checked with that calculated with MCNP code. To modify the error caused by ignoring the neutron landscape orientation leaking, the reflector that can't be modeled in a simple one-dimension model was dealt by extending landscape orientation scale. On this condition the neutron fluence rate distribution and the energy spectrum in the thermal column of SPRR-300 reactor were calculated with one-dimensional code ANISN, and the results of Cd ratio are well accorded with the experimental results. The deviation between them is less than 5% and it isn't above 10% in one or two special positions. It indicates that neutron fluence rate distribution and energy spectrum in the thermal column can be well calculated with one-dimensional code ANISN. (authors)
Verification of combined thermal-hydraulic and heat conduction analysis code FLOWNET/TRUMP
International Nuclear Information System (INIS)
Maruyama, Soh; Fujimoto, Nozomu; Sudo, Yukio; Kiso, Yoshihiro; Murakami, Tomoyuki.
1988-09-01
This report presents the verification results of the combined thermal-hydraulic and heat conduction analysis code, FLOWNET/TRUMP which has been utilized for the core thermal hydraulic design, especially for the analysis of flow distribution among fuel block coolant channels, the determination of thermal boundary conditions for fuel block stress analysis and the estimation of fuel temperature in the case of fuel block coolant channel blockage accident in the design of the High Temperature Engineering Test Reactor(HTTR), which the Japan Atomic Energy Research Institute has been planning to construct in order to establish basic technologies for future advanced very high temperature gas-cooled reactors and to be served as an irradiation test reactor for promotion of innovative high temperature new frontier technologies. The verification of the code was done through the comparison between the analytical results and experimental results of the Helium Engineering Demonstration Loop Multi-channel Test Section(HENDEL T 1-M ) with simulated fuel rods and fuel blocks. (author)
Verification of combined thermal-hydraulic and heat conduction analysis code FLOWNET/TRUMP
Maruyama, Soh; Fujimoto, Nozomu; Kiso, Yoshihiro; Murakami, Tomoyuki; Sudo, Yukio
1988-09-01
This report presents the verification results of the combined thermal-hydraulic and heat conduction analysis code, FLOWNET/TRUMP which has been utilized for the core thermal hydraulic design, especially for the analysis of flow distribution among fuel block coolant channels, the determination of thermal boundary conditions for fuel block stress analysis and the estimation of fuel temperature in the case of fuel block coolant channel blockage accident in the design of the High Temperature Engineering Test Reactor(HTTR), which the Japan Atomic Energy Research Institute has been planning to construct in order to establish basic technologies for future advanced very high temperature gas-cooled reactors and to be served as an irradiation test reactor for promotion of innovative high temperature new frontier technologies. The verification of the code was done through the comparison between the analytical results and experimental results of the Helium Engineering Demonstration Loop Multi-channel Test Section(HENDEL T(sub 1-M)) with simulated fuel rods and fuel blocks.
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.
Thermal energy storage devices, systems, and thermal energy storage device monitoring methods
Tugurlan, Maria; Tuffner, Francis K; Chassin, David P.
2016-09-13
Thermal energy storage devices, systems, and thermal energy storage device monitoring methods are described. According to one aspect, a thermal energy storage device includes a reservoir configured to hold a thermal energy storage medium, a temperature control system configured to adjust a temperature of the thermal energy storage medium, and a state observation system configured to provide information regarding an energy state of the thermal energy storage device at a plurality of different moments in time.
International Nuclear Information System (INIS)
2001-05-01
An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for
Thermal Comfort and Ventilation Criteria for low Energy Residential Buildings in Building Codes
DEFF Research Database (Denmark)
Cao, Guangyu; Kurnitski, Jarek; Awbi, Hazim
2012-01-01
of the indoor air quality in such buildings. Currently, there are no global guidelines for specifying the indoor thermal environment in such low-energy buildings. The objective of this paper is to analyse the classification of indoor thermal comfort levels and recommended ventilation rates for different low...
International Nuclear Information System (INIS)
Rector, D.R.; Wheeler, C.L.; Lombardo, N.J.
1986-11-01
COBRA-SFS (Spent Fuel Storage) is a general thermal-hydraulic analysis computer code used to predict temperatures and velocities in a wide variety of systems. The code was refined and specialized for spent fuel storage system analyses for the US Department of Energy's Commercial Spent Fuel Management Program. The finite-volume equations governing mass, momentum, and energy conservation are written for an incompressible, single-phase fluid. The flow equations model a wide range of conditions including natural circulation. The energy equations include the effects of solid and fluid conduction, natural convection, and thermal radiation. The COBRA-SFS code is structured to perform both steady-state and transient calculations: however, the transient capability has not yet been validated. This volume describes the finite-volume equations and the method used to solve these equations. It is directed toward the user who is interested in gaining a more complete understanding of these methods
Thermal energy storage apparatus, controllers and thermal energy storage control methods
Hammerstrom, Donald J.
2016-05-03
Thermal energy storage apparatus, controllers and thermal energy storage control methods are described. According to one aspect, a thermal energy storage apparatus controller includes processing circuitry configured to access first information which is indicative of surpluses and deficiencies of electrical energy upon an electrical power system at a plurality of moments in time, access second information which is indicative of temperature of a thermal energy storage medium at a plurality of moments in time, and use the first and second information to control an amount of electrical energy which is utilized by a heating element to heat the thermal energy storage medium at a plurality of moments in time.
Thermal-hydraulic codes validation for safety analysis of NPPs with RBMK
International Nuclear Information System (INIS)
Brus, N.A.; Ioussoupov, O.E.
2000-01-01
This work is devoted to validation of western thermal-hydraulic codes (RELAP5/MOD3 .2 and ATHLET 1.1 Cycle C) in application to Russian designed light water reactors. Such validation is needed due to features of RBMK reactor design and thermal-hydraulics in comparison with PWR and BWR reactors, for which these codes were developed and validated. These validation studies are concluded with a comparison of calculation results of modeling with the thermal-hydraulics codes with the experiments performed earlier using the thermal-hydraulics test facilities with the experimental data. (authors)
Comparison of the thermal neutron scattering treatment in MCNP6 and GEANT4 codes
Tran, H. N.; Marchix, A.; Letourneau, A.; Darpentigny, J.; Menelle, A.; Ott, F.; Schwindling, J.; Chauvin, N.
2018-06-01
To ensure the reliability of simulation tools, verification and comparison should be made regularly. This paper describes the work performed in order to compare the neutron transport treatment in MCNP6.1 and GEANT4-10.3 in the thermal energy range. This work focuses on the thermal neutron scattering processes for several potential materials which would be involved in the neutron source designs of Compact Accelerator-based Neutrons Sources (CANS), such as beryllium metal, beryllium oxide, polyethylene, graphite, para-hydrogen, light water, heavy water, aluminium and iron. Both thermal scattering law and free gas model, coming from the evaluated data library ENDF/B-VII, were considered. It was observed that the GEANT4.10.03-patch2 version was not able to account properly the coherent elastic process occurring in crystal lattice. This bug is treated in this work and it should be included in the next release of the code. Cross section sampling and integral tests have been performed for both simulation codes showing a fair agreement between the two codes for most of the materials except for iron and aluminium.
Energy Savings Analysis of the Proposed NYStretch-Energy Code 2018
Energy Technology Data Exchange (ETDEWEB)
Liu, Bing [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Zhang, Jian [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Chen, Yan [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Edelson, Jim [New Buildings Inst. (NBI), Portland, OR (United States); Lyles, Mark [New Buildings Inst. (NBI), Portland, OR (United States)
2018-01-20
This study was conducted by the Pacific Northwest National Laboratory (PNNL) in support of the stretch energy code development led by the New York State Energy Research and Development Authority (NYSERDA). In 2017 NYSERDA developed its 2016 Stretch Code Supplement to the 2016 New York State Energy Conservation Construction Code (hereinafter referred to as “NYStretch-Energy”). NYStretch-Energy is intended as a model energy code for statewide voluntary adoption that anticipates other code advancements culminating in the goal of a statewide Net Zero Energy Code by 2028. Since then, NYSERDA continues to develop the NYStretch-Energy Code 2018 edition. To support the effort, PNNL conducted energy simulation analysis to quantify the energy savings of proposed commercial provisions of the NYStretch-Energy Code (2018) in New York. The focus of this project is the 20% improvement over existing commercial model energy codes. A key requirement of the proposed stretch code is that it be ‘adoptable’ as an energy code, meaning that it must align with current code scope and limitations, and primarily impact building components that are currently regulated by local building departments. It is largely limited to prescriptive measures, which are what most building departments and design projects are most familiar with. This report describes a set of energy-efficiency measures (EEMs) that demonstrate 20% energy savings over ANSI/ASHRAE/IES Standard 90.1-2013 (ASHRAE 2013) across a broad range of commercial building types and all three climate zones in New York. In collaboration with New Building Institute, the EEMs were developed from national model codes and standards, high-performance building codes and standards, regional energy codes, and measures being proposed as part of the on-going code development process. PNNL analyzed these measures using whole building energy models for selected prototype commercial buildings and multifamily buildings representing buildings in New
Thermal-hydraulic analysis of the Three Mile Island Unit 2 reactor accident with THALES code
International Nuclear Information System (INIS)
Hashimoto, Kazuichiro; Soda, Kunihisa
1991-10-01
The OECD Nuclear Energy Agency (NEA) has established a Task Group in the Committee on the Safety of Nuclear Installations (CSNI) to perform an analysis of Three Mile Island Unit 2 (TMI-2) accident as a standard problem to benchmark severe accident computer codes and to assess the capability of the codes. The TMI-2 Analysis Exercise was performed at the Japan Atomic Energy Research Institute (JAERI) using the THALES (Thermal-Hydraulic Analysis of Loss-of-Coolant, Emergency Core Cooling and Severe Core Damage) - PM1/TMI code. The purpose of the analysis is to verify the capability of THALES-PM1/TMI code to describe accident progression in the actual plant. The present paper describes the final result of the TMI-2 Analysis Exercise performed at JAERI. (author)
75 FR 20833 - Building Energy Codes
2010-04-21
...-0012] Building Energy Codes AGENCY: Office of Energy Efficiency and Renewable Energy, Department of... the current model building energy codes or their equivalent. DOE is interested in better understanding... codes, Standard 90.1-2007, Energy Standard for Buildings Except Low-Rise Residential Buildings (or...
Development of realistic thermal hydraulic system analysis code
Energy Technology Data Exchange (ETDEWEB)
Lee, Won Jae; Chung, B. D; Kim, K. D. [and others
2002-05-01
The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others.
Development of realistic thermal hydraulic system analysis code
International Nuclear Information System (INIS)
Lee, Won Jae; Chung, B. D; Kim, K. D.
2002-05-01
The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others
Canadian energy standards : residential energy code requirements
Energy Technology Data Exchange (ETDEWEB)
Cooper, K. [SAR Engineering Ltd., Burnaby, BC (Canada)
2006-09-15
A survey of residential energy code requirements was discussed. New housing is approximately 13 per cent more efficient than housing built 15 years ago, and more stringent energy efficiency requirements in building codes have contributed to decreased energy use and greenhouse gas (GHG) emissions. However, a survey of residential energy codes across Canada has determined that explicit demands for energy efficiency are currently only present in British Columbia (BC), Manitoba, Ontario and Quebec. The survey evaluated more than 4300 single-detached homes built between 2000 and 2005 using data from the EnerGuide for Houses (EGH) database. House area, volume, airtightness and construction characteristics were reviewed to create archetypes for 8 geographic areas. The survey indicated that in Quebec and the Maritimes, 90 per cent of houses comply with ventilation system requirements of the National Building Code, while compliance in the rest of Canada is much lower. Heat recovery ventilation use is predominant in the Atlantic provinces. Direct-vent or condensing furnaces constitute the majority of installed systems in provinces where natural gas is the primary space heating fuel. Details of Insulation levels for walls, double-glazed windows, and building code insulation standards were also reviewed. It was concluded that if R-2000 levels of energy efficiency were applied, total average energy consumption would be reduced by 36 per cent in Canada. 2 tabs.
Energy Technology Data Exchange (ETDEWEB)
Washington State Energy Code Program
1992-05-01
This manual is designed to provide building department personnel with specific inspection and plan review skills and information on provisions of the 1991 edition of the Washington State Energy Code (WSEC). It also provides information on provisions of the new stand-alone Ventilation and Indoor Air Quality (VIAQ) Code.The intent of the WSEC is to reduce the amount of energy used by requiring energy-efficient construction. Such conservation reduces energy requirements, and, as a result, reduces the use of finite resources, such as gas or oil. Lowering energy demand helps everyone by keeping electricity costs down. (It is less expensive to use existing electrical capacity efficiently than it is to develop new and additional capacity needed to heat or cool inefficient buildings.) The new VIAQ Code (effective July, 1991) is a natural companion to the energy code. Whether energy-efficient or not, an homes have potential indoor air quality problems. Studies have shown that indoor air is often more polluted than outdoor air. The VIAQ Code provides a means of exchanging stale air for fresh, without compromising energy savings, by setting standards for a controlled ventilation system. It also offers requirements meant to prevent indoor air pollution from building products or radon.
International Nuclear Information System (INIS)
Matausek, M. V.
1966-06-01
A combination of multigroup method and P 3 approximation of spherical harmonics method was chosen for calculating space-energy distribution of thermal neutron flux in elementary reactor cell. Application of these methods reduced solution of complicated transport equation to the problem of solving an inhomogeneous system of six ordinary firs-order differential equations. A procedure is proposed which avoids numerical solution and enables analytical solution when applying certain approximations. Based on this approach, computer codes were written for ZUSE-Z-23 computer: SIGMA code for calculating group constants for a given material; MULTI code which uses results of SIGMA code as input and calculates spatial ana energy distribution of thermal neutron flux in a reactor cell. Calculations of thermal neutron spectra for a number of reactor cells were compared to results available from literature. Agreement was satisfactory in all the cases, which proved the correctness of the applied method. Some possibilities for improving the precision and acceleration of the calculation process were found during calculation. (author)
Two-dimensional disruption thermal analysis code DREAM
International Nuclear Information System (INIS)
Yamazaki, Seiichiro; Kobayashi, Takeshi; Seki, Masahiro.
1988-08-01
When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing components such as first wall and divertor/limiter are subjected to an intense heat load with very high heat flux and short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs, it causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes (melting/evaporation) and radiation heat loss is required in the design of these components. This paper describes the computer code DREAM developed to perform the two-dimensional transient thermal analysis that takes phase changes and radiation into account. The input and output of the code and a sample analysis on a disruption simulation experiment are also reported. The user's input manual is added as an appendix. The profiles and time variations of temperature, and melting and evaporated thicknesses of the material subjected to intense heat load can be obtained, using this computer code. This code also gives the temperature data for elastoplastic analysis with FEM structural analysis codes (ADINA, MARC, etc.) to evaluate the thermal stress and crack propagation behavior within the wall materials. (author)
Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS
Energy Technology Data Exchange (ETDEWEB)
Lee, Young Jin; Chung, Bub Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
2004-07-01
MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures.
Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS
International Nuclear Information System (INIS)
Lee, Young Jin; Chung, Bub Dong
2004-01-01
MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures
Gas-cooled reactor thermal-hydraulics using CAST3M and CRONOS2 codes
International Nuclear Information System (INIS)
Studer, E.; Coulon, N.; Stietel, A.; Damian, F.; Golfier, H.; Raepsaet, X.
2003-01-01
The CEA R and D program on advanced Gas Cooled Reactors (GCR) relies on different concepts: modular High Temperature Reactor (HTR), its evolution dedicated to hydrogen production (Very High Temperature Reactor) and Gas Cooled Fast Reactors (GCFR). Some key safety questions are related to decay heat removal during potential accident. This is strongly connected to passive natural convection (including gas injection of Helium, CO 2 , Nitrogen or Argon) or forced convection using active safety systems (gas blowers, heat exchangers). To support this effort, thermal-hydraulics computer codes will be necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Accurate and efficient modeling of heat transfer by conduction, convection or thermal radiation as well as energy storage are necessary requirements to obtain a high level of confidence in the thermal-hydraulic simulations. To achieve that goal a thorough validation process has to ve conducted. CEA's CAST3M code dedicated to GCR thermal-hydraulics has been validated against different test cases: academic interaction between natural convection and thermal radiation, small scale in-house THERCE experiments and large scale High Temperature Test Reactor benchmarks such as HTTR-VC benchmark. Coupling with neutronics is also an important modeling aspect for the determination of neutronic parameters such as neutronic coefficient (Doppler, moderator,...), critical position of control rods...CEA's CAST3M and CRONOS2 computer codes allow this coupling and a first example of coupled thermal-hydraulics/neutronics calculations has been performed. Comparison with experimental data will be the next step with High Temperature Test Reactor experimental results at nominal power
Investigation of coupling scheme for neutronic and thermal-hydraulic codes
International Nuclear Information System (INIS)
Wang Guoli; Yu Jianfeng; Pen Muzhang; Zhang Yuman.
1988-01-01
Recently, a number of coupled neutronics/thermal-hydraulics codes have been used in reaction design and safty analysis, which have been obtained by coupling previous neutronic and thermal-hydraulic codes. The different coupling schemes affect computer time and accuracy of calculation results. Numberical experiments of several different coupling schemes and some heuristic results are described
Current and anticipated uses of thermal-hydraulic codes in Germany
Energy Technology Data Exchange (ETDEWEB)
Teschendorff, V.; Sommer, F.; Depisch, F.
1997-07-01
In Germany, one third of the electrical power is generated by nuclear plants. ATHLET and S-RELAP5 are successfully applied for safety analyses of the existing PWR and BWR reactors and possible future reactors, e.g. EPR. Continuous development and assessment of thermal-hydraulic codes are necessary in order to meet present and future needs of licensing organizations, utilities, and vendors. Desired improvements include thermal-hydraulic models, multi-dimensional simulation, computational speed, interfaces to coupled codes, and code architecture. Real-time capability will be essential for application in full-scope simulators. Comprehensive code validation and quantification of uncertainties are prerequisites for future best-estimate analyses.
Current and anticipated uses of thermal-hydraulic codes in Germany
International Nuclear Information System (INIS)
Teschendorff, V.; Sommer, F.; Depisch, F.
1997-01-01
In Germany, one third of the electrical power is generated by nuclear plants. ATHLET and S-RELAP5 are successfully applied for safety analyses of the existing PWR and BWR reactors and possible future reactors, e.g. EPR. Continuous development and assessment of thermal-hydraulic codes are necessary in order to meet present and future needs of licensing organizations, utilities, and vendors. Desired improvements include thermal-hydraulic models, multi-dimensional simulation, computational speed, interfaces to coupled codes, and code architecture. Real-time capability will be essential for application in full-scope simulators. Comprehensive code validation and quantification of uncertainties are prerequisites for future best-estimate analyses
International Nuclear Information System (INIS)
Choi, Sun Rock; Back, Min Ho; Park, Won Seok; Kim, Sang Ji
2012-01-01
Since a fuel cladding failure is the most important parameter in a core thermal-hydraulic design, the conceptual design stage only involves fuel assemblies. However, although non-fuel assemblies such as control rod, reflector, and B4C generate a relatively smaller thermal power compared to fuel assemblies, they also require independent flow allocation to properly cool down each assembly. The thermal power in non-fuel assemblies is produced from both neutron and gamma energy, and thus the core thermal-hydraulic design including non-fuel assemblies should consider an energy redistribution by the gamma energy transport. To design non-fuel assemblies, the design-limiting parameters should be determined considering the thermal failure modes. While fuel assemblies set a limiting factor with cladding creep temperature to prevent a fission product ejection from the fuel rods, non-fuel assemblies restrict their outlet temperature to minimize thermally induced stress on the upper internal structure (UIS). This work employs a heat generation distribution reflecting both neutron and gamma transport. The whole core thermal-hydraulic design including fuel and non-fuel assemblies is then conducted using the SLTHEN (Steady-State LMR Thermal-Hydraulic Analysis Code Based on ENERGY Model) code. The other procedures follow from the previous conceptual design
Transitioning from interpretive to predictive in thermal hydraulic codes
International Nuclear Information System (INIS)
Mousseau, V.A.
2004-01-01
The current thermal hydraulic codes in use in the US, RELAP and TRAC, where originally written in the mid to late 1970's. At that time computers were slow, expensive, and had small memories. Because of these constraints, sacrifices had to be made, both in physics and numerical methods, which resulted in limitations on the accuracy of the solutions. Significant changes have occurred that induce very different requirements for the thermal hydraulic codes to be used for the future GEN-IV nuclear reactors. First, computers speed and memory grow at an exponential rate while the costs hold constant or decrease. Second, passive safety systems in modern designs stretch the length of relevant transients to many days. Finally, costs of experiments have grown very rapidly. Because of these new constraints, modern thermal hydraulic codes will be relied on for a significantly larger portion of bringing a nuclear reactor on line. Simulation codes will have to define in which part of state space experiments will be run. They will then have to be able to extend the small number of experiments to cover the large state space in which the reactors will operate. This data extrapolation mode will be referred to as 'predictive'. One of the keys to analyzing the accuracy of a simulation is to consider the entire domain being simulated. For example, in a reactor design where the containment is coupled to the reactor cooling system through radiative heat transfer, the accuracy of a transient includes the containment, the radiation heat transfer, the fluid flow in the cooling system, the thermal conduction in the solid, and the neutron transport in the reactor. All of this physics is coupled together in one nonlinear system through material properties, cross sections, heat transfer coefficients, and other mechanisms that exchange mass, momentum, and energy. Traditionally, these different physical domains, (containment, cooling system, nuclear fuel, etc.) have been solved in different
The fuel and channel thermal/mechanical behaviour code FACTAR 2.0 (LOCA)
International Nuclear Information System (INIS)
Westbye, C.J.; Mackinnon, J.C.; Gu, B.W.
1996-01-01
The computer code FACTAR 2.0 (LOCA) models the thermal and mechanical response of components within a single CANDU fuel channel under loss-of-coolant accident conditions. This code version is the successor to the FACTAR 1.x code series, and features many modelling enhancements over its predecessor. In particular, the thermal hydraulic treatment has been extended to model reverse and bi-directional coolant flow, and the axial variation in coolant flow rate. Thermal radiation is calculated by a detailed surface-to-surface model, and the ability to represent a greater range of geometries (including experimental configurations employed in code validation) has been implemented. Details of these new code treatments are described in this paper. (author)
International Nuclear Information System (INIS)
2001-01-01
An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for
Directory of Open Access Journals (Sweden)
Luca Evangelisti
2014-07-01
Full Text Available This study aims to highlight the importance of thermal inertia in buildings. Nowadays, it is possible to use energy analysis software to simulate the building energy performance. Considering Italian standards, these analyses are based on the UNI TS 11300 that defines the procedures for the national implementation of the UNI EN ISO 13790. These standards require an energy analysis under steady-state condition, underestimating the thermal inertia of the building. In order to understand the inertial behavior of walls, a cubic Test-Cell was modelled through the dynamic calculation code TRNSYS and three different wall types were tested. Different stratigraphies, characterized by the same thermal transmittance value, composed by massive elements and insulating layers in different order, were simulated. Through TRNSYS, it was possible to define maximum surface temperatures and to calculate thermal lag between maximum values, both external and internal. Moreover, the attenuation between external surface temperatures and internal ones during summer (July was calculated. Finally, the comparison between Test-Cell’s annual energy demands, performed by using a commercial code based on the Italian standard UNITS 11300 and the dynamic code, TRNSYS, was carried out.
International Nuclear Information System (INIS)
Singh, Tej; Kumar, Jainendra; Mazumdar, Tanay; Raina, V.K.
2013-01-01
Highlights: • A point reactor kinetics code coupled with thermal hydraulics of plate type fuel is developed. • This code is applicable for two phase flow of coolant. • Safety analysis of IAEA benchmark reactor core is carried out. • Results agree well with the results available in literature. - Abstract: A point reactor kinetics code SAC-RIT, acronym of Safety Analysis Code for Reactivity Initiated Transient, coupled with thermal hydraulics of two phase coolant flow for plate type fuel, is developed to calculate reactivity initiated transient analysis of nuclear research and test reactors. Point kinetics equations are solved by fourth order Runge Kutta method. Reactivity feedback effect is included into the code. Solution of kinetics equations gives neutronic power and it is then fed into a thermal hydraulic code where mass, momentum and thermal energy conservation equations are solved by explicit finite difference method to find out fuel, clad and coolant temperatures during transients. In this code, all possible flow regimes including laminar flow, transient flow and turbulent flow have been covered. Various heat transfer coefficients suitable for single liquid, sub-cooled boiling, saturation boiling, film boiling and single vapor phases are incorporated in the thermal hydraulic code
International Nuclear Information System (INIS)
Seubert, A.; Velkov, K.; Langenbuch, S.
2008-01-01
This paper describes the time-dependent 3D discrete ordinates transport code TORT-TD. Thermal-hydraulic feedback is considered by coupling TORT-TD with the thermal-hydraulics system code ATHLET. The coupled code TORT-TD/ATHLET allows 3D pin-by-pin analyses of transients in few energy groups and anisotropic scattering by solving the time-dependent transport equation using the unconditionally stable implicit method. The nuclear cross sections are interpolated between pre-calculated table values of fuel temperature, moderator density and boron concentration. For verification of the implementation, selected test cases have been calculated by TORT-TD/ATHLET. They include a control rod ejection transient in a small PWR fuel assembly arrangement and a local boron concentration change in a single PWR fuel assembly. In the latter, special attention has been paid to study the influence of the thermal-hydraulic feedback modelling in ATHLET. The results obtained for a control rod ejection accident in a PWR quarter core demonstrate the applicability of TORT-TD/ATHLET. (authors)
An improved UO2 thermal conductivity model in the ELESTRES computer code
International Nuclear Information System (INIS)
Chassie, G.G.; Tochaie, M.; Xu, Z.
2010-01-01
This paper describes the improved UO 2 thermal conductivity model for use in the ELESTRES (ELEment Simulation and sTRESses) computer code. The ELESTRES computer code models the thermal, mechanical and microstructural behaviour of a CANDU® fuel element under normal operating conditions. The main purpose of the code is to calculate fuel temperatures, fission gas release, internal gas pressure, fuel pellet deformation, and fuel sheath strains for fuel element design and assessment. It is also used to provide initial conditions for evaluating fuel behaviour during high temperature transients. The thermal conductivity of UO 2 fuel is one of the key parameters that affect ELESTRES calculations. The existing ELESTRES thermal conductivity model has been assessed and improved based on a large amount of thermal conductivity data from measurements of irradiated and un-irradiated UO 2 fuel with different densities. The UO 2 thermal conductivity data cover 90% to 99% theoretical density of UO 2 , temperature up to 3027 K, and burnup up to 1224 MW·h/kg U. The improved thermal conductivity model, which is recommended for a full implementation in the ELESTRES computer code, has reduced the ELESTRES code prediction biases of temperature, fission gas release, and fuel sheath strains when compared with the available experimental data. This improved thermal conductivity model has also been checked with a test version of ELESTRES over the full ranges of fuel temperature, fuel burnup, and fuel density expected in CANDU fuel. (author)
GAPCON-THERMAL-3 code description
International Nuclear Information System (INIS)
Lanning, D.D.; Mohr, C.L.; Panisko, F.E.; Stewart, K.B.
1978-01-01
GAPCON-3 is a computer program that predicts the thermal and mechanical behavior of an operating fuel rod during its normal lifetime. The code calculates temperatures, dimensions, stresses, and strains for the fuel and the cladding in both the radial and axial directions for each step of the user specified power history. The method of weighted residuals is for the steady state temperature calculation, and is combined with a finite difference approximation of the time derivative for transient conditions. The stress strain analysis employs an iterative axisymmetric finite element procedure that includes plasticity and creep for normal and pellet-clad mechanical interaction loads. GAPCON-3 can solve steady state and operational transient problems. Comparisons of GAPCON-3 predictions to both closed form analytical solutions and actual inpile instrumented fuel rod data have demonstrated the ability of the code to calculate fuel rod behavior. GAPCON-3 features a restart capability and an associated plot package unavailable in previous GAPCON series codes
GAPCON-THERMAL-3 code description
Energy Technology Data Exchange (ETDEWEB)
Lanning, D.D.; Mohr, C.L.; Panisko, F.E.; Stewart, K.B.
1978-01-01
GAPCON-3 is a computer program that predicts the thermal and mechanical behavior of an operating fuel rod during its normal lifetime. The code calculates temperatures, dimensions, stresses, and strains for the fuel and the cladding in both the radial and axial directions for each step of the user specified power history. The method of weighted residuals is for the steady state temperature calculation, and is combined with a finite difference approximation of the time derivative for transient conditions. The stress strain analysis employs an iterative axisymmetric finite element procedure that includes plasticity and creep for normal and pellet-clad mechanical interaction loads. GAPCON-3 can solve steady state and operational transient problems. Comparisons of GAPCON-3 predictions to both closed form analytical solutions and actual inpile instrumented fuel rod data have demonstrated the ability of the code to calculate fuel rod behavior. GAPCON-3 features a restart capability and an associated plot package unavailable in previous GAPCON series codes.
An assessment of the CORCON-MOD3 code. Part 1: Thermal-hydraulic calculations
International Nuclear Information System (INIS)
Strizhov, V.; Kanukova, V.; Vinogradova, T.; Askenov, E.; Nikulshin, V.
1996-09-01
This report deals with the subject of CORCON-Mod3 code validation (thermal-hydraulic modeling capability only) based on MCCI (molten core concrete interaction) experiments conducted under different programs in the past decade. Thermal-hydraulic calculations (i.e., concrete ablation, melt temperature, melt energy, concrete temperature, and condensible and non-condensible gas generation) were performed with the code, and compared with the data from 15 experiments, conducted at different scales using both simulant (metallic and oxidic) and prototypic melt materials, using different concrete types, and with and without an overlying water pool. Sensitivity studies were performed in a few cases involving, for example, heat transfer from melt to concrete, condensed phase chemistry, etc. Further, special analysis was performed using the ACE L8 experimental data to illustrate the differences between the experimental and the reactor conditions, and to demonstrate that with proper corrections made to the code, the calculated results were in better agreement with the experimental data. Generally, in the case of dry cavity and metallic melts, CORCON-Mod3 thermal-hydraulic calculations were in good agreement with the test data. For oxidic melts in a dry cavity, uncertainties in heat transfer models played an important role for two melt configurations--a stratified geometry with segregated metal and oxide layers, and a heterogeneous mixture. Some discrepancies in the gas release data were noted in a few cases
The 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.
International Nuclear Information System (INIS)
Buonomano, Annamaria; Palombo, Adolfo
2014-01-01
Highlights: • A new dynamic simulation code for building energy performance analysis is presented. • The thermal behavior of each building element is modeled by a thermal RC network. • The physical models implemented in the code are illustrated. • The code was validated by the BESTEST standard procedure. • We investigate residential buildings, offices and stores in different climates. - Abstract: A novel dynamic simulation model for the building envelope energy performance analysis is presented in this paper. This tool helps the investigation of many new building technologies to increase the system energy efficiency and it can be carried out for scientific research purposes. In addition to the yearly heating and cooling load and energy demand, the obtained output is the dynamic temperature profile of indoor air and surfaces and the dynamic profile of the thermal fluxes through the building elements. The presented simulation model is also validated through the BESTEST standard procedure. Several new case studies are developed for assessing, through the presented code, the energy performance of three different building envelopes with several different weather conditions. In particular, dwelling and commercial buildings are analysed. Light and heavyweight envelopes as well as different glazed surfaces areas have been used for every case study. With the achieved results interesting design and operating guidelines can be obtained. Such data have been also compared vs. those calculated by TRNSYS and EnergyPlus. The detected deviation of the obtained results vs. those of such standard tools are almost always lower than 10%
International Nuclear Information System (INIS)
Carlson, K.E.; Ransom, V.H.; Roth, P.A.
1987-03-01
The ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) code has been developed to perform transient simulation of the thermal hydraulic systems that may be found in fusion reactors, space reactors, and other advanced systems. As an assessment of current capability the code was applied to a number of physical problems, both conceptual and actual experiments. Results indicate that the numerical solution to the basic conservation equations is technically sound, and that generally good agreement can be obtained when modeling relevant hydrodynamic experiments. The assessment also demonstrates basic fusion system modeling capability and verifies compatibility of the code with both CDC and CRAY mainframes. Areas where improvements could be made include constitutive modeling, which describes the interfacial exchange term. 13 refs., 84 figs
Visual and intelligent transients and accidents analyzer based on thermal-hydraulic system code
International Nuclear Information System (INIS)
Meng Lin; Rui Hu; Yun Su; Ronghua Zhang; Yanhua Yang
2005-01-01
Full text of publication follows: Many thermal-hydraulic system codes were developed in the past twenty years, such as RELAP5, RETRAN, ATHLET, etc. Because of their general and advanced features in thermal-hydraulic computation, they are widely used in the world to analyze transients and accidents. But there are following disadvantages for most of these original thermal-hydraulic system codes. Firstly, because models are built through input decks, so the input files are complex and non-figurative, and the style of input decks is various for different users and models. Secondly, results are shown in off-line data file form. It is not convenient for analysts who may pay more attention to dynamic parameters trend and changing. Thirdly, there are few interfaces with other program in these original thermal-hydraulic system codes. This restricts the codes expanding. The subject of this paper is to develop a powerful analyzer based on these thermal-hydraulic system codes to analyze transients and accidents more simply, accurately and fleetly. Firstly, modeling is visual and intelligent. Users build the thermalhydraulic system model using component objects according to their needs, and it is not necessary for them to face bald input decks. The style of input decks created automatically by the analyzer is unified and can be accepted easily by other people. Secondly, parameters concerned by analyst can be dynamically communicated to show or even change. Thirdly, the analyzer provide interface with other programs for the thermal-hydraulic system code. Thus parallel computation between thermal-hydraulic system code and other programs become possible. In conclusion, through visual and intelligent method, the analyzer based on general and advanced thermal-hydraulic system codes can be used to analysis transients and accidents more effectively. The main purpose of this paper is to present developmental activities, assessment and application results of the visual and intelligent
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)
Country Report on Building Energy Codes in China
Energy Technology Data Exchange (ETDEWEB)
Shui, Bin; Evans, Meredydd; Lin, H.; Jiang, Wei; Liu, Bing; Song, Bo; Somasundaram, Sriram
2009-04-15
This report is part of a series of reports on building energy efficiency codes in countries associated with the Asian Pacific Partnership (APP) - Australia, South Korea, Japan, China, India, and the United States of America (U.S.). This reports gives an overview of the development of building energy codes in China, including national energy policies related to building energy codes, history of building energy codes, recent national projects and activities to promote building energy codes. The report also provides a review of current building energy codes (such as building envelope and HVAC) for commercial and residential buildings in China.
Country Report on Building Energy Codes in Japan
Energy Technology Data Exchange (ETDEWEB)
Evans, Meredydd; Shui, Bin; Takagi, T.
2009-04-15
This report is part of a series of reports on building energy efficiency codes in countries associated with the Asian Pacific Partnership (APP) - Australia, South Korea, Japan, China, India, and the United States of America (U.S.). This reports gives an overview of the development of building energy codes in Japan, including national energy policies related to building energy codes, history of building energy codes, recent national projects and activities to promote building energy codes. The report also provides a review of current building energy codes (such as building envelope, HVAC, and lighting) for commercial and residential buildings in Japan.
Country Report on Building Energy Codes in Australia
Energy Technology Data Exchange (ETDEWEB)
Shui, Bin; Evans, Meredydd; Somasundaram, Sriram
2009-04-02
This report is part of a series of reports on building energy efficiency codes in countries associated with the Asian Pacific Partnership (APP) - Australia, South Korea, Japan, China, India, and the United States of America (U.S.). This reports gives an overview of the development of building energy codes in Australia, including national energy policies related to building energy codes, history of building energy codes, recent national projects and activities to promote building energy codes. The report also provides a review of current building energy codes (such as building envelope, HVAC, and lighting) for commercial and residential buildings in Australia.
Country Report on Building Energy Codes in Canada
Energy Technology Data Exchange (ETDEWEB)
Shui, Bin; Evans, Meredydd
2009-04-06
This report is part of a series of reports on building energy efficiency codes in countries associated with the Asian Pacific Partnership (APP) - Australia, South Korea, Japan, China, India, and the United States of America . This reports gives an overview of the development of building energy codes in Canada, including national energy policies related to building energy codes, history of building energy codes, recent national projects and activities to promote building energy codes. The report also provides a review of current building energy codes (such as building envelope, HVAC, lighting, and water heating) for commercial and residential buildings in Canada.
Mesh generation and energy group condensation studies for the jaguar deterministic transport code
International Nuclear Information System (INIS)
Kennedy, R. A.; Watson, A. M.; Iwueke, C. I.; Edwards, E. J.
2012-01-01
The deterministic transport code Jaguar is introduced, and the modeling process for Jaguar is demonstrated using a two-dimensional assembly model of the Hoogenboom-Martin Performance Benchmark Problem. This single assembly model is being used to test and analyze optimal modeling methodologies and techniques for Jaguar. This paper focuses on spatial mesh generation and energy condensation techniques. In this summary, the models and processes are defined as well as thermal flux solution comparisons with the Monte Carlo code MC21. (authors)
Mesh generation and energy group condensation studies for the jaguar deterministic transport code
Energy Technology Data Exchange (ETDEWEB)
Kennedy, R. A.; Watson, A. M.; Iwueke, C. I.; Edwards, E. J. [Knolls Atomic Power Laboratory, Bechtel Marine Propulsion Corporation, P.O. Box 1072, Schenectady, NY 12301-1072 (United States)
2012-07-01
The deterministic transport code Jaguar is introduced, and the modeling process for Jaguar is demonstrated using a two-dimensional assembly model of the Hoogenboom-Martin Performance Benchmark Problem. This single assembly model is being used to test and analyze optimal modeling methodologies and techniques for Jaguar. This paper focuses on spatial mesh generation and energy condensation techniques. In this summary, the models and processes are defined as well as thermal flux solution comparisons with the Monte Carlo code MC21. (authors)
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.
CEDNBR: a computer code for transient thermal margin analysis of a reactor core
International Nuclear Information System (INIS)
Shesler, A.T.; Lehmann, C.R.
1976-09-01
The report describes the CEDNBR computer code. This code was developed for the transient thermal analysis of a pressurized water reactor core or a critical heat flux test. Included are the code structure, conservation equations, and correlations utilized by CEDNBR. The methods of modelling a reactor core and hot channel and a CHF test are presented. Comparisons of CEDNBR calculations are made with both empirical pressure loss data and simulated loss of flow test data. The code solves the one-dimensional conservation of mass, energy, and momentum equations and the equation of state for the fluid for either steady-state or transient conditions. Tabular time dependent functions of inlet temperatures, pressure, mass velocity, axial heat flux distributions, normalized heat flux, radial peaking factors, and incremental mixing factors are required input to the code. Transient effects are included in the calculation of enthalpy rise and fluid properties. The Departure from Nucleate Boiling Ratio (DNBR) is calculated by applying a Critical Heat Flux (CHF) correlation to the computed local fluid properties. A code user's guide is provided for preparing input to the code. In addition, descriptions of the sub-routines used by CEDNBR are given
CASKETSS: a computer code system for thermal and structural analysis of nuclear fuel shipping casks
International Nuclear Information System (INIS)
Ikushima, Takeshi
1989-02-01
A computer program CASKETSS has been developed for the purpose of thermal and structural analysis of nuclear fuel shipping casks. CASKETSS measn a modular code system for CASK Evaluation code system Thermal and Structural Safety. Main features of CASKETSS are as follow; (1) Thermal and structural analysis computer programs for one-, two-, three-dimensional geometries are contained in the code system. (2) Some of the computer programs in the code system has been programmed to provide near optimal speed on vector processing computers. (3) Data libralies fro thermal and structural analysis are provided in the code system. (4) Input data generator is provided in the code system. (5) Graphic computer program is provided in the code system. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)
Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures
Directory of Open Access Journals (Sweden)
Alessandro Petruzzi
2008-01-01
Full Text Available In the last four decades, large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Whereas the first system codes, developed at the beginning of the 1970s, utilized the homogenous equilibrium model with three balance equations to describe the two-phase flow, nowadays the more advanced system codes are based on the so-called “two-fluid model” with separation of the water and vapor phases, resulting in systems with at least six balance equations. The wide experimental campaign, constituted by the integral and separate effect tests, conducted under the umbrella of the OECD/CSNI was at the basis of the development and validation of the thermal-hydraulic system codes by which they have reached the present high degree of maturity. However, notwithstanding the huge amounts of financial and human resources invested, the results predicted by the code are still affected by errors whose origins can be attributed to several reasons as model deficiencies, approximations in the numerical solution, nodalization effects, and imperfect knowledge of boundary and initial conditions. In this context, the existence of qualified procedures for a consistent application of qualified thermal-hydraulic system code is necessary and implies the drawing up of specific criteria through which the code-user, the nodalization, and finally the transient results are qualified.
Energy Technology Data Exchange (ETDEWEB)
Watanabe, Shouichi; Yamane, Yuichi; Miyoshi, Yoshinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2003-03-01
Since exact information is not always acquired in the criticality accident of fuel solution, parametric survey calculations are required for grasping behaviors of the thermal-hydraulics. On the other hand, the practical methods of the calculation with can reduce the computation time with allowable accuracy will be also required, since the conventional method takes a long calculation time. In order to fulfill the requirement, a two-dimensional (R-Z) nuclear-kinetics analysis code considering thermal-hydraulic based on the multi-region kinetic equations with one-group neutron energy was created by incorporating with the thermal-hydraulics analysis code PHOENICS for all-purpose use the computation time of the code was shortened by separating time mesh intervals of the nuclear- and heat-calculations from that of the hydraulics calculation, and by regulating automatically the time mesh intervals in proportion to power change rate. A series of analysis were performed for the natural-cooling characteristic test using TRACY in which the power changed slowly for 5 hours after the transient power resulting from the reactivity insertion of a 0.5 dollar. It was found that the code system was able to calculate within the limit of practical time, and acquired the prospect of reproducing the experimental values considerably for the power and temperature change. (author)
Building Energy Codes: Policy Overview and Good Practices
Energy Technology Data Exchange (ETDEWEB)
Cox, Sadie [National Renewable Energy Lab. (NREL), Golden, CO (United States)
2016-02-19
Globally, 32% of total final energy consumption is attributed to the building sector. To reduce energy consumption, energy codes set minimum energy efficiency standards for the building sector. With effective implementation, building energy codes can support energy cost savings and complementary benefits associated with electricity reliability, air quality improvement, greenhouse gas emission reduction, increased comfort, and economic and social development. This policy brief seeks to support building code policymakers and implementers in designing effective building code programs.
76 FR 64931 - Building Energy Codes Cost Analysis
2011-10-19
...-0046] Building Energy Codes Cost Analysis AGENCY: Office of Energy Efficiency and Renewable Energy... reopening of the time period for submitting comments on the request for information on Building Energy Codes... the request for information on Building Energy Code Cost Analysis and provide docket number EERE-2011...
Energy Technology Data Exchange (ETDEWEB)
Ebert, D.
1997-07-01
This is a report on the CSNI Workshop on Transient Thermal-Hydraulic and Neutronic Codes Requirements held at Annapolis, Maryland, USA November 5-8, 1996. This experts` meeting consisted of 140 participants from 21 countries; 65 invited papers were presented. The meeting was divided into five areas: (1) current and prospective plans of thermal hydraulic codes development; (2) current and anticipated uses of thermal-hydraulic codes; (3) advances in modeling of thermal-hydraulic phenomena and associated additional experimental needs; (4) numerical methods in multi-phase flows; and (5) programming language, code architectures and user interfaces. The workshop consensus identified the following important action items to be addressed by the international community in order to maintain and improve the calculational capability: (a) preserve current code expertise and institutional memory, (b) preserve the ability to use the existing investment in plant transient analysis codes, (c) maintain essential experimental capabilities, (d) develop advanced measurement capabilities to support future code validation work, (e) integrate existing analytical capabilities so as to improve performance and reduce operating costs, (f) exploit the proven advances in code architecture, numerics, graphical user interfaces, and modularization in order to improve code performance and scrutibility, and (g) more effectively utilize user experience in modifying and improving the codes.
International Nuclear Information System (INIS)
Ebert, D.
1997-07-01
This is a report on the CSNI Workshop on Transient Thermal-Hydraulic and Neutronic Codes Requirements held at Annapolis, Maryland, USA November 5-8, 1996. This experts' meeting consisted of 140 participants from 21 countries; 65 invited papers were presented. The meeting was divided into five areas: (1) current and prospective plans of thermal hydraulic codes development; (2) current and anticipated uses of thermal-hydraulic codes; (3) advances in modeling of thermal-hydraulic phenomena and associated additional experimental needs; (4) numerical methods in multi-phase flows; and (5) programming language, code architectures and user interfaces. The workshop consensus identified the following important action items to be addressed by the international community in order to maintain and improve the calculational capability: (a) preserve current code expertise and institutional memory, (b) preserve the ability to use the existing investment in plant transient analysis codes, (c) maintain essential experimental capabilities, (d) develop advanced measurement capabilities to support future code validation work, (e) integrate existing analytical capabilities so as to improve performance and reduce operating costs, (f) exploit the proven advances in code architecture, numerics, graphical user interfaces, and modularization in order to improve code performance and scrutibility, and (g) more effectively utilize user experience in modifying and improving the codes
3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors
Energy Technology Data Exchange (ETDEWEB)
Langenbuch, S.; Velkov, K. [GRS, Garching (Germany); Lizorkin, M. [Kurchatov-Institute, Moscow (Russian Federation)] [and others
1997-07-01
This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.
FLICA-4 (version 1). A computer code for three dimensional thermal analysis of nuclear reactor cores
International Nuclear Information System (INIS)
Raymond, P.; Allaire, G.; Boudsocq, G.; Caruge, D.; Gramont, T. de; Toumi, I.
1995-01-01
FLICA-4 is a thermal-hydraulic computer code, developed at the French Atomic Energy Commission (CEA) for three-dimensional steady-state or transient two-phase flow, and aimed at design and safety thermal analysis of nuclear reactor cores. It is available for various UNIX workstations and CRAY computers under UNICOS.It is based on four balance equations which include three balance equations for the mixture and a mass balance equation for the less concentrated phase which allows for the calculation of non equilibrium flows such as sub-cooled boiling and superheated steam. A drift velocity model takes into account the velocity unbalance between phases. The equations are solved using a finite volume numerical scheme. Typical running time, specific features (coupling with other codes) and auxiliary programs are presented. 1 tab., 9 refs
Aquifer Thermal Energy Storage for Seasonal Thermal Energy Balance
Rostampour, Vahab; Bloemendal, Martin; Keviczky, Tamas
2017-04-01
Aquifer Thermal Energy Storage (ATES) systems allow storing large quantities of thermal energy in subsurface aquifers enabling significant energy savings and greenhouse gas reductions. This is achieved by injection and extraction of water into and from saturated underground aquifers, simultaneously. An ATES system consists of two wells and operates in a seasonal mode. One well is used for the storage of cold water, the other one for the storage of heat. In warm seasons, cold water is extracted from the cold well to provide cooling to a building. The temperature of the extracted cold water increases as it passes through the building climate control systems and then gets simultaneously, injected back into the warm well. This procedure is reversed during cold seasons where the flow direction is reversed such that the warmer water is extracted from the warm well to provide heating to a building. From the perspective of building climate comfort systems, an ATES system is considered as a seasonal storage system that can be a heat source or sink, or as a storage for thermal energy. This leads to an interesting and challenging optimal control problem of the building climate comfort system that can be used to develop a seasonal-based energy management strategy. In [1] we develop a control-oriented model to predict thermal energy balance in a building climate control system integrated with ATES. Such a model however cannot cope with off-nominal but realistic situations such as when the wells are completely depleted, or the start-up phase of newly installed wells, etc., leading to direct usage of aquifer ambient temperature. Building upon our previous work in [1], we here extend the mathematical model for ATES system to handle the above mentioned more realistic situations. Using our improved models, one can more precisely predict system behavior and apply optimal control strategies to manage the building climate comfort along with energy savings and greenhouse gas reductions
Baumeister, Joseph F.
1994-01-01
A non-flowing, electrically heated test rig was developed to verify computer codes that calculate radiant energy propagation from nozzle geometries that represent aircraft propulsion nozzle systems. Since there are a variety of analysis tools used to evaluate thermal radiation propagation from partially enclosed nozzle surfaces, an experimental benchmark test case was developed for code comparison. This paper briefly describes the nozzle test rig and the developed analytical nozzle geometry used to compare the experimental and predicted thermal radiation results. A major objective of this effort was to make available the experimental results and the analytical model in a format to facilitate conversion to existing computer code formats. For code validation purposes this nozzle geometry represents one validation case for one set of analysis conditions. Since each computer code has advantages and disadvantages based on scope, requirements, and desired accuracy, the usefulness of this single nozzle baseline validation case can be limited for some code comparisons.
Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels
Energy Technology Data Exchange (ETDEWEB)
Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro, E-mail: duvan.castellanos@ufabc.edu.br, E-mail: joao.moreira@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: pedro.rossi@ufabc.edu.br, E-mail: pedro.carajilescov10@gmail.com [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil). Centro de Engenharias, Modelagem e Ciências Sociais Aplicadas
2017-07-01
To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat
Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels
International Nuclear Information System (INIS)
Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro
2017-01-01
To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat
More Efficient Solar Thermal-Energy Receiver
Dustin, M. O.
1987-01-01
Thermal stresses and reradiation reduced. Improved design for solar thermal-energy receiver overcomes three major deficiencies of solar dynamic receivers described in literature. Concentrator and receiver part of solar-thermal-energy system. Receiver divided into radiation section and storage section. Concentrated solar radiation falls on boiling ends of heat pipes, which transmit heat to thermal-energy-storage medium. Receiver used in number of applications to produce thermal energy directly for use or to store thermal energy for subsequent use in heat engine.
Thermal energy systems design and analysis
Penoncello, Steven G
2015-01-01
IntroductionThermal Energy Systems Design and AnalysisSoftwareThermal Energy System TopicsUnits and Unit SystemsThermophysical PropertiesEngineering DesignEngineering EconomicsIntroductionCommon Engineering Economics NomenclatureEconomic Analysis Tool: The Cash Flow DiagramTime Value of MoneyTime Value of Money ExamplesUsing Software to Calculate Interest FactorsEconomic Decision MakingDepreciation and TaxesProblemsAnalysis of Thermal Energy SystemsIntroductionNomenclatureThermophysical Properties of SubstancesSuggested Thermal Energy Systems Analysis ProcedureConserved and Balanced QuantitiesConservation of MassConservation of Energy (The First Law of Thermodynamics)Entropy Balance (The Second Law of Thermodynamics)Exergy Balance: The Combined LawEnergy and Exergy Analysis of Thermal Energy CyclesDetailed Analysis of Thermal Energy CyclesProblemsFluid Transport in Thermal Energy SystemsIntroductionPiping and Tubing StandardsFluid Flow FundamentalsValves and FittingsDesign and Analysis of Pipe NetworksEconomi...
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.
Waste energy harvesting mechanical and thermal energies
Ling Bing, Kong; Hng, Huey Hoon; Boey, Freddy; Zhang, Tianshu
2014-01-01
Waste Energy Harvesting overviews the latest progress in waste energy harvesting technologies, with specific focusing on waste thermal mechanical energies. Thermal energy harvesting technologies include thermoelectric effect, storage through phase change materials and pyroelectric effect. Waste mechanical energy harvesting technologies include piezoelectric (ferroelectric) effect with ferroelectric materials and nanogenerators. The book aims to strengthen the syllabus in energy, materials and physics and is well suitable for students and professionals in the fields.
High energy particle transport code NMTC/JAM
International Nuclear Information System (INIS)
Niita, K.; Takada, H.; Meigo, S.; Ikeda, Y.
2001-01-01
We have developed a high energy particle transport code NMTC/JAM, which is an upgrade version of NMTC/JAERI97. The available energy range of NMTC/JAM is, in principle, extended to 200 GeV for nucleons and mesons including the high energy nuclear reaction code JAM for the intra-nuclear cascade part. We compare the calculations by NMTC/JAM code with the experimental data of thin and thick targets for proton induced reactions up to several 10 GeV. The results of NMTC/JAM code show excellent agreement with the experimental data. From these code validation, it is concluded that NMTC/JAM is reliable in neutronics optimization study of the high intense spallation neutron utilization facility. (author)
High Energy Transport Code HETC
International Nuclear Information System (INIS)
Gabriel, T.A.
1985-09-01
The physics contained in the High Energy Transport Code (HETC), in particular the collision models, are discussed. An application using HETC as part of the CALOR code system is also given. 19 refs., 5 figs., 3 tabs
Preliminary development of thermal nuclear cell homogenization code
International Nuclear Information System (INIS)
Su'ud, Z.; Shafii, M. A.; Yudha, S. P.; Waris, A.; Rijal, K.
2012-01-01
Nuclear fuel cell homogenization for thermal reactors usually include three main parts, i.e., fast energy resonance part which usually adopt narrow resonance approximation to treat the resonance, low (intermediate) energy region in which the resonance can not be treated accurately using NR approximation and therefore we should use intermediate resonance treatment, and thermal energy region (very low) in which the effect of thermal must be treated properly. In n this study the application of the intermediate resonance approximation treatment for low energy nuclear resonance is discussed. The method is iterative based. As a sample the method is applied in U-235 low lying resonance and the result is presented and discussed.
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.
International Nuclear Information System (INIS)
Lightston, M.F.; Rock, R.
1996-01-01
This paper presents the results of a study of the thermal mixing of single-phase coolant in 28-element CANDU fuel bundles under steady-state conditions. The study, which is based on simulations performed using the ASSERT-PV thermal hydraulic code, consists of two main parts. In the first part the various physical mechanisms that contribute to coolant mixing are identified and their impact is isolated via ASSERT-PV simulations. The second part is concerned with development of a preliminary model suitable for use in the fuel and fuel channel code FACTAR to predict the thermal mixing that occurs between flow annuli. (author)
Roadmap for the Future of Commercial Energy Codes
Energy Technology Data Exchange (ETDEWEB)
Rosenberg, Michael I. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hart, Philip R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Zhang, Jian [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Athalye, Rahul A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
2015-01-01
Building energy codes have significantly increased building efficiency over the last 38 years, since the first national energy code was published in 1975. The most commonly used path in energy codes, the prescriptive path, appears to be reaching a point of diminishing returns. The current focus on prescriptive codes has limitations including significant variation in actual energy performance depending on which prescriptive options are chosen, a lack of flexibility for designers and developers, the inability to handle optimization that is specific to building type and use, the inability to account for project-specific energy costs, and the lack of follow-through or accountability after a certificate of occupancy is granted. It is likely that an approach that considers the building as an integrated system will be necessary to achieve the next real gains in building efficiency. This report provides a high-level review of different formats for commercial building energy codes, including prescriptive, prescriptive packages, capacity constrained, outcome based, and predictive performance approaches. This report also explores a next generation commercial energy code approach that places a greater emphasis on performance-based criteria.
Nuclear-thermal-coupled optimization code for the fusion breeding blanket conceptual design
International Nuclear Information System (INIS)
Li, Jia; Jiang, Kecheng; Zhang, Xiaokang; Nie, Xingchen; Zhu, Qinjun; Liu, Songlin
2016-01-01
Highlights: • A nuclear-thermal-coupled predesign code has been developed for optimizing the radial build arrangement of fusion breeding blanket. • Coupling module aims at speeding up the efficiency of design progress by coupling the neutronics calculation code with the thermal-hydraulic analysis code. • Radial build optimization algorithm aims at optimal arrangement of breeding blanket considering one or multiple specified objectives subject to the design criteria such as material temperature limit and available TBR. - Abstract: Fusion breeding blanket as one of the key in-vessel components performs the functions of breeding the tritium, removing the nuclear heat and heat flux from plasma chamber as well as acting as part of shielding system. The radial build design which determines the arrangement of function zones and material properties on the radial direction is the basis of the detailed design of fusion breeding blanket. For facilitating the radial build design, this study aims for developing a pre-design code to optimize the radial build of blanket with considering the performance of nuclear and thermal-hydraulic simultaneously. Two main features of this code are: (1) Coupling of the neutronics analysis with the thermal-hydraulic analysis to speed up the analysis progress; (2) preliminary optimization algorithm using one or multiple specified objectives subject to the design criteria in the form of constrains imposed on design variables and performance parameters within the possible engineering ranges. This pre-design code has been applied to the conceptual design of water-cooled ceramic breeding blanket in project of China fusion engineering testing reactor (CFETR).
Nuclear-thermal-coupled optimization code for the fusion breeding blanket conceptual design
Energy Technology Data Exchange (ETDEWEB)
Li, Jia, E-mail: lijia@ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027, Anhui (China); Jiang, Kecheng; Zhang, Xiaokang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, Anhui (China); Nie, Xingchen [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027, Anhui (China); Zhu, Qinjun; Liu, Songlin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, Anhui (China)
2016-12-15
Highlights: • A nuclear-thermal-coupled predesign code has been developed for optimizing the radial build arrangement of fusion breeding blanket. • Coupling module aims at speeding up the efficiency of design progress by coupling the neutronics calculation code with the thermal-hydraulic analysis code. • Radial build optimization algorithm aims at optimal arrangement of breeding blanket considering one or multiple specified objectives subject to the design criteria such as material temperature limit and available TBR. - Abstract: Fusion breeding blanket as one of the key in-vessel components performs the functions of breeding the tritium, removing the nuclear heat and heat flux from plasma chamber as well as acting as part of shielding system. The radial build design which determines the arrangement of function zones and material properties on the radial direction is the basis of the detailed design of fusion breeding blanket. For facilitating the radial build design, this study aims for developing a pre-design code to optimize the radial build of blanket with considering the performance of nuclear and thermal-hydraulic simultaneously. Two main features of this code are: (1) Coupling of the neutronics analysis with the thermal-hydraulic analysis to speed up the analysis progress; (2) preliminary optimization algorithm using one or multiple specified objectives subject to the design criteria in the form of constrains imposed on design variables and performance parameters within the possible engineering ranges. This pre-design code has been applied to the conceptual design of water-cooled ceramic breeding blanket in project of China fusion engineering testing reactor (CFETR).
TRIO-EF a general thermal hydraulics computer code applied to the Avlis process
International Nuclear Information System (INIS)
Magnaud, J.P.; Claveau, M.; Coulon, N.; Yala, P.; Guilbaud, D.; Mejane, A.
1993-01-01
TRIO(EF is a general purpose Fluid Mechanics 3D Finite Element Code. The system capabilities cover areas such as steady state or transient, laminar or turbulent, isothermal or temperature dependent fluid flows; it is applicable to the study of coupled thermo-fluid problems involving heat conduction and possibly radiative heat transfer. It has been used to study the thermal behaviour of the AVLIS process separation module. In this process, a linear electron beam impinges the free surface of a uranium ingot, generating a two dimensional curtain emission of vapour from a water-cooled crucible. The energy transferred to the metal causes its partial melting, forming a pool where strong convective motion increases heat transfer towards the crucible. In the upper part of the Separation Module, the internal structures are devoted to two main functions: vapor containment and reflux, irradiation and physical separation. They are subjected to very high temperature levels and heat transfer occurs mainly by radiation. Moreover, special attention has to be paid to electron backscattering. These two major points have been simulated numerically with TRIO-EF and the paper presents and comments the results of such a computation, for each of them. After a brief overview of the computer code, two examples of the TRIO-EF capabilities are given: a crucible thermal hydraulics model, a thermal analysis of the internal structures
Energy Technology Data Exchange (ETDEWEB)
Marguet, S.D. [Electricite de France (EDF), 92 - Clamart (France)
1997-12-31
Among all parameters, the fuel temperature has a significant influence on the reactivity of the core, because of the Doppler effect on cross-sections. Most neutronic codes use a straightforward method to calculate an average fuel temperature used in their specific feed-back models. For instance, EDF`s neutronic code COCCINELLE uses the Rowland`s formula using the temperatures of the center and the surface of the pellet. COCCINELLE is coupled to the 3D thermal-hydraulic code THYC with calculates TDoppler with is standard thermal model. In order to improve the accuracy of such calculations, we have developed the coupling of our two latest codes in thermal-hydraulics (THYC) and thermo-mechanics (CYRANO3). THYC calculates two-phase flows in pipes or rod bundles and is used for transient calculations such as steam-line break, boron dilution accidents, DNB predictions, steam generator and condenser studies. CYRANO3 calculates most of the phenomena that take place in the fuel such as: 1) heat transfer induced by nuclear power; 2) thermal expansion of the fuel and the cladding; 3) release of gaseous fission`s products; 4) mechanical interaction between the pellet and the cladding. These two codes are now qualified in their own field and the coupling, using Parallel Virtual Machine (PVM) libraries customized in an home-made-easy-to-use package called CALCIUM, has been validated on `low` configurations (no thermal expansion, constant thermal characteristics) and used on accidental transients such as rod ejection and loss of coolant accident. (K.A.) 7 refs.
International Nuclear Information System (INIS)
Marguet, S.D.
1997-01-01
Among all parameters, the fuel temperature has a significant influence on the reactivity of the core, because of the Doppler effect on cross-sections. Most neutronic codes use a straightforward method to calculate an average fuel temperature used in their specific feed-back models. For instance, EDF's neutronic code COCCINELLE uses the Rowland's formula using the temperatures of the center and the surface of the pellet. COCCINELLE is coupled to the 3D thermal-hydraulic code THYC with calculates TDoppler with is standard thermal model. In order to improve the accuracy of such calculations, we have developed the coupling of our two latest codes in thermal-hydraulics (THYC) and thermo-mechanics (CYRANO3). THYC calculates two-phase flows in pipes or rod bundles and is used for transient calculations such as steam-line break, boron dilution accidents, DNB predictions, steam generator and condenser studies. CYRANO3 calculates most of the phenomena that take place in the fuel such as: 1) heat transfer induced by nuclear power; 2) thermal expansion of the fuel and the cladding; 3) release of gaseous fission's products; 4) mechanical interaction between the pellet and the cladding. These two codes are now qualified in their own field and the coupling, using Parallel Virtual Machine (PVM) libraries customized in an home-made-easy-to-use package called CALCIUM, has been validated on 'low' configurations (no thermal expansion, constant thermal characteristics) and used on accidental transients such as rod ejection and loss of coolant accident. (K.A.)
Energy Technology Data Exchange (ETDEWEB)
Suh, Kune Yull; Yoon, Sang Hyuk; Noh, Sang Woo; Lee, Il Suk [Seoul National University, Seoul (Korea)
2002-03-01
This study is concerned with developing a multidimensional flow model required for the system analysis code MARS to more mechanistically simulate a variety of thermal hydraulic phenomena in the nuclear stem supply system. The capability of the MARS code as a thermal hydraulic analysis tool for optimized system design can be expanded by improving the current calculational methods and adding new models. In this study the relevant literature was surveyed on the multidimensional flow models that may potentially be applied to the multidimensional analysis code. Research items were critically reviewed and suggested to better predict the multidimensional thermal hydraulic behavior and to identify test requirements. A small-scale preliminary test was performed in the downcomer formed by two vertical plates to analyze multidimensional flow pattern in a simple geometry. The experimental result may be applied to the code for analysis of the fluid impingement to the reactor downcomer wall. Also, data were collected to find out the controlling parameters for the one-dimensional and multidimensional flow behavior. 22 refs., 40 figs., 7 tabs. (Author)
Application of thermal-hydraulic codes in the nuclear sector
International Nuclear Information System (INIS)
Queral, C.; Coriso, M.; Garcia Sedano, P. J.; Ruiz, J. A.; Posada, J. M.; Jimenez Varas, G.; Sol, I.; Herranz, L. E.
2011-01-01
Use of thermal-hydraulic codes is extended all over many different aspects of nuclear engineering. This article groups and briefly describes the main features of some of the well known codes as an introduction to their recent applications in the Spain nuclear sector. the broad range and quality of applications highlight the maturity achieved both in industry and research organizations and universities within the Spanish nuclear sector. (Author)
Energy Efficiency Program Administrators and Building Energy Codes
Explore how energy efficiency program administrators have helped advance building energy codes at federal, state, and local levels—using technical, institutional, financial, and other resources—and discusses potential next steps.
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
Computer code for the thermal-hydraulic analysis of ITU TRIGA Mark-II reactor
International Nuclear Information System (INIS)
Ustun, G.; Durmayaz, A.
2002-01-01
Istanbul Technical University (ITU) TRIGA Mark-II reactor core consists of ninety vertical cylindrical elements located in five rings. Sixty-nine of them are fuel elements. The reactor is operated and cooled with natural convection by pool water, which is also cooled and purified in external coolant circuits by forced convection. This characteristic leads to consider both the natural and forced convection heat transfer in a 'porous-medium analysis'. The safety analysis of the reactor requires a thermal-hydraulic model of the reactor to determine the thermal-hydraulic parameters in each mode of operation. In this study, a computer code cooled TRIGA-PM (TRIGA - Porous Medium) for the thermal-hydraulic analysis of ITU is considered. TRIGA Mark-II reactor code has been developed to obtain velocity, pressure and temperature distributions in the reactor pool as a function of core design parameters and pool configuration. The code is a transient, thermal-hydraulic code and requires geometric and physical modelling parameters. In the model, although the reactor is considered as only porous medium, the other part of the reactor pool is considered partly as continuum and partly as porous medium. COMMIX-1C code is used for the benchmark purpose of TRIGA-PM code. For the normal operating conditions of the reactor, estimations of TRIGA-PM are in good agreement with those of COMMIX-1C. After some more improvements, this code will be employed for the estimation of LOCA scenario, which can not be analyses by COMMIX-1C and the other multi-purpose codes, considering a break at one of the beam tubes of the reactor
Study of Aquifer Thermal Energy Storage
Okuyama, Masaaki; Umemiya, Hiromichi; Shibuya, Ikuko; Haga, Eiji
Yamagata University 'Aquifer Thermal Energy Storage (ATES)' is the experimental system which has been running since 1982. From the results for along terms of experiments, we obtain many important knowledge. This paper presents the accomplishments for 16 years and the characteristics of thermal energy storage in thermal energy storage well. The conclusions show as follows. 1)In recent years, the thermal recovery factor of warm energy storage well becomes almost constant at about 60%. 2) The thermal recovery factor of cool energy storage well increases gradually and becomes at about 15%. 3) Since the ferric colloidal dam is formed in aquifer, thermal recovery factor increase year after year. 4) Back wash can remove clogging for ferric colloidal dam. 5) The apparent thermal diffusivity decrease gradually due to ferric colloidal dam.
Energy Technology Data Exchange (ETDEWEB)
Arai, Kenji; Ebata, Shigeo [Toshiba Corp., Yokohama (Japan)
1997-07-01
This paper summarizes the current and anticipated use of the thermal-hydraulic and neutronic codes for the BWR transient and accident analyses in Japan. The codes may be categorized into the licensing codes and the best estimate codes for the BWR transient and accident analyses. Most of the licensing codes have been originally developed by General Electric. Some codes have been updated based on the technical knowledge obtained in the thermal hydraulic study in Japan, and according to the BWR design changes. The best estimates codes have been used to support the licensing calculations and to obtain the phenomenological understanding of the thermal hydraulic phenomena during a BWR transient or accident. The best estimate codes can be also applied to a design study for a next generation BWR to which the current licensing model may not be directly applied. In order to rationalize the margin included in the current BWR design and develop a next generation reactor with appropriate design margin, it will be required to improve the accuracy of the thermal-hydraulic and neutronic model. In addition, regarding the current best estimate codes, the improvement in the user interface and the numerics will be needed.
Verification of thermal-irradiation stress analytical code VIENUS of graphite block
International Nuclear Information System (INIS)
Iyoku, Tatsuo; Ishihara, Masahiro; Shiozawa, Shusaku; Shirai, Hiroshi; Minato, Kazuo.
1992-02-01
The core graphite components of the High Temperature Engineering Test Reactor (HTTR) show both the dimensional change (irradiation shrinkage) and creep behavior due to fast neutron irradiation under the temperature and the fast neutron irradiation conditions of the HTTR. Therefore, thermal/irradiation stress analytical code, VIENUS, which treats these graphite irradiation behavior, is to be employed in order to design the core components such as fuel block etc. of the HTTR. The VIENUS is a two dimensional finite element viscoelastic stress analytical code to take account of changes in mechanical properties, thermal strain, irradiation-induced dimensional change and creep in the fast neutron irradiation environment. Verification analyses were carried out in order to prove the validity of this code based on the irradiation tests of the 8th OGL-1 fuel assembly and the fuel element of the Peach Bottom reactor. This report describes the outline of the VIENUS code and its verification analyses. (author)
76 FR 57982 - Building Energy Codes Cost Analysis
2011-09-19
... DEPARTMENT OF ENERGY Office of Energy Efficiency and Renewable Energy [Docket No. EERE-2011-BT-BC-0046] Building Energy Codes Cost Analysis Correction In notice document 2011-23236 beginning on page...-23236 Filed 9-16-11; 8:45 am] BILLING CODE 1505-01-P ...
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.
Comparison for the interfacial and wall friction models in thermal-hydraulic system analysis codes
International Nuclear Information System (INIS)
Hwang, Moon Kyu; Park, Jee Won; Chung, Bub Dong; Kim, Soo Hyung; Kim, See Dal
2007-07-01
The average equations employed in the current thermal hydraulic analysis codes need to be closed with the appropriate models and correlations to specify the interphase phenomena along with fluid/structure interactions. This includes both thermal and mechanical interactions. Among the closure laws, an interfacial and wall frictions, which are included in the momentum equations, not only affect pressure drops along the fluid flow, but also have great effects for the numerical stability of the codes. In this study, the interfacial and wall frictions are reviewed for the commonly applied thermal-hydraulic system analysis codes, i.e. RELAP5-3D, MARS-3D, TRAC-M, and CATHARE
Country Report on Building Energy Codes in the United States
Energy Technology Data Exchange (ETDEWEB)
Halverson, Mark A.; Shui, Bin; Evans, Meredydd
2009-04-30
This report is part of a series of reports on building energy efficiency codes in countries associated with the Asian Pacific Partnership (APP) - Australia, South Korea, Japan, China, India, and the United States of America (U.S.). This reports gives an overview of the development of building energy codes in U.S., including national energy policies related to building energy codes, history of building energy codes, recent national projects and activities to promote building energy codes. The report also provides a review of current building energy codes (such as building envelope, HVAC, lighting, and water heating) for commercial and residential buildings in the U.S.
Thermal energy storage and losses in a room-Trombe wall system located in Mexico
International Nuclear Information System (INIS)
Hernández-López, I.; Xamán, J.; Chávez, Y.; Hernández-Pérez, I.; Alvarado-Juárez, R.
2016-01-01
A thermal evaluation of a R-TW system (room with a Trombe wall) is presented. Hourly climatic data of the coldest and the warmest days of 2014 was used to assess the behavior of the R-TW in two cities of Mexico with cold climate (Huitzilac and Toluca). The simulations were done with an in-house code based on the Finite Volume Method. It was found that thermal energy losses through the semitransparent wall are about 60% of the solar radiation incident on the system (G_s_o_l). Despite of the thermal losses, the system gets enough energy to keep the air inside the room with a temperature above 35 °C. For both cities during the coldest day, the maximum energy stored is about 109 MJ and during the warmest day is about 70 MJ. This energy is supplied from the storage wall to the air inside the room during periods without insolation. - Highlights: • Thermal performance of a Room-Trombe Wall system was evaluated under two cold cities. • Thermal energy losses through the semitransparent wall were about 60% of the solar radiation incident of the system. • The maximum energy stored by the Trombe Wall was 109 MJ during the coldest day. • The maximum energy stored by the Trombe Wall was 70 MJ during the warmest day.
Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2
Energy Technology Data Exchange (ETDEWEB)
Coryell, E.W.; Siefken, L.J.; Harvego, E.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States)] [and others
1997-07-01
The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures. The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds.
Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2
International Nuclear Information System (INIS)
Coryell, E.W.; Siefken, L.J.; Harvego, E.A.
1997-01-01
The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures. The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds
Office of Codes and Standards resource book. Section 1, Building energy codes and standards
Energy Technology Data Exchange (ETDEWEB)
Hattrup, M.P.
1995-01-01
The US Department of Energy`s (DOE`s) Office of Codes and Standards has developed this Resource Book to provide: A discussion of DOE involvement in building codes and standards; a current and accurate set of descriptions of residential, commercial, and Federal building codes and standards; information on State contacts, State code status, State building construction unit volume, and State needs; and a list of stakeholders in the building energy codes and standards arena. The Resource Book is considered an evolving document and will be updated occasionally. Users are requested to submit additional data (e.g., more current, widely accepted, and/or documented data) and suggested changes to the address listed below. Please provide sources for all data provided.
Thermal-hydraulic analysis of SMART steam generator tube rupture using TASS/SMR-S code
International Nuclear Information System (INIS)
Kim, Hee-Kyung; Kim, Soo Hyoung; Chung, Young-Jong; Kim, Hyeon-Soo
2013-01-01
Highlights: ► The analysis was performed from the viewpoint of primary coolant leakage. ► The thermal hydraulic responses and the maximum leakage have been identified. ► There is no direct release into the atmosphere caused by an SGTR accident. ► SMART safety system works well against an SGTR accident. - Abstract: A steam generator tube rupture (SGTR) accident analysis for SMART was performed using the TASS/SMR-S code. SMART with a rated thermal power of 330 MWt has been developed at the Korea Atomic Energy Research Institute. The TASS/SMR-S code can analyze the thermal hydraulic phenomena of SMART in a full range of reactor operating conditions. An SGTR is one of the most important accidents from a thermal hydraulic and radiological viewpoint. A conservative analysis against a SMART SGTR was performed. The major concern of this analysis is to find the thermal hydraulic responses and maximum leakage amount from a primary to a secondary side caused by an SGTR accident. A sensitivity study searching for the conservative thermal hydraulic conditions, break locations, reactivity and other conditions was performed. The dominant parameters related with the integral leak are the high RCS pressure, low core inlet coolant temperature and low break location of the SG cassette. The largest integral leak comes to 28 tons in the most conservative case during 1 h. But there is no direct release into the atmosphere because the secondary system pressure is maintained with a sufficient margin for the design pressure. All leaks go to the condenser. The analysis results show that the primary and secondary system pressures are maintained below the design pressure and the SMART safety system is working well against an SGTR accident
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.
Scaling of Thermal-Hydraulic Phenomena and System Code Assessment
International Nuclear Information System (INIS)
Wolfert, K.
2008-01-01
In the last five decades large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Many separate effects tests and integral system tests were carried out to establish a data base for code development and code validation. In this context the question has to be answered, to what extent the results of down-scaled test facilities represent the thermal-hydraulic behaviour expected in a full-scale nuclear reactor under accidental conditions. Scaling principles, developed by many scientists and engineers, present a scientific technical basis and give a valuable orientation for the design of test facilities. However, it is impossible for a down-scaled facility to reproduce all physical phenomena in the correct temporal sequence and in the kind and strength of their occurrence. The designer needs to optimize a down-scaled facility for the processes of primary interest. This leads compulsorily to scaling distortions of other processes with less importance. Taking into account these weak points, a goal oriented code validation strategy is required, based on the analyses of separate effects tests and integral system tests as well as transients occurred in full-scale nuclear reactors. The CSNI validation matrices are an excellent basis for the fulfilling of this task. Separate effects tests in full scale play here an important role.
Local Thermal Insulating Materials For Thermal Energy Storage ...
African Journals Online (AJOL)
Thermal insulation is one of the most important components of a thermal energy storage system. In this paper the thermal properties of selected potential local materials which can be used for high temperature insulation are presented. Thermal properties of seven different samples were measured. Samples consisted of: ...
VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 2. User's manual
International Nuclear Information System (INIS)
Cuta, J.M.; Koontz, A.S.; Stewart, C.W.; Montgomery, S.D.
1983-04-01
VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear energy reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 2: User's Manual) describes the input requirements of VIPRE and its auxiliary programs, SPECSET, ASP and DECCON, and lists the input instructions for each code
Energy Technology Data Exchange (ETDEWEB)
Unal, C., E-mail: cu@lanl.gov [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Williams, B.J. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Yacout, A. [Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, IL 60439 (United States); Higdon, D.M. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)
2013-10-15
Highlights: ► The application of advanced validation techniques (sensitivity, calibration and prediction) to nuclear performance codes FRAPCON and LIFE-4 is the focus of the paper. ► A sensitivity ranking methodology narrows down the number of selected modeling parameters from 61 to 24 for FRAPCON and from 69 to 35 for LIFE-4. ► Fuel creep, fuel thermal conductivity, fission gas transport/release, crack/boundary, and fuel gap conductivity models of LIFE-4 are identified for improvements. ► FRAPCON sensitivity results indicated the importance of the fuel thermal conduction and the fission gas release models. -- Abstract: Evolving nuclear energy programs expect to use enhanced modeling and simulation (M and S) capabilities, using multiscale, multiphysics modeling approaches, to reduce both cost and time from the design through the licensing phases. Interest in the development of the multiscale, multiphysics approach has increased in the last decade because of the need for predictive tools for complex interacting processes as a means of eliminating the limited use of empirically based model development. Complex interacting processes cannot be predicted by analyzing each individual component in isolation. In most cases, the mathematical models of complex processes and their boundary conditions are nonlinear. As a result, the solutions of these mathematical models often require high-performance computing capabilities and resources. The use of multiscale, multiphysics (MS/MP) models in conjunction with high-performance computational software and hardware introduces challenges in validating these predictive tools—traditional methodologies will have to be modified to address these challenges. The advanced MS/MP codes for nuclear fuels and reactors are being developed within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program of the US Department of Energy (DOE) – Nuclear Energy (NE). This paper does not directly address challenges in calibration
Deciphering neuronal population codes for acute thermal pain
Chen, Zhe; Zhang, Qiaosheng; Phuong Sieu Tong, Ai; Manders, Toby R.; Wang, Jing
2017-06-01
Objective. Pain is defined as an unpleasant sensory and emotional experience associated with actual or potential tissue damage, or described in terms of such damage. Current pain research mostly focuses on molecular and synaptic changes at the spinal and peripheral levels. However, a complete understanding of pain mechanisms requires the physiological study of the neocortex. Our goal is to apply a neural decoding approach to read out the onset of acute thermal pain signals, which can be used for brain-machine interface. Approach. We used micro wire arrays to record ensemble neuronal activities from the primary somatosensory cortex (S1) and anterior cingulate cortex (ACC) in freely behaving rats. We further investigated neural codes for acute thermal pain at both single-cell and population levels. To detect the onset of acute thermal pain signals, we developed a novel latent state-space framework to decipher the sorted or unsorted S1 and ACC ensemble spike activities, which reveal information about the onset of pain signals. Main results. The state space analysis allows us to uncover a latent state process that drives the observed ensemble spike activity, and to further detect the ‘neuronal threshold’ for acute thermal pain on a single-trial basis. Our method achieved good detection performance in sensitivity and specificity. In addition, our results suggested that an optimal strategy for detecting the onset of acute thermal pain signals may be based on combined evidence from S1 and ACC population codes. Significance. Our study is the first to detect the onset of acute pain signals based on neuronal ensemble spike activity. It is important from a mechanistic viewpoint as it relates to the significance of S1 and ACC activities in the regulation of the acute pain onset.
A Literature Review of Sealed and Insulated Attics—Thermal, Moisture and Energy Performance
Energy Technology Data Exchange (ETDEWEB)
Less, Brennan [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Walker, Iain [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Levinson, Ronnen [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)
2016-08-01
In this literature review and analysis, we focus on the thermal, moisture and energy performance of sealed and insulated attics in California climates. Thermal. Sealed and insulated attics are expected to maintain attic air temperatures that are similar to those in the house within +/- 10°F. Thermal stress on the assembly, namely high shingle and sheathing temperatures, are of minimal concern. In the past, many sealed and insulated attics were constructed with insufficient insulation levels (~R-20) and with too much air leakage to outside, leading to poor thermal performance. To ensure high performance, sealed and insulated attics in new California homes should be insulated at levels at least equivalent to the flat ceiling requirements in the code, and attic envelopes and ducts should be airtight. We expect that duct systems in well-constructed sealed and insulated attics should have less than 2% HVAC system leakage to outside. Moisture. Moisture risk in sealed and insulated California attics will increase with colder climate regions and more humid outside air in marine zones. Risk is considered low in the hot-dry, highly populated regions of the state, where most new home construction occurs. Indoor humidity levels should be controlled by following code requirements for continuous whole-house ventilation and local exhaust. Pending development of further guidance, we recommend that the air impermeable insulation requirements of the International Residential Code (2012) be used, as they vary with IECC climate region and roof finish. Energy. Sealed and insulated attics provide energy benefits only if HVAC equipment is located in the attic volume, and the benefits depend strongly on the insulation and airtightness of the attic and ducts. Existing homes with leaky, uninsulated ducts in the attic should have major savings. When compared with modern, airtight duct systems in a vented attic, sealed and insulated attics in California may still provide substantial benefit
Validation of thermal hydraulic codes for fusion reactors safety
International Nuclear Information System (INIS)
Sardain, P.; Gulden, W.; Massaut, V.; Takase, K.; Merill, B.; Caruso, G.
2006-01-01
A significant effort has been done worldwide on the validation of thermal hydraulic codes, which can be used for the safety assessment of fusion reactors. This work is an item of an implementing agreement under the umbrella of the International Energy Agency. The European part is supported by EFDA. Several programmes related to transient analysis in water-cooled fusion reactors were run in order to assess the capabilities of the codes to treat the main physical phenomena governing the accidental sequences related to water/steam discharge into the vacuum vessel or the cryostat. The typical phenomena are namely the pressurization of a volume at low initial pressure, the critical flow, the flashing, the relief into an expansion volume, the condensation of vapor in a pressure suppression system, the formation of ice on a cryogenic structure, the heat transfer between walls and fluid in various thermodynamic conditions. · A benchmark exercise has been done involving different types of codes, from homogeneous equilibrium to six equations non-equilibrium models. Several cases were defined, each one focusing on a particular phenomenon. · The ICE (Ingress of Coolant Event) facility has been operated in Japan. It has simulated an in-vessel LOCA and the discharge of steam into a pressure suppression system. · The EVITA (European Vacuum Impingement Test Apparatus) facility has been operated in France. It has simulated ingress of coolant into the cryostat, i.e. into a volume at low initial pressure containing surfaces at cryogenic temperature. This paper gives the main lessons gained from these programs, in particular the possibilities for the improvement of the computer codes, extending their capabilities. For example, the water properties have been extended below the triple point. Ice formation models have been implemented. Work has also been done on condensation models. The remaining needs for R-and-D are also highlighted. (author)
Energy Referencing in LANL HE-EOS Codes
Energy Technology Data Exchange (ETDEWEB)
Leiding, Jeffery Allen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Coe, Joshua Damon [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2017-10-19
Here, We briefly describe the choice of energy referencing in LANL's HE-EOS codes, HEOS and MAGPIE. Understanding this is essential to comparing energies produced by different EOS codes, as well as to the correct calculation of shock Hugoniots of HEs and other materials. In all equations after (3) throughout this report, all energies, enthalpies and volumes are assumed to be molar quantities.
Directory of Open Access Journals (Sweden)
M. AKBARI
2013-12-01
Full Text Available Energy group structure has a significant effect on the results of multigroup transport calculations. It is known that UO2–PUO2 (MOX is a recently developed fuel which consumes recycled plutonium. For such fuel which contains various resonant nuclides, the selection of energy group structure is more crucial comparing to the UO2 fuels. In this paper, in order to improve the accuracy of the integral results in MOX thermal lattices calculated by WIMSD-5B code, a swarm intelligence method is employed to optimize the energy group structure of WIMS library. In this process, the NJOY code system is used to generate the 69 group cross sections of WIMS code for the specified energy structure. In addition, the multiplication factor and spectral indices are compared against the results of continuous energy MCNP-4C code for evaluating the energy group structure. Calculations performed in four different types of H2O moderated UO2–PuO2 (MOX lattices show that the optimized energy structure obtains more accurate results in comparison with the WIMS original structure.
Development of Nuclear Energy Security Code
International Nuclear Information System (INIS)
Shimamura, Takehisa; Suzuki, Atsuyuki; Okubo, Hiroo; Kikuchi, Masahiro.
1990-01-01
In establishing of the nuclear fuel cycle in Japan that have a vulnerability in own energy structure, an effectiveness of energy security should be taken into account as well as an economy based on the balance of supply and demand of nuclear fuels. NMCC develops the 'Nuclear Energy Security Code' which was able to evaluate the effectiveness of energy security. Evaluation method adopted in this code is 'Import Premium' which was proposed in 'World Oil', EMF Report 6. The viewpoints of evaluation are as follows: 1. How much uranium fuel quantity can be reduced by using plutonium fuel? 2. How much a sudden rise of fuel cost can be absorbed by establishing the plutonium cycle beforehand the energy crisis? (author)
Ocean Thermal Extractable Energy Visualization
Energy Technology Data Exchange (ETDEWEB)
Ascari, Matthew [Lockheed Martin Corporation, Bethesda, MD (United States)
2012-10-28
The Ocean Thermal Extractable Energy Visualization (OTEEV) project focuses on assessing the Maximum Practicably Extractable Energy (MPEE) from the world’s ocean thermal resources. MPEE is defined as being sustainable and technically feasible, given today’s state-of-the-art ocean energy technology. Under this project the OTEEV team developed a comprehensive Geospatial Information System (GIS) dataset and software tool, and used the tool to provide a meaningful assessment of MPEE from the global and domestic U.S. ocean thermal resources.
A generalized interface module for the coupling of spatial kinetics and thermal-hydraulics codes
Energy Technology Data Exchange (ETDEWEB)
Barber, D.A.; Miller, R.M.; Joo, H.G.; Downar, T.J. [Purdue Univ., West Lafayette, IN (United States). Dept. of Nuclear Engineering; Wang, W. [SCIENTECH, Inc., Rockville, MD (United States); Mousseau, V.A.; Ebert, D.D. [Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research
1999-03-01
A generalized interface module has been developed for the coupling of any thermal-hydraulics code to any spatial kinetics code. The coupling scheme was designed and implemented with emphasis placed on maximizing flexibility while minimizing modifications to the respective codes. In this design, the thermal-hydraulics, general interface, and spatial kinetics codes function independently and utilize the Parallel Virtual Machine software to manage cross-process communication. Using this interface, the USNRC version of the 3D neutron kinetics code, PARCX, has been coupled to the USNRC system analysis codes RELAP5 and TRAC-M. RELAP5/PARCS assessment results are presented for two NEACRP rod ejection benchmark problems and an NEA/OECD main steam line break benchmark problem. The assessment of TRAC-M/PARCS has only recently been initiated, nonetheless, the capabilities of the coupled code are presented for a typical PWR system/core model.
A generalized interface module for the coupling of spatial kinetics and thermal-hydraulics codes
International Nuclear Information System (INIS)
Barber, D.A.; Miller, R.M.; Joo, H.G.; Downar, T.J.; Mousseau, V.A.; Ebert, D.D.
1999-01-01
A generalized interface module has been developed for the coupling of any thermal-hydraulics code to any spatial kinetics code. The coupling scheme was designed and implemented with emphasis placed on maximizing flexibility while minimizing modifications to the respective codes. In this design, the thermal-hydraulics, general interface, and spatial kinetics codes function independently and utilize the Parallel Virtual Machine software to manage cross-process communication. Using this interface, the USNRC version of the 3D neutron kinetics code, PARCX, has been coupled to the USNRC system analysis codes RELAP5 and TRAC-M. RELAP5/PARCS assessment results are presented for two NEACRP rod ejection benchmark problems and an NEA/OECD main steam line break benchmark problem. The assessment of TRAC-M/PARCS has only recently been initiated, nonetheless, the capabilities of the coupled code are presented for a typical PWR system/core model
MCT: a Monte Carlo code for time-dependent neutron thermalization problems
International Nuclear Information System (INIS)
Cupini, E.; Simonini, R.
1974-01-01
In the Monte Carlo simulation of pulse source experiments, the neutron energy spectrum, spatial distribution and total density may be required for a long time after the pulse. If the assemblies are very small, as often occurs in the cases of interest, sophisticated Monte Carlo techniques must be applied which force neutrons to remain in the system during the time interval investigated. In the MCT code a splitting technique has been applied to neutrons exceeding assigned target times, and we have found that this technique compares very favorably with more usual ones, such as the expected leakage probability, giving large gains in computational time and variance. As an example, satisfactory asymptotic thermal spectra with a neutron attenuation of 10 -5 were quickly obtained. (U.S.)
TOWARDS ENERGY-AWARE CODING PRACTICES FOR ANDROID
Directory of Open Access Journals (Sweden)
João SARAIVA
2018-03-01
Full Text Available This paper studies how the use of different coding practices when developing Android applications influence energy consumption. We consider two common Java/Android programming practices, namely string operations and (non cached image loading, and we show the energy profile of different coding practices for doing them. With string operations, we compare the performance of the usage of the standard String class to the usage of the StringBuilder class, while with our second practice we evaluate the benefits of image caching with asynchronous loading. We externally measure energy consumption of the example applications using the Trepn profiler application by Qualcomm. Our preliminary results show that selected coding practices do significantly affect energy consumption, in the particular cases of our practice selection, this difference varies between 20% and 50%.
Energy Technology Data Exchange (ETDEWEB)
Carderi, A.
1991-01-01
The INCAS residential building energy savings design code comes complete with data banks giving designers all the necessary technical and economic information and energy savings options for the code's efficacious application to obtain the optimum energy efficient/cost beneficial solution. This user's manual contains descriptions of the types of data to be input, the code's methodology, the data banks, and complete instructions for the code's implementation. The energy savings alternatives include: choice of heating plant and fuel; choice, application location and thickness of thermal insulation. For the case of building heating system retrofits, the code takes into account the existing condition of building components.
Thermal energy storages analysis for high temperature in air solar systems
International Nuclear Information System (INIS)
Andreozzi, Assunta; Buonomo, Bernardo; Manca, Oronzio; Tamburrino, Salvatore
2014-01-01
In this paper a high temperature thermal storage in a honeycomb solid matrix is numerically investigated and a parametric analysis is accomplished. In the formulation of the model it is assumed that the system geometry is cylindrical, the fluid and the solid thermo physical properties are temperature independent and radiative heat transfer is taken into account whereas the effect of gravity is neglected. Air is employed as working fluid and the solid material is cordierite. The evaluation of the fluid dynamic and thermal behaviors is accomplished assuming the honeycomb as a porous medium. The Brinkman–Forchheimer–extended Darcy model is used in the governing equations and the local thermal non equilibrium is assumed. The commercial CFD Fluent code is used to solve the governing equations in transient regime. Numerical simulations are carried out with storage medium for different mass flow rates of the working fluid and different porosity values. Results in terms of temperature profiles, temperatures fields and stored thermal energy as function of time are presented. The effects of storage medium, different porosity values and mass flow rate on stored thermal energy and storage time are shown. - Highlights: • HTTES in a honeycomb solid matrix is numerically investigated. • The numerical analysis is carried out assuming the honeycomb as a porous medium. • The Brinkman–Forchheimer–extended Darcy model is used in the governing equations. • Results are carried out for different mass flow rates and porosity values. • The main effect is due to the porosity which set the thermal energy storage value
FITPULS: a code for obtaining analytic fits to aggregate fission-product decay-energy spectra
International Nuclear Information System (INIS)
LaBauve, R.J.; George, D.C.; England, T.R.
1980-03-01
The operation and input to the FITPULS code, recently updated to utilize interactive graphics, are described. The code is designed to retrieve data from a library containing aggregate fine-group spectra (150 energy groups) from fission products, collapse the data to few groups (up to 25), and fit the resulting spectra along the cooling time axis with a linear combination of exponential functions. Also given in this report are useful results for aggregate gamma and beta spectra from the decay of fission products released from 235 U irradiated with a pulse (10 -4 s irradiation time) of thermal neutrons. These fits are given in 22 energy groups that are the first 22 groups of the LASL 25-group decay-energy group structure, and the data are expressed both as MeV per fission second and particles per fission second; these pulse functions are readily folded into finite fission histories. 65 figures, 11 tables
INL Experimental Program Roadmap for Thermal Hydraulic Code Validation
Energy Technology Data Exchange (ETDEWEB)
Glenn McCreery; Hugh McIlroy
2007-09-01
Advanced computer modeling and simulation tools and protocols will be heavily relied on for a wide variety of system studies, engineering design activities, and other aspects of the Next Generation Nuclear Power (NGNP) Very High Temperature Reactor (VHTR), the DOE Global Nuclear Energy Partnership (GNEP), and light-water reactors. The goal is for all modeling and simulation tools to be demonstrated accurate and reliable through a formal Verification and Validation (V&V) process, especially where such tools are to be used to establish safety margins and support regulatory compliance, or to design a system in a manner that reduces the role of expensive mockups and prototypes. Recent literature identifies specific experimental principles that must be followed in order to insure that experimental data meet the standards required for a “benchmark” database. Even for well conducted experiments, missing experimental details, such as geometrical definition, data reduction procedures, and manufacturing tolerances have led to poor Benchmark calculations. The INL has a long and deep history of research in thermal hydraulics, especially in the 1960s through 1980s when many programs such as LOFT and Semiscle were devoted to light-water reactor safety research, the EBRII fast reactor was in operation, and a strong geothermal energy program was established. The past can serve as a partial guide for reinvigorating thermal hydraulic research at the laboratory. However, new research programs need to fully incorporate modern experimental methods such as measurement techniques using the latest instrumentation, computerized data reduction, and scaling methodology. The path forward for establishing experimental research for code model validation will require benchmark experiments conducted in suitable facilities located at the INL. This document describes thermal hydraulic facility requirements and candidate buildings and presents examples of suitable validation experiments related
Some neutronics and thermal-hydraulics codes for reactor analysis using personal computers
International Nuclear Information System (INIS)
Woodruff, W.L.
1990-01-01
Some neutronics and thermal-hydraulics codes formerly available only for main frame computers may now be run on personal computers. Brief descriptions of the codes are provided. Running times for some of the codes are compared for an assortment of personal and main frame computers. With some limitations in detail, personal computer versions of the codes can be used to solve many problems of interest in reactor analyses at very modest costs. 11 refs., 4 tabs
VIPRE-01: A thermal-hydraulic code for reactor cores
International Nuclear Information System (INIS)
Cuta, J.M.; Koontz, A.S.; Stewart, C.W.; Montgomery, S.D.; Nomura, K.K.
1989-08-01
The VIPRE-01 thermal hydraulics code for PWR and BWR analysis has undergone significant modifications and error correction. This manual for the updated code, designated as VIPRE-01 Mod-02, describes improvements that eliminate problems of slow convergence with the drift flux model in transient simulation. To update the VIPRE-01 code and its documentation the drift flux model of two-phase flow was implemented and error corrections developed during VIPRE-01 application were included. The project team modified the existing VIPRE-01 equations into drift flux model equations by developing additional terms. They also developed and implemented corrections for the errors identified during the last four years. They then validated the modified code against standard test data using selected test cases. The project team prepared documentation revisions reflecting code improvements and corrections to replace the corresponding sections in the original VIPRE documents. The revised VIPRE code, designated VIPRE-01 Mod-02, incorporates improvements that eliminate many shortcomings of the previous version. During the validation, the code produced satisfactory output compared with test data. The revised documentation is in the form of binder pages to replace existing pages in three of the original manuals
Energy Technology Data Exchange (ETDEWEB)
Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok [Korea Atomic Energy Research Institute, Taejon (Korea)
1998-06-01
A multi-dimensional realistic thermal-hydraulic system analysis code, MARS version 1.3 has been developed. Main purpose of MARS 1.3 development is to have the realistic analysis capability of transient two-phase thermal-hydraulics of Pressurized Water Reactors (PWRs) especially during Large Break Loss of Coolant Accidents (LBLOCAs) where the multi-dimensional phenomena domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, three-dimensional (3D) reactor vessel analysis code, and RELAP5/MOD3.2.1.2, one-dimensional (1D) reactor system analysis code., Developmental requirements for MARS are chosen not only to best utilize the existing capability of the codes but also to have the enhanced capability in code maintenance, user accessibility, user friendliness, code portability, code readability, and code flexibility. For the maintenance of existing codes capability and the enhancement of code maintenance capability, user accessibility and user friendliness, MARS has been unified to be a single code consisting of 1D module (RELAP5) and 3D module (COBRA-TF). This is realized by implicitly integrating the system pressure matrix equations of hydrodynamic models and solving them simultaneously, by modifying the 1D/3D calculation sequence operable under a single Central Processor Unit (CPU) and by unifying the input structure and the light water property routines of both modules. In addition, the code structure of 1D module is completely restructured using the modular data structure of standard FORTRAN 90, which greatly improves the code maintenance capability, readability and portability. For the code flexibility, a dynamic memory management scheme is applied in both modules. MARS 1.3 now runs on PC/Windows and HP/UNIX platforms having a single CPU, and users have the options to select the 3D module to model the 3D thermal-hydraulics in the reactor vessel or other
The Cost of Enforcing Building Energy Codes: Phase 1
Energy Technology Data Exchange (ETDEWEB)
Williams, Alison [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Vine, Ed [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Price, Sarah [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Sturges, Andrew [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Rosenquist, Greg [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)
2013-04-01
The purpose of this literature review is to summarize key findings regarding the costs associated with enforcing building energy code compliance—primarily focusing on costs borne by local government. The review takes into consideration over 150 documents that discuss, to some extent, code enforcement. This review emphasizes those documents that specifically focus on costs associated with energy code enforcement. Given the low rates of building energy code compliance that have been reported in existing studies, as well as the many barriers to both energy code compliance and enforcement, this study seeks to identify the costs of initiatives to improve compliance and enforcement. Costs are reported primarily as presented in the original source. Some costs are given on a per home or per building basis, and others are provided for jurisdictions of a certain size. This literature review gives an overview of state-based compliance rates, barriers to code enforcement, and U.S. Department of Energy (DOE) and key stakeholder involvement in improving compliance with building energy codes. In addition, the processes and costs associated with compliance and enforcement of building energy codes are presented. The second phase of this study, which will be presented in a different report, will consist of surveying 34 experts in the building industry at the national and state or local levels in order to obtain additional cost information, building on the findings from the first phase, as well as recommendations for where to most effectively spend money on compliance and enforcement.
Building Energy Efficiency in India: Compliance Evaluation of Energy Conservation Building Code
Energy Technology Data Exchange (ETDEWEB)
Yu, Sha; Evans, Meredydd; Delgado, Alison
2014-03-26
India is experiencing unprecedented construction boom. The country doubled its floorspace between 2001 and 2005 and is expected to add 35 billion m2 of new buildings by 2050. Buildings account for 35% of total final energy consumption in India today, and building energy use is growing at 8% annually. Studies have shown that carbon policies will have little effect on reducing building energy demand. Chaturvedi et al. predicted that, if there is no specific sectoral policies to curb building energy use, final energy demand of the Indian building sector will grow over five times by the end of this century, driven by rapid income and population growth. The growing energy demand in buildings is accompanied by a transition from traditional biomass to commercial fuels, particularly an increase in electricity use. This also leads to a rapid increase in carbon emissions and aggravates power shortage in India. Growth in building energy use poses challenges to the Indian government. To curb energy consumption in buildings, the Indian government issued the Energy Conservation Building Code (ECBC) in 2007, which applies to commercial buildings with a connected load of 100 kW or 120kVA. It is predicted that the implementation of ECBC can help save 25-40% of energy, compared to reference buildings without energy-efficiency measures. However, the impact of ECBC depends on the effectiveness of its enforcement and compliance. Currently, the majority of buildings in India are not ECBC-compliant. The United Nations Development Programme projected that code compliance in India would reach 35% by 2015 and 64% by 2017. Whether the projected targets can be achieved depends on how the code enforcement system is designed and implemented. Although the development of ECBC lies in the hands of the national government – the Bureau of Energy Efficiency under the Ministry of Power, the adoption and implementation of ECBC largely relies on state and local governments. Six years after ECBC
Temperature Distribution and Thermal Performance of an Aquifer Thermal Energy Storage System
Ganguly, Sayantan
2017-04-01
Energy conservation and storage has become very crucial to make use of excess energy during times of future demand. Excess thermal energy can be captured and stored in aquifers and this technique is termed as Aquifer Thermal Energy Storage (ATES). Storing seasonal thermal energy in water by injecting it into subsurface and extracting in time of demand is the principle of an ATES system. Using ATES systems leads to energy savings, reduces the dependency on fossil fuels and thus leads to reduction in greenhouse gas emission. This study numerically models an ATES system to store seasonal thermal energy and evaluates the performance of it. A 3D thermo-hydrogeological numerical model for a confined ATES system is presented in this study. The model includes heat transport processes of advection, conduction and heat loss to confining rock media. The model also takes into account regional groundwater flow in the aquifer, geothermal gradient and anisotropy in the aquifer. Results show that thermal injection into the aquifer results in the generation of a thermal-front which grows in size with time. Premature thermal-breakthrough causes thermal interference in the system when the thermal-front reaches the production well and consequences in the fall of system performance and hence should be avoided. This study models the transient temperature distribution in the aquifer for different flow and geological conditions. This may be effectively used in designing an efficient ATES project by ensuring safety from thermal-breakthrough while catering to the energy demand. Based on the model results a safe well spacing is proposed. The thermal energy discharged by the system is determined and strategy to avoid the premature thermal-breakthrough in critical cases is discussed. The present numerical model is applied to simulate an experimental field study which is found to approximate the field results quite well.
Applications of the Los Alamos High Energy Transport code
International Nuclear Information System (INIS)
Waters, L.; Gavron, A.; Prael, R.E.
1992-01-01
Simulation codes reliable through a large range of energies are essential to analyze the environment of vehicles and habitats proposed for space exploration. The LAHET monte carlo code has recently been expanded to track high energy hadrons with FLUKA, while retaining the original Los Alamos version of HETC at lower energies. Electrons and photons are transported with EGS4, and an interface to the MCNP monte carlo code is provided to analyze neutrons with kinetic energies less than 20 MeV. These codes are benchmarked by comparison of LAHET/MCNP calculations to data from the Brookhaven experiment E814 participant calorimeter
Energy Technology Data Exchange (ETDEWEB)
Park, Youngjae; Kim, Iljin; Kim, Hyungdae [Kyung Hee University, Yongin (Korea, Republic of)
2015-10-15
Diverse integral/small-modular reactors (SMRs) have been developed. Once-through steam generator (OTSG) which generates superheated steam without steam separator and dryer was used in the SMRs to reduce volume of steam generator. It would be possible to design a new steam generator with best estimate thermal-hydraulic codes such as RELAP and MARS. However, it is not convenience to use the general purpose thermal-hydraulic analysis code to design a specific component of nuclear power plants. A widely used simulation tool for thermal-hydraulic analysis of drum-type steam generators is ATHOS, which allows 3D analysis. On the other hand, a simple 1D thermal-hydraulic analysis code might be accurate enough for the conceptual design of OTSG. In this study, thermal-hydraulic analysis code for conceptual design of OTSG was developed using 1D homogeneous equilibrium model (HEM). A benchmark calculation was also conducted to verify and validate the prediction accuracy of the developed code by comparing with the analysis results with MARS. Finally, conceptual design of OTSG was conducted by the developed code. A simple 1D thermal-hydraulic analysis code was developed for the purpose of conceptual design OTSG for SMRs. A set of benchmark calculations was conducted to verify and validate the analysis accuracy of the developed code by comparing results obtained with a best-estimated thermal-hydraulic analysis code, MARS. Finally, analysis of two different OTSG design concepts with superheating and recirculation was demonstrated using the developed code.
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
International Nuclear Information System (INIS)
Yudov, Y.V.
2001-01-01
The functional part of the KORSAR computer code is based on the computational unit for the reactor system thermal-hydraulics and other thermal power systems with water cooling. The two-phase flow dynamics of the thermal-hydraulic network is modelled by KORSAR in one-dimensional two-fluid (non-equilibrium and nonhomogeneous) approximation with the same pressure of both phases. Each phase is characterized by parameters averaged over the channel sections, and described by the conservation equations for mass, energy and momentum. The KORSAR computer code relies upon a novel approach to mathematical modelling of two-phase dispersed-annular flows. This approach allows a two-fluid model to differentiate the effects of the liquid film and droplets in the gas core on the flow characteristics. A semi-implicit numerical scheme has been chosen for deriving discrete analogs the conservation equations in KORSAR. In the semi-implicit numerical scheme, solution of finite-difference equations is reduced to the problem of determining the pressure field at a new time level. For the one-channel case, the pressure field is found from the solution of a system of linear algebraic equations by using the tri-diagonal matrix method. In the branched network calculation, the matrix of coefficients in the equations describing the pressure field is no longer tri-diagonal but has a sparseness structure. In this case, the system of linear equations for the pressure field can be solved with any of the known classical methods. Such an approach is implemented in the existing best-estimate thermal-hydraulic computer codes (TRAC, RELAP5, etc.) For the KORSAR computer code, we have developed a new non-iterative method for calculating the pressure field in the network of any topology. This method is based on the tri-diagonal matrix method and performs well when solving the thermal-hydraulic network problems. (author)
Thermal-hydraulic analysis of PWR core including intermediate flow mixers with the THYC code
International Nuclear Information System (INIS)
Mur, J.; Meignin, J.C.
1997-07-01
Departure from nucleate boiling (DNB) is one of the major limiting factors of pressurized water reactors (PWRs). Safety requires that occurrence of DNB should be precluded under normal or incidental operating conditions. The thermal-hydraulic THYC code developed by EDF is described. The code is devoted to heat and mass transfer in nuclear components. Critical Heat Flux (CHF) is predicted from local thermal-hydraulic parameters such as pressure, mass flow rate, and quality. A three stage methodology to evaluate thermal margins in order to perform standard core design is described. (K.A.)
Thermal-hydraulic analysis of PWR core including intermediate flow mixers with the THYC code
Energy Technology Data Exchange (ETDEWEB)
Mur, J. [Electricite de France (EDF), 78 - Chatou (France); Meignin, J.C. [Electricite de France (EDF), 69 - Villeurbanne (France)
1997-07-01
Departure from nucleate boiling (DNB) is one of the major limiting factors of pressurized water reactors (PWRs). Safety requires that occurrence of DNB should be precluded under normal or incidental operating conditions. The thermal-hydraulic THYC code developed by EDF is described. The code is devoted to heat and mass transfer in nuclear components. Critical Heat Flux (CHF) is predicted from local thermal-hydraulic parameters such as pressure, mass flow rate, and quality. A three stage methodology to evaluate thermal margins in order to perform standard core design is described. (K.A.) 8 refs.
LWR containment thermal hydraulic codes benchmark demona B3 exercise
International Nuclear Information System (INIS)
Della Loggia, E.; Gauvain, J.
1988-01-01
Recent discussion about the aerosol codes currently used for the analysis of containment retention capabilities have revealed a number of questions concerning the reliabilities and verifications of the thermal-hydraulic modules of these codes with respect to the validity of implemented physical models and the stability and effectiveness of numerical schemes. Since these codes are used for the calculation of the Source Term for the assessment of radiological consequences of severe accidents, they are an important part of reactor safety evaluation. For this reason the Commission of European Communities (CEC), following the recommendation mode by experts from Member Stades, is promoting research in this field with the aim also of establishing and increasing collaboration among Research Organisations of member countries. In view of the results of the studies, the CEC has decided to carry out a Benchmark exercise for severe accident containment thermal hydraulics codes. This exercise is based on experiment B3 in the DEMONA programme. The main objective of the benchmark exercise has been to assess the ability of the participating codes to predict atmosphere saturation levels and bulk condensation rates under conditions similar to those predicted to follow a severe accident in a PWR. This exercise follows logically on from the LA-4 exercise, which, is related to an experiment with a simpler internal geometry. We present here the results obtained so far and from them preliminary conclusions are drawn, concerning condensation temperature, pressure, flow rates, in the reactor containment
Development of subchannel analysis code MATRA-LMR for KALIMER subassembly thermal-hydraulics
International Nuclear Information System (INIS)
Won-Seok Kim; Young-Gyun Kim
2000-01-01
In the sodium cooled liquid metal reactors, the design limit are imposed on the maximum temperatures of claddings and fuel pins. Thus an accurate prediction of core coolant/fuel temperature distribution is essential to the LMR core thermal-hydraulic design. The detailed subchannel thermal-hydraulic analysis code MATRA-LMR (Multichannel Analyzer for Steady States and Transients in Rod Arrays for Liquid Metal Reactors) is being developed for KALIMER core design and analysis, based on COBRA-IV-i and MATRA. The major modifications and improvements implemented into MATRA-LMR are as follows: a) nonuniform axial noding capability, b) sodium properties calculation subprogram, c) sodium coolant heat transfer correlations, and d) most recent pressure drop correlations, such as Novendstern, Chiu-Rohsenow-Todreas and Cheng-Todreas. To assess the development status of this code, the benchmark calculations were performed with the ORNL 19 pin tests and EBR-II seven-assembly SLTHEN calculation results. The calculation results of MATRA-LMR for ORNL 19-pin assembly tests and EBR-II 91-pin experiments were compared to the measurements, and to SABRE4 and SLTHEN code calculation results, respectively. In this comparison, the differences are found among the three codes because of the pressure drop and the thermal mixing modellings. Finally, the major technical results of the conceptual design for the KALIMER 98.03 core have been compared with the calculations of MATRA-LMR, SABRE4 and SLTHEN codes. (author)
Parallelization methods study of thermal-hydraulics codes
International Nuclear Information System (INIS)
Gaudart, Catherine
2000-01-01
The variety of parallelization methods and machines leads to a wide selection for programmers. In this study we suggest, in an industrial context, some solutions from the experience acquired through different parallelization methods. The study is about several scientific codes which simulate a large variety of thermal-hydraulics phenomena. A bibliography on parallelization methods and a first analysis of the codes showed the difficulty of our process on the whole applications to study. Therefore, it would be necessary to identify and extract a representative part of these applications and parallelization methods. The linear solver part of the codes forced itself. On this particular part several parallelization methods had been used. From these developments one could estimate the necessary work for a non initiate programmer to parallelize his application, and the impact of the development constraints. The different methods of parallelization tested are the numerical library PETSc, the parallelizer PAF, the language HPF, the formalism PEI and the communications library MPI and PYM. In order to test several methods on different applications and to follow the constraint of minimization of the modifications in codes, a tool called SPS (Server of Parallel Solvers) had be developed. We propose to describe the different constraints about the optimization of codes in an industrial context, to present the solutions given by the tool SPS, to show the development of the linear solver part with the tested parallelization methods and lastly to compare the results against the imposed criteria. (author) [fr
Verification of the thermal module in the ELESIM code and the associated uncertainty analysis
International Nuclear Information System (INIS)
Arimescu, V.I.; Williams, A.F.; Klein, M.E.; Richmond, W.R.; Couture, M.
1997-01-01
Temperature is a critical parameter in fuel modelling because most of the physical processes that occur in fuel elements during irradiation are thermally activated. The focus of this paper is the temperature distribution calculation used in the computer code ELESIM, developed at AECL to model the steady state behaviour of CANDU fuel. A validation procedure for fuel codes is described and applied to ELESIM's thermal calculation
International Nuclear Information System (INIS)
Santos, Thiago A. dos; Maiorino, José R.; Stefanni, Giovanni L. de
2017-01-01
In order to project a nuclear reactor, the neutronic calculus must be validated, so that its thermal limits and safety parameters are respected. Considering this issue, this research aims to evaluate the APTh-100 reactor thermal limits. This PWR is a project developed in Universidade Federal do ABC (UFABC) using fuel composed of Uranium and Thorium oxide mixed (U,Th)O 2 . For this purpose, a simplified, although conservative, code was developed in a MATLAB environment named STC-MOX-Th 'Simplified Thermal-hydraulics Code-Mixed Oxide Thorium'. This code provides axial and radial temperature distribution, as well as DNBR distribution over the hottest channel of the reactor core. Moreover, it brings other hydraulic quantities, such as pressure drop over the fuel rod, considering any fuel proportion of (U,Th)O 2 .The software uses basic laws of conservation of mass, momentum and energy, it also calculates the thermal conduction equation, considering the thermal conductive coefficient as a temperature function. In order to solve this equation, the finite elements method was used. Furthermore, the proportion of 36% of UO 2 was used to evaluate the temperature over the fuel rod and DNBR minimum in three burn conditions: beginning, middle and ending. The program has proven to be efficient in every condition and the results evidenced that the APTh-1000 reactor, in an initial analysis, has its thermal limits within the recommended security parameters. (author)
Thermal performance and heat transport in aquifer thermal energy storage
Sommer, W.T.; Doornenbal, P.J.; Drijver, B.C.; Gaans, van P.F.M.; Leusbrock, I.; Grotenhuis, J.T.C.; Rijnaarts, H.H.M.
2014-01-01
Aquifer thermal energy storage (ATES) is used for seasonal storage of large quantities of thermal energy. Due to the increasing demand for sustainable energy, the number of ATES systems has increased rapidly, which has raised questions on the effect of ATES systems on their surroundings as well as
A comparison of thermal algorithms of fuel rod performance code systems
International Nuclear Information System (INIS)
Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C.
2003-11-01
The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance
A comparison of thermal algorithms of fuel rod performance code systems
Energy Technology Data Exchange (ETDEWEB)
Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C
2003-11-01
The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance.
Solar energy thermally powered electrical generating system
Owens, William R. (Inventor)
1989-01-01
A thermally powered electrical generating system for use in a space vehicle is disclosed. The rate of storage in a thermal energy storage medium is controlled by varying the rate of generation and dissipation of electrical energy in a thermally powered electrical generating system which is powered from heat stored in the thermal energy storage medium without exceeding a maximum quantity of heat. A control system (10) varies the rate at which electrical energy is generated by the electrical generating system and the rate at which electrical energy is consumed by a variable parasitic electrical load to cause storage of an amount of thermal energy in the thermal energy storage system at the end of a period of insolation which is sufficient to satisfy the scheduled demand for electrical power to be generated during the next period of eclipse. The control system is based upon Kalman filter theory.
A two-compartment thermal-hydraulic experiment (LACE-LA4) analyzed by ESCADRE code
International Nuclear Information System (INIS)
Passalacqua, R.
1994-01-01
Large scale experiments show that whenever a Loss of Coolant Accident (LOCA) occurs, water pools are generated. Stratifications of steam saturated gas develop above water pools causing a two-compartment thermal-hydraulics. The LACE (LWR Advanced Containment Experiment) LA4 experiment, performed at the Hanford Engineering Development Laboratory (HEDL), exhibited a strong stratification, at all times, above a growing water pool. JERICHO and AEROSOLS-B2 are part of the ESCADRE code system (Ensemble de Systemes de Codes d'Analyse d'accident Des Reacteurs A Eau), a tool for evaluating the response of a nuclear plant to severe accidents. These two codes are here used to simulate respectively the thermal-hydraulics and the associated aerosol behavior. Code results have shown that modelling large containment thermal-hydraulics without taking account of the stratification phenomenon leads to large overpredictions of containment pressure and temperature. If the stratification is modelled as a zone with a higher steam condensation rate and a higher thermal resistance, ESCADRE predictions match quite well experimental data. The stratification thermal-hydraulics is controlled by power (heat fluxes) repartition in the lower compartment between the water pool and the nearby walls. Therefore the total, direct heat exchange between the two compartment is reduced. Stratification modelling is believed to be important for its influence on aerosol behavior: aerosol deposition through the inter-face of the two subcompartments is improved by diffusiophoresis and thermophoresis. In addition the aerosol concentration gradient, through the stratification, will cause a driving force for motion of smaller particles towards the pool. (author)
Solar-thermal conversion and thermal energy storage of graphene foam-based composite
Zhang, Lianbin
2016-07-11
Among various utilizations of solar energy, solar-thermal conversion has recently gained renewed research interest due to its extremely high energy efficiency. However, one limiting factor common to all solar-based energy conversion technologies is the intermittent nature of solar irradiation, which makes them unable to stand-alone to satisfy continuous energy need. Herein, we report a three-dimensional (3D) graphene foam and phase change material (PCM) composite for the seamlessly combined solar-thermal conversion and thermal storage for sustained energy release. The composite is obtained by infiltrating the 3D graphene foam with a commonly used PCM, paraffin wax. The high macroporosity and low density of the graphene foam allow for high weight fraction of the PCM to be incorporated, which enhances heat storage capacity of the composite. The interconnected graphene sheets in the composite provide (1) the solar-thermal conversion capability, (2) high thermal conductivity and (3) form stability of the composite. Under light irradiation, the composite effectively collects and converts the light energy into thermal energy, and the converted thermal energy is stored in the PCM and released in an elongated period of time for sustained utilization. This study provides a promising route for sustainable utilization of solar energy.
Solar-thermal conversion and thermal energy storage of graphene foam-based composites.
Zhang, Lianbin; Li, Renyuan; Tang, Bo; Wang, Peng
2016-08-14
Among various utilizations of solar energy, solar-thermal conversion has recently gained renewed research interest due to its extremely high energy efficiency. However, one limiting factor common to all solar-based energy conversion technologies is the intermittent nature of solar irradiation, which makes them unable to stand-alone to satisfy the continuous energy need. Herein, we report a three-dimensional (3D) graphene foam and phase change material (PCM) composite for the seamlessly combined solar-thermal conversion and thermal storage for sustained energy release. The composite is obtained by infiltrating the 3D graphene foam with a commonly used PCM, paraffin wax. The high macroporosity and low density of the graphene foam allow for high weight fraction of the PCM to be incorporated, which enhances the heat storage capacity of the composite. The interconnected graphene sheets in the composite provide (1) the solar-thermal conversion capability, (2) high thermal conductivity and (3) form stability of the composite. Under light irradiation, the composite effectively collects and converts the light energy into thermal energy, and the converted thermal energy is stored in the PCM and released in an elongated period of time for sustained utilization. This study provides a promising route for sustainable utilization of solar energy.
Energy Technology Data Exchange (ETDEWEB)
Peniguel, C [Electricite de France (EDF), 78 - Chatou (France). Direction des Etudes et Recherches; Rupp, I [SIMULOG, 78 - Guyancourt (France)
1997-06-01
EDF has developed numerical codes for modeling the conductive, radiative and convective thermal transfers and their couplings in complex industrial configurations: the convection in a fluid is solved by Estet in finite volumes or N3S in finite elements, the conduction is solved by Syrthes and the wall-to-wall thermal radiation is modelled by Syrthes with the help of a radiosity method. Syrthes controls the different heat exchanges which may occur between fluid and solid domains, using an explicit iterative method. An extension of Syrthes has been developed in order to allow the consideration of configurations where several fluid codes operate simultaneously, using ``message passing`` tools such as PVM (Parallel Virtual Machine) and the Calcium code coupler developed at EDF. Application examples are given
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
Current status of high energy nucleon-meson transport code
Energy Technology Data Exchange (ETDEWEB)
Takada, Hiroshi; Sasa, Toshinobu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1998-03-01
Current status of design code of accelerator (NMTC/JAERI code), outline of physical model and evaluation of accuracy of code were reported. To evaluate the nuclear performance of accelerator and strong spallation neutron origin, the nuclear reaction between high energy proton and target nuclide and behaviors of various produced particles are necessary. The nuclear design of spallation neutron system used a calculation code system connected the high energy nucleon{center_dot}meson transport code and the neutron{center_dot}photon transport code. NMTC/JAERI is described by the particle evaporation process under consideration of competition reaction of intranuclear cascade and fission process. Particle transport calculation was carried out for proton, neutron, {pi}- and {mu}-meson. To verify and improve accuracy of high energy nucleon-meson transport code, data of spallation and spallation neutron fragment by the integral experiment were collected. (S.Y.)
Application of RELAP5-3D code for thermal analysis of the ADS reactor core
International Nuclear Information System (INIS)
Fernandes, Gustavo Henrique Nazareno
2018-01-01
Nuclear power is essential to supply global energy demand. Therefore, in order to use nuclear fuel more efficiently, more efficient nuclear reactors technologies researches have been intensified, such as hybrid systems, composed of particle accelerators coupled into nuclear reactors. In order to add knowledge to such studies, an innovative reactor design was considered where the RELAP5-3D thermal-hydraulic analysis code was used to perform a thermal analysis of the core, either in stationary operation or in situations transitory. The addition of new kind of coolants, such as, liquid salts, among them Flibe, lead, lead-bismuth, sodium, lithium-bismuth and lithium-lead was an important advance in this version of the code, making possible to do the thermal simulation of reactors that use these types of coolants. The reactor, object of study in this work, is an innovative reactor, due to its ability to operate in association with an Accelerator Driven System (ADS), considered a predecessor system of the next generation of nuclear reactors (GEN IV). The reactor selected was the MYRRHA (Multi-purpose Hybrid Research Reactor for High tech Applications) due to the availability of data to perform the simulation. In the modeling of the reactor with the code RELAP5-3D, the core was simulated using nodules with 1, 7, 15 and 51 thermohydraulic channels and eutectic lead-bismuth (LBE) as coolant. The parameters, such as, pressure, mass flow and coolant and heat structure temperature were analyzed. In addition, the thermal behavior of the core was evaluated by varying the type of coolant (sodium) in substitution for the LBE of the original design using the model with 7 thermohydraulic channels. The results of the steady-state calculations were compared with data from the literature and the proposed models were verified certifying the ability of the RELAP5-3D code to simulate this innovative reactor. After this step, it was analysed cases of transients with loss of coolant flow
Thermal energy at the nanoscale
Fisher, Timothy S
2014-01-01
These lecture notes provide a detailed treatment of the thermal energy storage and transport by conduction in natural and fabricated structures. Thermal energy in two carriers, i.e. phonons and electrons -- are explored from first principles. For solid-state transport, a common Landauer framework is used for heat flow. Issues including the quantum of thermal conductance, ballistic interface resistance, and carrier scattering are elucidated. Bulk material properties, such as thermal and electrical conductivity, are derived from particle transport theories, and the effects of spatial confinement on these properties are established. Readership: Students and professionals in physics and engineering.
Verification of the thermal module in the ELESIM code and the associated uncertainty analysis
International Nuclear Information System (INIS)
Arimescu, V.I.; Williams, A.F.; Klein, M.E.; Richmond, W.R.; Couture, M.
1997-09-01
Temperature is a critical parameter in fuel modelling because most of the physical processes that occur in fuel elements during irradiation are thermally activated. The focus of this paper is the temperature distribution calculation used in the computer code ELESIM, developed at AECL to model the steady-state behaviour of CANDU fuel. A validation procedure for fuel codes is described and applied to ELESIM's thermal calculation.The effects of uncertainties in model parameters, like Uranium Dioxide thermal conductivity, and input variables, such as fuel element linear power, are accounted for through an uncertainty analysis using Response Surface and Monte Carlo techniques
Thermal comfort in residential buildings: Comfort values and scales for building energy simulation
Energy Technology Data Exchange (ETDEWEB)
Peeters, Leen; D' haeseleer, William [Division of Applied Mechanics and Energy Conversion, University of Leuven (K.U.Leuven), Celestijnenlaan 300 A, B-3001 Leuven (Belgium); Dear, Richard de [Division of Environmental and Life Sciences, Macquarie University, Sydney (Australia); Hensen, Jan [Faculty of Architecture, Building and Planning, Technische Universiteit Eindhoven, Vertigo 6.18, P.O. Box 513, 5600 MB Eindhoven (Netherlands)
2009-05-15
Building Energy Simulation (BES) programmes often use conventional thermal comfort theories to make decisions, whilst recent research in the field of thermal comfort clearly shows that important effects are not incorporated. The conventional theories of thermal comfort were set up based on steady state laboratory experiments. This, however, is not representing the real situation in buildings, especially not when focusing on residential buildings. Therefore, in present analysis, recent reviews and adaptations are considered to extract acceptable temperature ranges and comfort scales. They will be defined in an algorithm, easily implementable in any BES code. The focus is on comfortable temperature levels in the room, more than on the detailed temperature distribution within that room. (author)
International Nuclear Information System (INIS)
Banerjee, S.; Hassan, Y.A.
1995-01-01
Condensation in the presence of noncondensible gases plays an important role in the nuclear industry. The RELAP5/MOD3 thermal hydraulic code was used to study the ability of the code to predict this phenomenon. Two separate effects experiments were simulated using this code. These were the Massachusetts Institute of Technology's (MIT) Pressurizer Experiment, the MIT Single Tube Experiment. A new iterative approach to calculate the interface temperature and the degraded heat transfer coefficient was developed and implemented in the RELAP5/MOD3 thermal hydraulic code. This model employs the heat transfer simultaneously. This model was found to perform much better than the reduction factor approach. The calculations using the new model were found to be in much better agreement with the experimental values
Energy Technology Data Exchange (ETDEWEB)
Banerjee, S.; Hassan, Y.A. [Texas A& M Univ., College Station, TX (United States)
1995-09-01
Condensation in the presence of noncondensible gases plays an important role in the nuclear industry. The RELAP5/MOD3 thermal hydraulic code was used to study the ability of the code to predict this phenomenon. Two separate effects experiments were simulated using this code. These were the Massachusetts Institute of Technology`s (MIT) Pressurizer Experiment, the MIT Single Tube Experiment. A new iterative approach to calculate the interface temperature and the degraded heat transfer coefficient was developed and implemented in the RELAP5/MOD3 thermal hydraulic code. This model employs the heat transfer simultaneously. This model was found to perform much better than the reduction factor approach. The calculations using the new model were found to be in much better agreement with the experimental values.
International Nuclear Information System (INIS)
Ichikawa, Ryoko; Masuhara, Yasuhiro; Kasahara, Fumio
2012-01-01
The Best Estimate Plus Uncertainty (BEPU) method has been prepared for the regulatory cross-check analysis at Japan Nuclear Energy Safety Organization (JNES) on base of the three-dimensional neutron-kinetics/thermal-hydraulics coupled code SKETCH-INS/TRACE5.0. In the preparation, TRACE5.0 is verified against the large-scale thermal-hydraulic tests carried out with NUPEC facility. These tests were focused on the pressure drop of steam-liquid two phase flow and void fraction distribution. From the comparison of the experimental data with other codes (RELAP5/MOD3.3 and TRAC-BF1), TRACE5.0 was judged better than other codes. It was confirmed that TRACE5.0 has high reliability for thermal hydraulics behavior and are used as a best-estimate code for the statistical safety evaluation. Next, the coupled code SKETCH-INS/TRACE5.0 was applied to turbine trip tests performed at the Peach Bottom-2 BWR4 Plant. The turbine trip event shows the rapid power peak due to the voids collapse with the pressure increase. The analyzed peak value of core power is better simulated than the previous version SKETCH-INS/TRAC-BF1. And the statistical safety evaluation using SKETCH-INS/TRACE5.0 was applied to the loss of load transient for examining the influence of the choice of sampling method. (author)
Current and anticipated uses of the thermal hydraulics codes at the NRC
Energy Technology Data Exchange (ETDEWEB)
Caruso, R.
1997-07-01
The focus of Thermal-Hydraulic computer code usage in nuclear regulatory organizations has undergone a considerable shift since the codes were originally conceived. Less work is being done in the area of {open_quotes}Design Basis Accidents,{close_quotes}, and much more emphasis is being placed on analysis of operational events, probabalistic risk/safety assessment, and maintenance practices. All of these areas need support from Thermal-Hydraulic computer codes to model the behavior of plant fluid systems, and they all need the ability to perform large numbers of analyses quickly. It is therefore important for the T/H codes of the future to be able to support these needs, by providing robust, easy-to-use, tools that produce easy-to understand results for a wider community of nuclear professionals. These tools need to take advantage of the great advances that have occurred recently in computer software, by providing users with graphical user interfaces for both input and output. In addition, reduced costs of computer memory and other hardware have removed the need for excessively complex data structures and numerical schemes, which make the codes more difficult and expensive to modify, maintain, and debug, and which increase problem run-times. Future versions of the T/H codes should also be structured in a modular fashion, to allow for the easy incorporation of new correlations, models, or features, and to simplify maintenance and testing. Finally, it is important that future T/H code developers work closely with the code user community, to ensure that the code meet the needs of those users.
Current and anticipated uses of the thermal hydraulics codes at the NRC
International Nuclear Information System (INIS)
Caruso, R.
1997-01-01
The focus of Thermal-Hydraulic computer code usage in nuclear regulatory organizations has undergone a considerable shift since the codes were originally conceived. Less work is being done in the area of Design Basis Accidents, , and much more emphasis is being placed on analysis of operational events, probabalistic risk/safety assessment, and maintenance practices. All of these areas need support from Thermal-Hydraulic computer codes to model the behavior of plant fluid systems, and they all need the ability to perform large numbers of analyses quickly. It is therefore important for the T/H codes of the future to be able to support these needs, by providing robust, easy-to-use, tools that produce easy-to understand results for a wider community of nuclear professionals. These tools need to take advantage of the great advances that have occurred recently in computer software, by providing users with graphical user interfaces for both input and output. In addition, reduced costs of computer memory and other hardware have removed the need for excessively complex data structures and numerical schemes, which make the codes more difficult and expensive to modify, maintain, and debug, and which increase problem run-times. Future versions of the T/H codes should also be structured in a modular fashion, to allow for the easy incorporation of new correlations, models, or features, and to simplify maintenance and testing. Finally, it is important that future T/H code developers work closely with the code user community, to ensure that the code meet the needs of those users
Analysis of the Phebus FPT0 containment thermal hydraulics with the Jericho and Trio-VF codes
International Nuclear Information System (INIS)
Layly, V.D.; Spitz, P.; Mailliat, A.
1994-01-01
This paper presents the analysis of the thermal hydraulic behavior of the containment, during the Phebus FPT0 test performed on December 2, 1993, with the Jericho code which deals with the thermal hydraulics of containment in the severe accident field. This code is part of Escadre which is the French system of codes in charge of predicting PWR severe accidents. After summarizing the relevant Jericho code characteristics and the preliminary assessment work for the Phebus conditions, we briefly describe the REPF 502 test facility and report the thermal hydraulic FPT0 experimental protocol. Then, the experiment / Jericho calculation comparisons are analysed. Because the Jericho code assumes a well-mixed atmosphere, some additional 3-D calculations have been carried out in order to get further insight on the convection flow patterns and qualify the well-mixed atmosphere assumption in the Phebus containment. (author). 9 refs., 12 figs
A thermal engine for underwater glider driven by ocean thermal energy
International Nuclear Information System (INIS)
Yang, Yanan; Wang, Yanhui; Ma, Zhesong; Wang, Shuxin
2016-01-01
Highlights: • Thermal engine with a double-tube structure is developed for underwater glider. • Isostatic pressing technology is effective to increase volumetric change rate. • Actual volumetric change rate reaches 89.2% of the theoretical value. • Long term sailing of 677 km and 27 days is achieved by thermal underwater glider. - Graphical Abstract: - Abstract: Underwater glider is one of the most popular platforms for long term ocean observation. Underwater glider driven by ocean thermal energy extends the duration and range of underwater glider powered by battery. Thermal engine is the core device of underwater glider to harvest ocean thermal energy. In this paper, (1) model of thermal engine was raised by thermodynamics method and the performance of thermal engine was investigated, (2) thermal engine with a double-tube structure was developed and isostatic pressing technology was applied to improve the performance for buoyancy driven, referencing powder pressing theory, (3) wall thickness of thermal engine was optimized to reduce the overall weight of thermal engine, (4) material selection and dimension determination were discussed for a faster heat transfer design, by thermal resistance analysis, (5) laboratory test and long term sea trail were carried out to test the performance of thermal engine. The study shows that volumetric change rate is the most important indicator to evaluating buoyancy-driven performance of a thermal engine, isostatic pressing technology is effective to improve volumetric change rate, actual volumetric change rate can reach 89.2% of the theoretical value and the average power is about 124 W in a typical diving profile. Thermal engine developed by Tianjin University is a superior thermal energy conversion device for underwater glider. Additionally, application of thermal engine provides a new solution for miniaturization of ocean thermal energy conversion.
Phase-Change Thermal Energy Storage
1989-11-01
The goal of this program is to advance the engineering and scientific understanding of solar thermal technology and to establish the technology base from which private industry can develop solar thermal power production options for introduction into the competitive energy market. Solar thermal technology concentrates the solar flux using tracking mirrors or lenses onto a receiver where the solar energy is absorbed as heat and converted into electricity or incorporated into products as process heat. The two primary solar thermal technologies, central receivers and distributed receivers, employ various point and line-focus optics to concentrate sunlight. Current central receiver systems use fields of heliostats (two-axes tracking mirrors) to focus the sun's radiant energy onto a single, tower-mounted receiver. Point focus concentrators up to 17 meters in diameter track the sun in two axes and use parabolic dish mirrors or Fresnel lenses to focus radiant energy onto a receiver. Troughs and bowls are line-focus tracking reflectors that concentrate sunlight onto receiver tubes along their focal lines. Concentrating collector modules can be used alone or in a multimodule system. The concentrated radiant energy absorbed by the solar thermal receiver is transported to the conversion process by a circulating working fluid. Receiver temperatures range from 100 C in low-temperature troughs to over 1500 C in dish and central receiver systems.
Assessment and Application of the ROSE Code for Reactor Outage Thermal-Hydraulic and Safety Analysis
International Nuclear Information System (INIS)
Liang, Thomas K.S.; Ko, F.-K.; Dai, L.-C.
2001-01-01
The currently available tools, such as RELAP5, RETRAN, and others, cannot easily and correctly perform the task of analyzing the system behavior during plant outages. Therefore, a medium-sized program aiming at reactor outage simulation and evaluation, such as midloop operation (MLO) with loss of residual heat removal (RHR), has been developed. Important thermal-hydraulic processes involved during MLO with loss of RHR can be properly simulated by the newly developed reactor outage simulation and evaluation (ROSE) code. The two-region approach with a modified two-fluid model has been adopted to be the theoretical basis of the ROSE code.To verify the analytical model in the first step, posttest calculations against the integral midloop experiments with loss of RHR have been performed. The excellent simulation capacity of the ROSE code against the Institute of Nuclear Energy Research Integral System Test Facility test data is demonstrated. To further mature the ROSE code in simulating a full-sized pressurized water reactor, assessment against the WGOTHIC code and the Maanshan momentary-loss-of-RHR event has been undertaken. The successfully assessed ROSE code is then applied to evaluate the abnormal operation procedure (AOP) with loss of RHR during MLO (AOP 537.4) for the Maanshan plant. The ROSE code also has been successfully transplanted into the Maanshan training simulator to support operator training. How the simulator was upgraded by the ROSE code for MLO will be presented in the future
International Nuclear Information System (INIS)
Ikushima, Takeshi
1991-08-01
A computer program CASKETSS-2 has been developed for the purpose of thermal and structural analysis of nuclear fuel shipping casks. CASKETSS-2 means a modular code system for CASK Evaluation code system Thermal and Structural Safety (Version 2). Main features of CASKETSS-2 are as follow; (1) Thermal and structural analysis computer programs for one-, two-, three-dimensional geometries are contained in the code system. (2) There are simplified computer programs and a detailed one in the structural analysis part in the code system. (3) Input data generator is provided in the code system. (4) Graphic computer program is provided in the code system. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)
Numerical analysis and nuclear standard code application to thermal fatigue
International Nuclear Information System (INIS)
Merola, M.
1992-01-01
The present work describes the Joint Research Centre Ispra contribution to the IAEA benchmark exercise 'Lifetime Behaviour of the First Wall of Fusion Machines'. The results of the numerical analysis of the reference thermal fatigue experiment are presented. Then a discussion on the numerical analysis of thermal stress is tackled, pointing out its particular aspects in view of their influence on the stress field evaluation. As far as the design-allowable number of cycles are concerned the American nuclear code ASME and the French code RCC-MR are applied and the reasons for the different results obtained are investigated. As regards a realistic fatigue lifetime evaluation, the main problems to be solved are brought out. This work, is intended as a preliminary basis for a discussion focusing on the main characteristics of the thermal fatigue problem from both a numerical and a lifetime assessment point of view. In fact the present margin of discretion left to the analyst may cause undue discrepancies in the results obtained. A sensitivity analysis of the main parameters involved is desirable and more precise design procedures should be stated
Solar energy thermalization and storage device
McClelland, J.F.
A passive solar thermalization and thermal energy storage assembly which is visually transparent is described. The assembly consists of two substantial parallel, transparent wall members mounted in a rectangular support frame to form a liquid-tight chamber. A semitransparent thermalization plate is located in the chamber, substantially paralled to and about equidistant from the transparent wall members to thermalize solar radiation which is stored in a transparent thermal energy storage liquid which fills the chamber. A number of the devices, as modules, can be stacked together to construct a visually transparent, thermal storage wall for passive solar-heated buildings.
1995 building energy codes and standards workshops: Summary and documentation
Energy Technology Data Exchange (ETDEWEB)
Sandahl, L.J.; Shankle, D.L.
1996-02-01
During the spring of 1995, Pacific Northwest National Laboratory (PNNL) conducted four two-day Regional Building Energy Codes and Standards workshops across the US. Workshops were held in Chicago, Denver, Rhode Island, and Atlanta. The workshops were designed to benefit state-level officials including staff of building code commissions, energy offices, public utility commissions, and others involved with adopting/updating, implementing, and enforcing building energy codes in their states. The workshops provided an opportunity for state and other officials to learn more about residential and commercial building energy codes and standards, the role of the US Department of Energy and the Building Standards and Guidelines Program at Pacific Northwest National Laboratory, Home Energy Rating Systems (HERS), Energy Efficient Mortgages (EEM), training issues, and other topics related to the development, adoption, implementation, and enforcement of building energy codes. Participants heard success stories, got tips on enforcement training, and received technical support materials. In addition to receiving information on the above topics, workshop participants had an opportunity to provide input on code adoption issues, building industry training issues, building design issues, and exemplary programs across the US. This paper documents the workshop planning, findings, and follow-up processes.
International Nuclear Information System (INIS)
Hirano, Masashi; Hada, Kazuhiko
1990-04-01
The THYDE-HTGR code has been developed for transient thermal-hydraulic analyses of high-temperature gas-cooled reactors, based on the THYDE-W code. THYDE-W is a code developed at JAERI for the simulation of Light Water Reactor plant dynamics during various types of transients including loss-of-coolant accidents. THYDE-HTGR solves the conservation equations of mass, momentum and energy for compressible gas, or single-phase or two-phase flow. The major code modification from THYDE-W is to treat helium loops as well as water loops. In parallel to this, modification has been made for the neutron kinetics to be applicable to helium-cooled graphite-moderated reactors, for the heat transfer models to be applicable to various types of heat exchangers, and so forth. In order to assess the validity of the modifications, analyses of some of the experiments conducted at the High Temperature Test Loop of ERANS have been performed. In this report, the models applied in THYDE-HTGR are described focusing on the present modifications and the results from the assessment calculations are presented. (author)
International Nuclear Information System (INIS)
Gonzalez, J.C.; Leal C, H.
1998-01-01
Some relative aspects to the development and current state of thermal solar energy are summarized, so much at domestic level as international. To facilitate the criteria understanding as the size of the facilities in thermal solar systems, topics as availability of the solar resource and its interactions with the matter are included. Finally, some perspectives for the development of this energetic alternative are presented
Implementation of Energy Code Controls Requirements in New Commercial Buildings
Energy Technology Data Exchange (ETDEWEB)
Rosenberg, Michael I. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hart, Philip R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hatten, Mike [Solarc Energy Group, LLC, Seattle, WA (United States); Jones, Dennis [Group 14 Engineering, Inc., Denver, CO (United States); Cooper, Matthew [Group 14 Engineering, Inc., Denver, CO (United States)
2017-03-24
Most state energy codes in the United States are based on one of two national model codes; ANSI/ASHRAE/IES 90.1 (Standard 90.1) or the International Code Council (ICC) International Energy Conservation Code (IECC). Since 2004, covering the last four cycles of Standard 90.1 updates, about 30% of all new requirements have been related to building controls. These requirements can be difficult to implement and verification is beyond the expertise of most building code officials, yet the assumption in studies that measure the savings from energy codes is that they are implemented and working correctly. The objective of the current research is to evaluate the degree to which high impact controls requirements included in commercial energy codes are properly designed, commissioned and implemented in new buildings. This study also evaluates the degree to which these control requirements are realizing their savings potential. This was done using a three-step process. The first step involved interviewing commissioning agents to get a better understanding of their activities as they relate to energy code required controls measures. The second involved field audits of a sample of commercial buildings to determine whether the code required control measures are being designed, commissioned and correctly implemented and functioning in new buildings. The third step includes compilation and analysis of the information gather during the first two steps. Information gathered during these activities could be valuable to code developers, energy planners, designers, building owners, and building officials.
Thermal hydraulic codes for LWR safety analysis - present status and future perspective
Energy Technology Data Exchange (ETDEWEB)
Staedtke, H. [Commission of the European Union, Ispra (Italy)
1997-07-01
The aim of the present paper is to give a review on the current status and future perspective of present best-estimate Thermal Hydraulic codes. Reference is made to internationally well-established codes which have reached a certain state of maturity. The first part of the paper deals with the common basic code features with respect to the physical modelling and their numerical methods used to describe complex two-phase flow and heat transfer processes. The general predictive capabilities are summarized identifying some remaining code deficiencies and their underlying limitations. The second part discusses various areas including physical modelling, numerical techniques and informatic structure where the codes could be substantially improved.
Thermal hydraulic codes for LWR safety analysis - present status and future perspective
International Nuclear Information System (INIS)
Staedtke, H.
1997-01-01
The aim of the present paper is to give a review on the current status and future perspective of present best-estimate Thermal Hydraulic codes. Reference is made to internationally well-established codes which have reached a certain state of maturity. The first part of the paper deals with the common basic code features with respect to the physical modelling and their numerical methods used to describe complex two-phase flow and heat transfer processes. The general predictive capabilities are summarized identifying some remaining code deficiencies and their underlying limitations. The second part discusses various areas including physical modelling, numerical techniques and informatic structure where the codes could be substantially improved
International Nuclear Information System (INIS)
Chaudri, Khurrum Saleem; Su Yali; Chen Ronghua; Tian Wenxi; Su Guanghui; Qiu Suizheng
2012-01-01
Highlights: ► A tool is developed for coupled neutronics/thermal-hydraulic analysis for SCWR. ► For thermal hydraulic analysis, a sub-channel code SACoS is developed and verified. ► Coupled analysis agree quite well with the reference calculations. ► Different choice of important parameters makes huge difference in design calculations. - Abstract: Supercritical Water Reactor (SCWR) is one of the promising reactors from the list of fourth generation of nuclear reactors. High thermal efficiency and low cost of electricity make it an attractive option in the era of growing energy demand. An almost seven fold density variation for coolant/moderator along the active height does not allow the use of constant density assumption for design calculations, as used for previous generations of reactors. The advancement in computer technology gives us the superior option of performing coupled analysis. Thermal hydraulics calculations of supercritical water systems present extra challenges as not many computational tools are available to perform that job. This paper introduces a new sub-channel code called Sub-channel Analysis Code of SCWR (SACoS) and its application in coupled analyses of High Performance Light Water Reactor (HPLWR). SACoS can compute the basic thermal hydraulic parameters needed for design studies of a supercritical water reactor. Multiple heat transfer and pressure drop correlations are incorporated in the code according to the flow regime. It has the additional capability of calculating the thermal hydraulic parameters of moderator flowing in water box and between fuel assemblies under co-current or counter current flow conditions. Using MCNP4c and SACoS, a coupled system has been developed for SCWR design analyses. The developed coupled system is verified by performing and comparing HPLWR calculations. The results were found to be in very good agreement. Significant difference between the results was seen when Doppler feedback effect was included in
Energy Technology Data Exchange (ETDEWEB)
Rosenberg, Michael I.; Hart, Philip R.; Athalye, Rahul A.; Zhang, Jian; Wang, Weimin
2016-02-15
The US Department of Energy’s most recent commercial energy code compliance evaluation efforts focused on determining a percent compliance rating for states to help them meet requirements under the American Recovery and Reinvestment Act (ARRA) of 2009. That approach included a checklist of code requirements, each of which was graded pass or fail. Percent compliance for any given building was simply the percent of individual requirements that passed. With its binary approach to compliance determination, the previous methodology failed to answer some important questions. In particular, how much energy cost could be saved by better compliance with the commercial energy code and what are the relative priorities of code requirements from an energy cost savings perspective? This paper explores an analytical approach and pilot study using a single building type and climate zone to answer those questions.
Thermal performance and heat transport in aquifer thermal energy storage
Sommer, W. T.; Doornenbal, P. J.; Drijver, B. C.; van Gaans, P. F. M.; Leusbrock, I.; Grotenhuis, J. T. C.; Rijnaarts, H. H. M.
2014-01-01
Aquifer thermal energy storage (ATES) is used for seasonal storage of large quantities of thermal energy. Due to the increasing demand for sustainable energy, the number of ATES systems has increased rapidly, which has raised questions on the effect of ATES systems on their surroundings as well as their thermal performance. Furthermore, the increasing density of systems generates concern regarding thermal interference between the wells of one system and between neighboring systems. An assessment is made of (1) the thermal storage performance, and (2) the heat transport around the wells of an existing ATES system in the Netherlands. Reconstruction of flow rates and injection and extraction temperatures from hourly logs of operational data from 2005 to 2012 show that the average thermal recovery is 82 % for cold storage and 68 % for heat storage. Subsurface heat transport is monitored using distributed temperature sensing. Although the measurements reveal unequal distribution of flow rate over different parts of the well screen and preferential flow due to aquifer heterogeneity, sufficient well spacing has avoided thermal interference. However, oversizing of well spacing may limit the number of systems that can be realized in an area and lower the potential of ATES.
High Efficiency and Low Cost Thermal Energy Storage System
Energy Technology Data Exchange (ETDEWEB)
Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Lv, Qiuping [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Moisseytsev, Anton [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Bucknor, Matthew [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division
2017-09-29
BgtL, LLC (BgtL) is focused on developing and commercializing its proprietary compact technology for processes in the energy sector. One such application is a compact high efficiency Thermal Energy Storage (TES) system that utilizes the heat of fusion through phase change between solid and liquid to store and release energy at high temperatures and incorporate state-of-the-art insulation to minimize heat dissipation. BgtL’s TES system would greatly improve the economics of existing nuclear and coal-fired power plants by allowing the power plant to store energy when power prices are low and sell power into the grid when prices are high. Compared to existing battery storage technology, BgtL’s novel thermal energy storage solution can be significantly less costly to acquire and maintain, does not have any waste or environmental emissions, and does not deteriorate over time; it can keep constant efficiency and operates cleanly and safely. BgtL’s engineers are experienced in this field and are able to design and engineer such a system to a specific power plant’s requirements. BgtL also has a strong manufacturing partner to fabricate the system such that it qualifies for an ASME code stamp. BgtL’s vision is to be the leading provider of compact systems for various applications including energy storage. BgtL requests that all technical information about the TES designs be protected as proprietary information. To honor that request, only non-proprietay summaries are included in this report.
International Nuclear Information System (INIS)
Arita, Yutaka; Kihara, Yuji; Mitsuhasi, Junichi; Niita, Koji; Takai, Mikio; Ogawa, Izumi; Kishimoto, Tadafumi; Yoshihara, Tsutomu
2007-01-01
The simulation of a thermal-neutron-induced single-event upset (SEU) was performed on a 0.4-μm-design-rule 4 Mbit static random access memory (SRAM) using particle and heavy-ion transport code system (PHITS): The SEU rates obtained by the simulation were in very good agreement with the result of experiments. PHITS is a useful tool for simulating SEUs in semiconductor devices. To further improve the accuracy of the simulation, additional methods for tallying the energy deposition are required for PHITS. (author)
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.
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
Energy Technology Data Exchange (ETDEWEB)
Menzel, Francine; Sabundjian, Gaianê, E-mail: franmenzel@gmail.com, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); D’Auria, Francesco, E-mail: f.dauria@ing.unipi.it [University of Pisa, San Piero a Grado Nuclear Research Group (Italy)
2017-07-01
Nuclear thermal hydraulic and accident analysis are based in three pillar activities, which consists in: Scaling, Coupling and V and V. Each of them are established technology, with key documents to describe and widely used. The final goal of this work is to apply the BEPU methodology in all parts of FSAR where analytical techniques are needed (BEPU-FSAR) and for that the crucial step is the transfer of the BEPU concepts into the other areas. In this sense, the issue is how to adapt to other disciplines the pillar activities presented in the thermal hydraulic area. For that we need to identify which elements can be applied in the other areas, to show that the proposed methodology is feasible. This work aims to discuss the first steps towards a BEPU-FSAR methodology and to show that the Scaling, Coupling and V and V elements, currently done for thermal-hydraulic codes, can be also done for different codes, which are used to perform different analysis included on a FSAR of a generic plant. (author)
Potential energy savings and thermal comfort
DEFF Research Database (Denmark)
Jensen, Karsten Ingerslev; Rudbeck, Claus Christian; Schultz, Jørgen Munthe
1996-01-01
The simulation results on the energy saving potential and influence on indoor thermal comfort by replacement of common windows with aerogel windows as well as commercial low-energy windows are described and analysed.......The simulation results on the energy saving potential and influence on indoor thermal comfort by replacement of common windows with aerogel windows as well as commercial low-energy windows are described and analysed....
Methodology for Evaluating Cost-effectiveness of Commercial Energy Code Changes
Energy Technology Data Exchange (ETDEWEB)
Hart, Philip R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Liu, Bing [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
2015-01-31
This document lays out the U.S. Department of Energy’s (DOE’s) method for evaluating the cost-effectiveness of energy code proposals and editions. The evaluation is applied to provisions or editions of the American Society of Heating, Refrigerating and Air-Conditioning Engineers (ASHRAE) Standard 90.1 and the International Energy Conservation Code (IECC). The method follows standard life-cycle cost (LCC) economic analysis procedures. Cost-effectiveness evaluation requires three steps: 1) evaluating the energy and energy cost savings of code changes, 2) evaluating the incremental and replacement costs related to the changes, and 3) determining the cost-effectiveness of energy code changes based on those costs and savings over time.
Building Standards and Codes for Energy Conservation
Gross, James G.; Pierlert, James H.
1977-01-01
Current activity intended to lead to energy conservation measures in building codes and standards is reviewed by members of the Office of Building Standards and Codes Services of the National Bureau of Standards. For journal availability see HE 508 931. (LBH)
Development of best estimate auditing code for CANDU thermal hydraulic safety analysis
Energy Technology Data Exchange (ETDEWEB)
Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1998-04-15
The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is first step of the whole project, thus focus to the establishment of improvement area. The study was performed by reconsideration of the previous code assessment works and investigation of AECL design analysis tools. In order to identify the thermal hydraulic phenomena for events, the whole system of CANDU plant was divided into main functional systems and subcomponents. Each phenomena was addressed to the each subcomponent. FinaIly improvement areas of model development for auditing tool were established based on the identified phenomena.
International Nuclear Information System (INIS)
2001-06-01
This report deals with an internationally agreed experimental test facility matrix for the validation of best estimate thermal-hydraulic computer codes applied for the analysis of VVER reactor primary systems in accident and transient conditions. Firstly, the main physical phenomena that occur during the considered accidents are identified, test types are specified, and test facilities that supplement the CSNI CCVMs and are 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. The construction of VVER Thermal-Hydraulic Code Validation Matrix follows the logic of the CSNI Code Validation Matrices (CCVM). Similar to the CCVM it is an attempt to collect together in a systematic way the best sets of available test data for VVER specific 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 in countries operating VVER reactors 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. (authors)
Solar applications of thermal energy storage. Final report
Energy Technology Data Exchange (ETDEWEB)
Lee, C.; Taylor, L.; DeVries, J.; Heibein, S.
1979-01-01
A technology assessment is presented on solar energy systems which use thermal energy storage. The study includes characterization of the current state-of-the-art of thermal energy storage, an assessment of the energy storage needs of solar energy systems, and the synthesis of this information into preliminary design criteria which would form the basis for detailed designs of thermal energy storage. (MHR)
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
Energy Technology Data Exchange (ETDEWEB)
Santos, Thiago A. dos; Maiorino, José R., E-mail: thiago.santos@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil); Stefanni, Giovanni L. de, E-mail: giovanni.stefanni@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)
2017-07-01
In order to project a nuclear reactor, the neutronic calculus must be validated, so that its thermal limits and safety parameters are respected. Considering this issue, this research aims to evaluate the APTh-100 reactor thermal limits. This PWR is a project developed in Universidade Federal do ABC (UFABC) using fuel composed of Uranium and Thorium oxide mixed (U,Th)O{sub 2}. For this purpose, a simplified, although conservative, code was developed in a MATLAB environment named STC-MOX-Th 'Simplified Thermal-hydraulics Code-Mixed Oxide Thorium'. This code provides axial and radial temperature distribution, as well as DNBR distribution over the hottest channel of the reactor core. Moreover, it brings other hydraulic quantities, such as pressure drop over the fuel rod, considering any fuel proportion of (U,Th)O{sub 2}.The software uses basic laws of conservation of mass, momentum and energy, it also calculates the thermal conduction equation, considering the thermal conductive coefficient as a temperature function. In order to solve this equation, the finite elements method was used. Furthermore, the proportion of 36% of UO{sub 2} was used to evaluate the temperature over the fuel rod and DNBR minimum in three burn conditions: beginning, middle and ending. The program has proven to be efficient in every condition and the results evidenced that the APTh-1000 reactor, in an initial analysis, has its thermal limits within the recommended security parameters. (author)
Optimization of thermal insulation to achieve energy savings in low energy house (refurbishment)
International Nuclear Information System (INIS)
Bojić, Milorad; Miletić, Marko; Bojić, Ljubiša
2014-01-01
Highlights: • For buildings that require heating, a thickness of their thermal insulation is optimized. • The objective was to improve energy efficiency of the building. • The optimization is performed by using EnergyPlus and Hooke–Jeeves method. • The embodied energy of thermal insulation and the entire life cycle of the house are taken into account. - Abstract: Due to the current environmental situation, saving energy and reducing CO 2 emission have become the leading drive in modern research. For buildings that require heating, one of the solutions is to optimize a thickness of their thermal insulation and thus improve energy efficiency and reduce energy needs. In this paper, for a small residential house in Serbia, an optimization in the thickness of its thermal insulation layer is investigated by using EnergyPlus software and Hooke–Jeeves direct search method. The embodied energy of thermal insulation is taken into account. The optimization is done for the entire life cycle of thermal insulation. The results show the optimal thickness of thermal insulation that yields the minimum primary energy consumption
WEC3: Wave Energy Converter Code Comparison Project: Preprint
Energy Technology Data Exchange (ETDEWEB)
Combourieu, Adrien; Lawson, Michael; Babarit, Aurelien; Ruehl, Kelley; Roy, Andre; Costello, Ronan; Laporte Weywada, Pauline; Bailey, Helen
2017-01-01
This paper describes the recently launched Wave Energy Converter Code Comparison (WEC3) project and present preliminary results from this effort. The objectives of WEC3 are to verify and validate numerical modelling tools that have been developed specifically to simulate wave energy conversion devices and to inform the upcoming IEA OES Annex VI Ocean Energy Modelling Verification and Validation project. WEC3 is divided into two phases. Phase 1 consists of a code-to-code verification and Phase II entails code-to-experiment validation. WEC3 focuses on mid-fidelity codes that simulate WECs using time-domain multibody dynamics methods to model device motions and hydrodynamic coefficients to model hydrodynamic forces. Consequently, high-fidelity numerical modelling tools, such as Navier-Stokes computational fluid dynamics simulation, and simple frequency domain modelling tools were not included in the WEC3 project.
Energy Technology Data Exchange (ETDEWEB)
Delbecq, J.M
1999-07-01
The Aster code is a 2D or 3D finite-element calculation code for structures developed by the R and D direction of Electricite de France (EdF). This dossier presents a complete overview of the characteristics and uses of the Aster code: introduction of version 4; the context of Aster (organisation of the code development, versions, systems and interfaces, development tools, quality assurance, independent validation); static mechanics (linear thermo-elasticity, Euler buckling, cables, Zarka-Casier method); non-linear mechanics (materials behaviour, big deformations, specific loads, unloading and loss of load proportionality indicators, global algorithm, contact and friction); rupture mechanics (G energy restitution level, restitution level in thermo-elasto-plasticity, 3D local energy restitution level, KI and KII stress intensity factors, calculation of limit loads for structures), specific treatments (fatigue, rupture, wear, error estimation); meshes and models (mesh generation, modeling, loads and boundary conditions, links between different modeling processes, resolution of linear systems, display of results etc..); vibration mechanics (modal and harmonic analysis, dynamics with shocks, direct transient dynamics, seismic analysis and aleatory dynamics, non-linear dynamics, dynamical sub-structuring); fluid-structure interactions (internal acoustics, mass, rigidity and damping); linear and non-linear thermal analysis; steels and metal industry (structure transformations); coupled problems (internal chaining, internal thermo-hydro-mechanical coupling, chaining with other codes); products and services. (J.S.)
International Nuclear Information System (INIS)
Hartmann, C.; Sanchez, V.; Tietsch, W.; Stieglitz, R.
2012-01-01
The KIT is involved in the development and qualification of best estimate methodologies for BWR transient analysis in cooperation with industrial partners. The goal is to establish the most advanced thermal hydraulic system codes coupled with 3D reactor dynamic codes to be able to perform a more realistic evaluation of the BWR behavior under accidental conditions. For this purpose a computational chain based on the lattice code (SCALE6/GenPMAXS), the coupled neutronic/thermal hydraulic code (TRACE/PARCS) as well as a Monte Carlo based uncertainty and sensitivity package (SUSA) has been established and applied to different kind of transients of a Boiling Water Reactor (BWR). This paper will describe the multidimensional models of the plant elaborated for TRACE and PARCS to perform the investigations mentioned before. For the uncertainty quantification of the coupled code TRACE/PARCS and specifically to take into account the influence of the kinetics parameters in such studies, the PARCS code has been extended to facilitate the change of model parameters in such a way that the SUSA package can be used in connection with TRACE/PARCS for the U and S studies. This approach will be presented in detail. The results obtained for a rod drop transient with TRACE/PARCS using the SUSA-methodology showed clearly the importance of some kinetic parameters on the transient progression demonstrating that the coupling of a best-estimate coupled codes with uncertainty and sensitivity tools is very promising and of great importance for the safety assessment of nuclear reactors. (authors)
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.
Thermal energy accumulators. A bibliographical study
International Nuclear Information System (INIS)
Charlety, Paul
1971-01-01
Energy storage is a challenge, notably for spacecraft, submarines and non-polluting automotive vehicles. After a comparison of mass energies of different principles of energy accumulation (magnetic, electrostatic, solid elasticity, kinetic energy, gaseous elasticity, electro-chemistry, sensitive heat, freezing heat, fuels, radioactivity, nuclear fission or fusion, mass energy), the author discusses the choice of thermal storage, presents the main bodies used for thermal energy accumulation (molten salts such as lithium hydride or lithium salt eutectics, or other compounds such as alumina, paraffins), and gives an overview of the main theoretical problems [fr
International Nuclear Information System (INIS)
Peng Muzhang; Zhang Quan; Wang Guoli; Zhang Yuman
1988-01-01
TISKTH-3 is a coupled neutronics/thermal-hydraulics code for the transient analysis. A 3-dimensional neutron kinetics equation solved by the Nodal Green's Function Method is used for the neutronics model of the code. A homogeneous equilibrium model with a complete boiling curve and two numerical solutions of the implicit and explicit scheme is used for the thermal-hydraulics model of the code. A 2-dimensional heat conduction equation with variable conductivity solved by the method of weighted residuals is used for the fuel rod heat transfer model of the code. TISKTH-3 is able to analyze the fast transient process and complicate accident situations in the core. The initative applications have shown that the stability and convergency in the calculations with the code are satisfactory
Energy Technology Data Exchange (ETDEWEB)
Muzhang, Peng; Quan, Zhang; Guoli, Wang; Yuman, Zhang
1988-03-01
TISKTH-3 is a coupled neutronics/thermal-hydraulics code for the transient analysis. A 3-dimensional neutron kinetics equation solved by the Nodal Green's Function Method is used for the neutronics model of the code. A homogeneous equilibrium model with a complete boiling curve and two numerical solutions of the implicit and explicit scheme is used for the thermal-hydraulics model of the code. A 2-dimensional heat conduction equation with variable conductivity solved by the method of weighted residuals is used for the fuel rod heat transfer model of the code. TISKTH-3 is able to analyze the fast transient process and complicate accident situations in the core. The initative applications have shown that the stability and convergency in the calculations with the code are satisfactory.
Best estimate LB LOCA approach based on advanced thermal-hydraulic codes
International Nuclear Information System (INIS)
Sauvage, J.Y.; Gandrille, J.L.; Gaurrand, M.; Rochwerger, D.; Thibaudeau, J.; Viloteau, E.
2004-01-01
Improvements achieved in thermal-hydraulics with development of Best Estimate computer codes, have led number of Safety Authorities to preconize realistic analyses instead of conservative calculations. The potentiality of a Best Estimate approach for the analysis of LOCAs urged FRAMATOME to early enter into the development with CEA and EDF of the 2nd generation code CATHARE, then of a LBLOCA BE methodology with BWNT following the Code Scaling Applicability and Uncertainty (CSAU) proceeding. CATHARE and TRAC are the basic tools for LOCA studies which will be performed by FRAMATOME according to either a deterministic better estimate (dbe) methodology or a Statistical Best Estimate (SBE) methodology. (author)
Gilbraith, Nathaniel; Azevedo, Inês L; Jaramillo, Paulina
2014-12-16
The federal government has the goal of decreasing commercial building energy consumption and pollutant emissions by incentivizing the adoption of commercial building energy codes. Quantitative estimates of code benefits at the state level that can inform the size and allocation of these incentives are not available. We estimate the state-level climate, environmental, and health benefits (i.e., social benefits) and reductions in energy bills (private benefits) of a more stringent code (ASHRAE 90.1-2010) relative to a baseline code (ASHRAE 90.1-2007). We find that reductions in site energy use intensity range from 93 MJ/m(2) of new construction per year (California) to 270 MJ/m(2) of new construction per year (North Dakota). Total annual benefits from more stringent codes total $506 million for all states, where $372 million are from reductions in energy bills, and $134 million are from social benefits. These total benefits range from $0.6 million in Wyoming to $49 million in Texas. Private benefits range from $0.38 per square meter in Washington State to $1.06 per square meter in New Hampshire. Social benefits range from $0.2 per square meter annually in California to $2.5 per square meter in Ohio. Reductions in human/environmental damages and future climate damages account for nearly equal shares of social benefits.
Evaluation of the RELAP4/MOD6 thermal-hydraulic code
International Nuclear Information System (INIS)
Haigh, W.S.; Margolis, S.G.; Rice, R.E.
1978-01-01
The NRC RELAP4/MOD6 computer code was recently released to the public for use in thermal-hydraulic analysis. This code has a unique new capability permitting analysis of both the blowdown and reflood portions of a postulated pressurized water reactor (PWR) loss-of-coolant accident (LOCA). A principal code evaluation objective is to assess the accuracy of the code for computing LOCA behavior over a wide range of system sizes and scaling concepts. The scales of interest include all LOCA experiments and will ultimately encompass full-sized PWR systems for which no experiments or data are available. Quantitative assessment of the accuracy of the code when it is applied to large PWR systems is still in the future. With RELAP4/MOD6, however, a technique has been demonstrated for using results derived from small-scale blowdown and reflood experiments to predict the accuracy of calculations for similar experiments of significantly different scale or component size. This demonstration is considered a first step in establishing confidence levels for the accuracy of calculations of a postulated LOCA
Compliance Verification Paths for Residential and Commercial Energy Codes
Energy Technology Data Exchange (ETDEWEB)
Conover, David R.; Makela, Eric J.; Fannin, Jerica D.; Sullivan, Robin S.
2011-10-10
This report looks at different ways to verify energy code compliance and to ensure that the energy efficiency goals of an adopted document are achieved. Conformity assessment is the body of work that ensures compliance, including activities that can ensure residential and commercial buildings satisfy energy codes and standards. This report identifies and discusses conformity-assessment activities and provides guidance for conducting assessments.
THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code
International Nuclear Information System (INIS)
Vondy, D.R.
1984-07-01
The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations
THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code
Energy Technology Data Exchange (ETDEWEB)
Vondy, D.R.
1984-07-01
The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.
THERMAL: A routine designed to calculate neutron thermal scattering. Revision 1
International Nuclear Information System (INIS)
Cullen, D.E.
1995-01-01
THERMAL is designed to calculate neutron thermal scattering that is elastic and isotropic in the center of mass system. At low energy thermal motion will be included. At high energies the target nuclei are assumed to be stationary. The point of transition between low and high energies has been defined to insure a smooth transition. It is assumed that at low energy the elastic cross section is constant in the relative system. At high energy the cross section can be of any form. You can use this routine for all energies where the elastic scattering is isotropic in the center of mass system. In most materials this will be a fairly high energy, e.g., the keV energy range. The THERMAL method is simple, clean, easy to understand, and most important very efficient; on a SUN SPARC-10 workstation, at low energies with thermal scattering it can do almost 6 million scatters a minute and at high energy over 13 million. Warning: This version of THERMAL completely supersedes the original version described in the same report number, dated February 24, 1995. The method used in the original code is incorrect, as explained in this report
Thyc, a 3D thermal-hydraulic code for rod bundles. Recent developments and validation tests
International Nuclear Information System (INIS)
Caremoli, C.; Rascle, P.; Aubry, S.; Olive, J.
1993-09-01
PWR or LMFBR cores or fuel assemblies, PWR steam generators, condensers, tubular heat exchangers, are basic components of a nuclear power plant involving two-phase flows in tube or rod bundles. A deep knowledge of the detailed flow patterns on the shell side is necessary to evaluate DNB margins in reactor cores, singularity effects (grids, wire spacers, support plates, baffles), corrosion on steam generator tube sheet, bypass effects and vibration risks. For that purpose, Electricite de France has developed, since 1986, a general purpose code named THYC (Thermal HYdraulic Code) designed to study three-dimensional single and two phase flows in rod or tube bundles (pressurized water reactor cores, steam generators, condensers, heat exchangers). It considers the three-dimensional domain to contain two kinds of components: fluid and solids. The THYC model is obtained by space-time averaging of the instantaneous equations (mass, momentum and energy) of each phase over control volumes including fluid and solids. This paper briefly presents the physical model and the numerical method used in THYC. Then, validation tests (comparison with experiments) and applications (coupling with three-dimensional neutronics code and DNB predictions) are presented. They emphasize the last developments and new capabilities of the code. (authors). 10 figs., 3 tabs., 21 refs
MIF-SCD computer code for thermal hydraulic calculation of supercritical water cooled reactor core
International Nuclear Information System (INIS)
Galina P Bogoslovskaia; Alexander A Karpenko; Pavel L Kirillov; Alexander P Sorokin
2005-01-01
Full text of publication follows: Supercritical pressure power plants constitute the basis of heat power engineering in many countries to day. Starting from a long-standing experience of their operation, it is proposed to develop a new type of fast breeder reactor cooled by supercritical water, which enables the economical indices of NPP to be substantially improved. In the Thermophysical Department of SSC RF-IPPE, an attempt is made to provide thermal-hydraulic validation of the reactor under discussion. The paper presents the results of analysis of the thermal-hydraulic characteristics of fuel subassemblies cooled by supercritical water based on subchannel analysis. Modification of subchannel code MIF - MIF-SCD Code - developed in the SSC RF IPPE is designed as block code and permits one to calculate the coolant temperature and velocity distributions in fuel subassembly channels, the temperature of fuel pin claddings and fuel subassembly wrapper under conditions of irregular geometry and non-uniform axial and radial power generation. The thermal hydraulics under supercritical pressure of water exhibits such peculiarities as abrupt variation of the thermal physical properties in the range of pseudo-critical temperature, the absence of such phenomenon as the critical heat flux which can lead to fuel element burnout in WWERs. As compared with subchannel code for light water, in order to take account of the variation of the coolant properties versus temperature in more detail, a block for evaluating the thermal physical properties of supercritical water versus the local coolant temperature in the fuel subassembly channels was added. The peculiarities of the geometry and power generation in the fuel subassembly of the supercritical reactor are considered as well in special blocks. The results of calculations have shown that considerable preheating of supercritical coolant (several hundreds degrees) can occur in the fuel subassembly. The test calculations according to
The Cost of Enforcing Building Energy Codes: Phase 2
Energy Technology Data Exchange (ETDEWEB)
Williams, Alison [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Price, Sarah K. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Vine, Ed [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)
2014-10-15
The purpose of this study is to present key findings regarding costs associated with enforcing building energy code compliance–primarily focusing on costs borne by local government. Building codes, if complied with, have the ability to save a significant amount of energy. However, energy code compliance rates have been significantly lower than 100%. Renewed interest in building energy codes has focused efforts on increasing compliance, particularly as a result of the 2009 American Recovery and Reinvestment Act (ARRA) requirement that in order for states to receive additional energy grants, they must have “a plan for the jurisdiction achieving compliance with the building energy code…in at least 90 percent of new and renovated residential and commercial building space” by 2017 (Public Law 111-5, Section 410(2)(C)). One study by the Institute for Market Transformation (IMT) estimated the costs associated with reaching 90% compliance to be $810 million, or $610 million in additional funding over existing expenditures, a non-trivial value. [Majersik & Stellberg 2010] In this context, Lawrence Berkeley National Laboratory (LBNL) conducted a study to better pinpoint the costs of enforcement through a two-phase process.
COMTA - a computer code for fuel mechanical and thermal analysis
International Nuclear Information System (INIS)
Basu, S.; Sawhney, S.S.; Anand, A.K.; Anantharaman, K.; Mehta, S.K.
1979-01-01
COMTA is a generalized computer code for integrity analysis of the free standing fuel cladding, with natural UO 2 or mixed oxide fuel pellets. Thermal and Mechanical analysis is done simultaneously for any power history of the fuel pin. For analysis, the fuel cladding is assumed to be axisymmetric and is subjected to axisymmetric load due to contact pressure, gas pressure, coolant pressure and thermal loads. Axial variation of load is neglected and creep and plasticity are assumed to occur at constant volume. The pellet is assumed to be made of concentric annuli. The fission gas release integral is dependent on the temperature and the power produced in each annulus. To calculate the temperature distribution in the fuel pin, the variation of bulk coolant temperature is given as an input to the code. Gap conductance is calculated at every time step, considering fuel densification, fuel relocation and gap closure, filler gas dilution by released fission gas, gap closure by expansion and irradiation swelling. Overall gap conductance is contributed by heat transfer due to the three modes; conduction convection and radiation as per modified Ross and Stoute model. Equilibrium equations, compatibility equations, stress strain relationships (including thermal strains and permanent strains due to creep and plasticity) are used to obtain triaxial stresses and strains. Thermal strain is assumed to be zero at hot zero power conditions. The boundary conditions are obtained for radial stresses at outside and inside surfaces by making these equal to coolant pressure and internal pressure respectively. A multi-mechanism creep model which accounts for thermal and irradiation creep is used to calculate the overall creep rate. Effective plastic strain is a function of effective stress and material constants. (orig.)
FLICA-4 (version 1) a computer code for three dimensional thermal analysis of nuclear reactor cores
International Nuclear Information System (INIS)
Raymond, P.; Allaire, G.; Boudsocq, G.
1995-01-01
FLICA-4 is a thermal-hydraulic computer code developed at the French Energy Atomic Commission (CEA) for three dimensional steady state or transient two phase flow for design and safety thermal analysis of nuclear reactor cores. The two phase flow model of FLICA-4 is based on four balance equations for the fluid which includes: three balance equations for the mixture and a mass balance equation for the less concentrated phase which permits the calculation of non-equilibrium flows as sub cooled boiling and superheated steam. A drift velocity model takes into account the velocity disequilibrium between phases. The thermal behaviour of fuel elements can be computed by a one dimensional heat conduction equation in plane, cylindrical or spherical geometries and coupled to the fluid flow calculation. Convection and diffusion of solution products which are transported either by the liquid or by the gas, can be evaluated by solving specific mass conservation equations. A one dimensional two phase flow model can also be used to compute 1-D flow in pipes, guide tubes, BWR assemblies or RBMK channels. The FLICA-4 computer code uses fast running time steam-water functions. Phasic and saturation physical properties are computed by using bi-cubic spline functions. Polynomial coefficients are tabulated from 0.1 to 22 MPa and 0 to 800 degrees C. Specific modules can be utilised in order to generate the spline coefficients for any other fluid properties
Hybrid energy harvesting using active thermal backplane
Kim, Hyun-Wook; Lee, Dong-Gun
2016-04-01
In this study, we demonstrate the concept of a new hybrid energy harvesting system by combing solar cells with magneto-thermoelectric generator (MTG, i.e., thermal energy harvesting). The silicon solar cell can easily reach high temperature under normal operating conditions. Thus the heated solar cell becomes rapidly less efficient as the temperature of solar cell rises. To increase the efficiency of the solar cell, air or water-based cooling system is used. To surpass conventional cooling devices requiring additional power as well as large working space for air/water collectors, we develop a new technology of pairing an active thermal backplane (ATB) to solar cell. The ATB design is based on MTG technology utilizing the physics of the 2nd order phase transition of active ferromagnetic materials. The MTG is cost-effective conversion of thermal energy to electrical energy and is fundamentally different from Seebeck TEG devices. The ATB (MTG) is in addition to being an energy conversion system, a very good conveyor of heat through both conduction and convection. Therefore, the ATB can provide dual-mode for the proposed hybrid energy harvesting. One is active convective and conductive cooling for heated solar cell. Another is active thermal energy harvesting from heat of solar cell. These novel hybrid energy harvesting device have potentially simultaneous energy conversion capability of solar and thermal energy into electricity. The results presented can be used for better understanding of hybrid energy harvesting system that can be integrated into commercial applications.
Graphene Thermal Properties: Applications in Thermal Management and Energy Storage
Directory of Open Access Journals (Sweden)
Jackie D. Renteria
2014-11-01
Full Text Available We review the thermal properties of graphene, few-layer graphene and graphene nanoribbons, and discuss practical applications of graphene in thermal management and energy storage. The first part of the review describes the state-of-the-art in the graphene thermal field focusing on recently reported experimental and theoretical data for heat conduction in graphene and graphene nanoribbons. The effects of the sample size, shape, quality, strain distribution, isotope composition, and point-defect concentration are included in the summary. The second part of the review outlines thermal properties of graphene-enhanced phase change materials used in energy storage. It is shown that the use of liquid-phase-exfoliated graphene as filler material in phase change materials is promising for thermal management of high-power-density battery parks. The reported experimental and modeling results indicate that graphene has the potential to outperform metal nanoparticles, carbon nanotubes, and other carbon allotropes as filler in thermal management materials.
Thermal Distribution System | Energy Systems Integration Facility | NREL
Thermal Distribution System Thermal Distribution System The Energy Systems Integration Facility's . Photo of the roof of the Energy Systems Integration Facility. The thermal distribution bus allows low as 10% of its full load level). The 60-ton chiller cools water with continuous thermal control
Basiri, H.; Tavakoli-Anbaran, H.
2018-01-01
Am-Be neutrons source is based on (α, n) reaction and generates neutrons in the energy range of 0-11 MeV. Since the thermal neutrons are widely used in different fields, in this work, we investigate how to improve the source configuration in order to increase the thermal flux. These suggested changes include a spherical moderator instead of common cylindrical geometry, a reflector layer and an appropriate materials selection in order to achieve the maximum thermal flux. All calculations were done by using MCNP1 Monte Carlo code. Our final results indicated that a spherical paraffin moderator, a layer of beryllium as a reflector can efficiently increase the thermal neutron flux of Am-Be source.
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. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
2003-04-15
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 first step of the 3 year project, and the main researches were focused on identifying the candidate thermal hydraulic models for improvement and to develop prototypical model development. During the current year, the verification calculations submitted for the APR 1400 design certification have been reviewed, the experimental data from the MIDAS DVI experiment facility in KAERI have been analyzed and evaluated, candidate thermal hydraulic models for improvement have been identified, prototypical models for the improved thermal hydraulic models have been developed, items for experiment in connection with the model development have been identified, and preliminary design of the experiment has been carried out.
Development of thermal hydraulic models for the reliable regulatory auditing code
International Nuclear Information System (INIS)
Chung, B. D.; Song, C. H.; Lee, Y. J.; Kwon, T. S.
2003-04-01
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 first step of the 3 year project, and the main researches were focused on identifying the candidate thermal hydraulic models for improvement and to develop prototypical model development. During the current year, the verification calculations submitted for the APR 1400 design certification have been reviewed, the experimental data from the MIDAS DVI experiment facility in KAERI have been analyzed and evaluated, candidate thermal hydraulic models for improvement have been identified, prototypical models for the improved thermal hydraulic models have been developed, items for experiment in connection with the model development have been identified, and preliminary design of the experiment has been carried out
International Nuclear Information System (INIS)
Bera, S.; Pradhan, S.K.; Dubey, S.K.; Gupta, S.K.
2011-01-01
In general safety analyses of design basis accident of NPPs are being carried out using system thermal hydraulics code like RELAP. In RELAP, power is calculated based on point kinetics approximation, which virtually ignores the space and energy dependence of neutron flux. To include the space and energy dependence of neutron flux, three-dimensional neutronics code TRIHEXFA has been externally coupled with RELAP through interface program, TRIHEXFA-RELAP Interface Program (TRIP). Calculation methodology of TRIP program is based on adiabatic approximation. In the adiabatic approximation the neutron flux is being factored into spatial and amplitude part. Spatial part of flux is slowly varying with time whereas amplitude part is strongly varying function. The RELAP controls the transient time steps. Transient time is divided into several major and minor time steps. Minor time step is the sub-step of major time step. Thermal hydraulics and neutronics data are exchanged at each major time step. Spatial part of neutron flux has been updated at each major time step using TRIHEXFA code. But amplitude part of the neutron flux is calculated at each minor time step using RELAP code. Convergence of results of the coupled code, TRIP has been checked through coupling time step descritization study. This study determines the minimum coupling time step. Transient concerning VVER-1000 Main Steam Line Break, MSLB has been considered to investigate the space-time effect on point kinetics. MSLB occurs as a consequence of the rupture of one steam line upstream of main steam line isolation valves. Reference design and data from Kudankulam Nuclear Power Plant (KK-NPP) are used for the analysis. From this investigation it is found that TRIP significantly overestimates the maximum reactor power against uncoupled RELAP result. The time of scram also occur six seconds earlier in TRIP calculation compared to the RELAP. This exercise has also shown a proof of principle that coupling 3D
Solar Thermal Energy; Energia Solar Termica
Energy Technology Data Exchange (ETDEWEB)
Perez-Martinez, M; Cuesta-Santianes, M J; Cabrera Jimenez, J A
2008-07-01
Approximately, 50 % of worldwide primary energy consumption is done in the form of heat in applications with a temperature lower than 250 degree centigree (low-medium temperature heat). These data clearly demonstrate the great potential of solar thermal energy to substitute conventional fossil fuels, which are becoming more expensive and are responsible for global warming. Low-medium temperature solar thermal energy is mainly used to obtain domestic hot water and provide space heating. Active solar thermal systems are those related to the use of solar thermal collectors. This study is dealing with low temperature solar thermal applications, mainly focusing on active solar thermal systems. This kind of systems has been extensively growing worldwide during the last years. At the end of 2006, the collector capacity in operation worldwide equalled 127.8 GWth. The technology is considered to be already developed and actions should be aimed at favouring a greater market penetration: diffusion, financial support, regulations establishment, etc. China and USA are the leading countries with a technology based on evacuated tube collectors and unglazed collectors, respectively. The rest of the world markets are dominated by the flat glazed collectors technology. (Author) 15 refs.
[Thermal energy utilization analysis and energy conservation measures of fluidized bed dryer].
Xing, Liming; Zhao, Zhengsheng
2012-07-01
To propose measures for enhancing thermal energy utilization by analyzing drying process and operation principle of fluidized bed dryers,in order to guide optimization and upgrade of fluidized bed drying equipment. Through a systematic analysis on drying process and operation principle of fluidized beds,the energy conservation law was adopted to calculate thermal energy of dryers. The thermal energy of fluidized bed dryers is mainly used to make up for thermal consumption of water evaporation (Qw), hot air from outlet equipment (Qe), thermal consumption for heating and drying wet materials (Qm) and heat dissipation to surroundings through hot air pipelines and cyclone separators. Effective measures and major approaches to enhance thermal energy utilization of fluidized bed dryers were to reduce exhaust gas out by the loss of heat Qe, recycle dryer export air quantity of heat, preserve heat for dry towers, hot air pipes and cyclone separators, dehumidify clean air in inlets and reasonably control drying time and air temperature. Such technical parameters such air supply rate, air inlet temperature and humidity, material temperature and outlet temperature and humidity are set and controlled to effectively save energy during the drying process and reduce the production cost.
An intercomparison of medium energy cross-section codes
International Nuclear Information System (INIS)
Pearlstein, S.
1988-05-01
Five medium energy proton reaction cases are selected for benchmarking nuclear model codes. The quantities calculated are isotopic activation yields for 180 MeV protons on Al and 40-200 MeV protons on Co, and double differential neutron emission spectra from Al, Zr-90 and Pb-208 for 35, 80, 160, 318, and 800 presented consist of three types: a closed form preequilibrium plus evaporation model, an intranuclear-cascade and evaporation model, and a model relying on nuclear systematics. The characteristics of each code are described. There are orders of magnitude differences in the time for each type of code to calculate neutron emission spectra, with codes using systematics, preequilibrium and intranuclear-cascade models requiring seconds, minutes and hours, respectively. Calculations are not compared with experiment in this initial study. For double differential neutron emission spectra, there is good overall agreement in magnitude among the different types of codes at forward angles. Differences where they occur at forward angles are greatest for the mid-energy neutrons emitted. At back angles the incident energy at which the best overall agreement is obtained is 160 MeV and the material for which the best overall agreement is obtained is Al. 4 refs., 7 tabs
Wang, Song
2017-05-10
Thermal diodes, or devices that transport thermal energy asymmetrically, analogous to electrical diodes, hold promise for thermal energy harvesting and conservation, as well as for phononics or information processing. The junction of a phase change material and phase invariant material can form a thermal diode; however, there are limited constituent materials available for a given target temperature, particularly near ambient. In this work, we demonstrate that a micro and nanoporous polystyrene foam can house a paraffin-based phase change material, fused to PMMA, to produce mechanically robust, solid-state thermal diodes capable of ambient operation with Young\\'s moduli larger than 11.5 MPa and 55.2 MPa above and below the melting transition point, respectively. Moreover, the composites show significant changes in thermal conductivity above and below the melting point of the constituent paraffin and rectification that is well-described by our previous theory and the Maxwell–Eucken model. Maximum thermal rectifications range from 1.18 to 1.34. We show that such devices perform reliably enough to operate in thermal diode bridges, dynamic thermal circuits capable of transforming oscillating temperature inputs into single polarity temperature differences – analogous to an electrical diode bridge with widespread implications for transient thermal energy harvesting and conservation. Overall, our approach yields mechanically robust, solid-state thermal diodes capable of engineering design from a mathematical model of phase change and thermal transport, with implications for energy harvesting.
Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system
International Nuclear Information System (INIS)
Iga, Kiminori; Takada, Hiroshi; Nagao, Tadashi.
1998-01-01
In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B 4 C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)
Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system
Energy Technology Data Exchange (ETDEWEB)
Iga, Kiminori [Kyushu Univ., Fukuoka (Japan); Takada, Hiroshi; Nagao, Tadashi
1998-01-01
In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B{sub 4}C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)
Enhancing radiative energy transfer through thermal extraction
Tan, Yixuan; Liu, Baoan; Shen, Sheng; Yu, Zongfu
2016-06-01
Thermal radiation plays an increasingly important role in many emerging energy technologies, such as thermophotovoltaics, passive radiative cooling and wearable cooling clothes [1]. One of the fundamental constraints in thermal radiation is the Stefan-Boltzmann law, which limits the maximum power of far-field radiation to P0 = σT4S, where σ is the Boltzmann constant, S and T are the area and the temperature of the emitter, respectively (Fig. 1a). In order to overcome this limit, it has been shown that near-field radiations could have an energy density that is orders of magnitude greater than the Stefan-Boltzmann law [2-7]. Unfortunately, such near-field radiation transfer is spatially confined and cannot carry radiative heat to the far field. Recently, a new concept of thermal extraction was proposed [8] to enhance far-field thermal emission, which, conceptually, operates on a principle similar to oil immersion lenses and light extraction in light-emitting diodes using solid immersion lens to increase light output [62].Thermal extraction allows a blackbody to radiate more energy to the far field than the apparent limit of the Stefan-Boltzmann law without breaking the second law of thermodynamics. Thermal extraction works by using a specially designed thermal extractor to convert and guide the near-field energy to the far field, as shown in Fig. 1b. The same blackbody as shown in Fig. 1a is placed closely below the thermal extractor with a spacing smaller than the thermal wavelength. The near-field coupling transfers radiative energy with a density greater than σT4. The thermal extractor, made from transparent and high-index or structured materials, does not emit or absorb any radiation. It transforms the near-field energy and sends it toward the far field. As a result, the total amount of far-field radiative heat dissipated by the same blackbody is greatly enhanced above SσT4, where S is the area of the emitter. This paper will review the progress in thermal
Development of disruption thermal analysis code DREAM
Energy Technology Data Exchange (ETDEWEB)
Yamazaki, Seiichiro; Kobayahsi, Takeshi [Kawasaki Heavy Industries Ltd., Kobe (Japan); Seki, Masahiro
1989-07-01
When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing componenets such as first wall and divertor/limiter are subjected to a intensse heat load in a short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs. It causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes and radiation heat loss in required in the design of these components. This paper describes the computer code DREAM, developed to perform the disruption thermal analysis, taking phase changes and radiation into account. (author).
Development of disruption thermal analysis code DREAM
International Nuclear Information System (INIS)
Yamazaki, Seiichiro; Kobayahsi, Takeshi; Seki, Masahiro.
1989-01-01
When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing componenets such as first wall and divertor/limiter are subjected to a intensse heat load in a short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs. It causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes and radiation heat loss in required in the design of these components. This paper describes the computer code DREAM, developed to perform the disruption thermal analysis, taking phase changes and radiation into account. (author)
Thermal energy storage in granular deposits
Ratuszny, Paweł
2017-10-01
Energy storage technology is crucial for the development of the use of renewable energy sources. This is a substantial constraint, however it can, to some extent, be solved by storing energy in its various forms: electrical, mechanical, chemical and thermal. This article presents the results of research in thermal properties of granular deposits. Correlation between temperature changes in the stores over a period of time and their physical properties has been studied. The results of the research have practical application in designing thermal stores based on bulk materials and ground deposits. Furthermore, the research results are significant for regeneration of the lower ground sources for heat pumps and provide data for designing ground heat exchangers for ventilation systems.
High energy particle transport code NMTC/JAM
International Nuclear Information System (INIS)
Niita, Koji; Meigo, Shin-ichiro; Takada, Hiroshi; Ikeda, Yujiro
2001-03-01
We have developed a high energy particle transport code NMTC/JAM, which is an upgraded version of NMTC/JAERI97. The applicable energy range of NMTC/JAM is extended in principle up to 200 GeV for nucleons and mesons by introducing the high energy nuclear reaction code JAM for the intra-nuclear cascade part. For the evaporation and fission process, we have also implemented a new model, GEM, by which the light nucleus production from the excited residual nucleus can be described. According to the extension of the applicable energy, we have upgraded the nucleon-nucleus non-elastic, elastic and differential elastic cross section data by employing new systematics. In addition, the particle transport in a magnetic field has been implemented for the beam transport calculations. In this upgrade, some new tally functions are added and the format of input of data has been improved very much in a user friendly manner. Due to the implementation of these new calculation functions and utilities, consequently, NMTC/JAM enables us to carry out reliable neutronics study of a large scale target system with complex geometry more accurately and easily than before. This report serves as a user manual of the code. (author)
Application of the French codes to the pressurized thermal shocks assessment
Energy Technology Data Exchange (ETDEWEB)
Chen, Mingya; Wang, Rong Shan; Yu, Weiwei; Lu, Feng; Zhang, Guo Dong; Xue, Fei; Chen, Zhilin [Suzhou Nuclear Power Research Institute, Life Management Center, Suzhou (China); Qian, Guian [Paul Scherrer Institute, Nuclear Energy and Safety Department, Villigen (Switzerland); Shi, Jinhua [Amec Foster Wheeler, Clean Energy Department, Gloucester (United Kingdom)
2016-12-15
The integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the 'screening criterion' for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no 'screening criterion'. In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF) may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed.
Application of the French Codes to the Pressurized Thermal Shocks Assessment
Directory of Open Access Journals (Sweden)
Mingya Chen
2016-12-01
Full Text Available The integrity of a reactor pressure vessel (RPV related to pressurized thermal shocks (PTSs has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the “screening criterion” for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no “screening criterion”. In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed.
Application of the French codes to the pressurized thermal shocks assessment
International Nuclear Information System (INIS)
Chen, Mingya; Wang, Rong Shan; Yu, Weiwei; Lu, Feng; Zhang, Guo Dong; Xue, Fei; Chen, Zhilin; Qian, Guian; Shi, Jinhua
2016-01-01
The integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the 'screening criterion' for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no 'screening criterion'. In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF) may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed
HETFIS: High-Energy Nucleon-Meson Transport Code with Fission
International Nuclear Information System (INIS)
Barish, J.; Gabriel, T.A.; Alsmiller, F.S.; Alsmiller, R.G. Jr.
1981-07-01
A model that includes fission for predicting particle production spectra from medium-energy nucleon and pion collisions with nuclei (Z greater than or equal to 91) has been incorporated into the nucleon-meson transport code, HETC. This report is primarily concerned with the programming aspects of HETFIS (High-Energy Nucleon-Meson Transport Code with Fission). A description of the program data and instructions for operating the code are given. HETFIS is written in FORTRAN IV for the IBM computers and is readily adaptable to other systems
Summary of papers on current and anticipated uses of thermal-hydraulic codes
Energy Technology Data Exchange (ETDEWEB)
Caruso, R.
1997-07-01
The author reviews a range of recent papers which discuss possible uses and future development needs for thermal/hydraulic codes in the nuclear industry. From this review, eight common recommendations are extracted. They are: improve the user interface so that more people can use the code, so that models are easier and less expensive to prepare and maintain, and so that the results are scrutable; design the code so that it can easily be coupled to other codes, such as core physics, containment, fission product behaviour during severe accidents; improve the numerical methods to make the code more robust and especially faster running, particularly for low pressure transients; ensure that future code development includes assessment of code uncertainties as integral part of code verification and validation; provide extensive user guidelines or structure the code so that the `user effect` is minimized; include the capability to model multiple fluids (gas and liquid phase); design the code in a modular fashion so that new models can be added easily; provide the ability to include detailed or simplified component models; build on work previously done with other codes (RETRAN, RELAP, TRAC, CATHARE) and other code validation efforts (CSAU, CSNI SET and IET matrices).
Summary of papers on current and anticipated uses of thermal-hydraulic codes
International Nuclear Information System (INIS)
Caruso, R.
1997-01-01
The author reviews a range of recent papers which discuss possible uses and future development needs for thermal/hydraulic codes in the nuclear industry. From this review, eight common recommendations are extracted. They are: improve the user interface so that more people can use the code, so that models are easier and less expensive to prepare and maintain, and so that the results are scrutable; design the code so that it can easily be coupled to other codes, such as core physics, containment, fission product behaviour during severe accidents; improve the numerical methods to make the code more robust and especially faster running, particularly for low pressure transients; ensure that future code development includes assessment of code uncertainties as integral part of code verification and validation; provide extensive user guidelines or structure the code so that the 'user effect' is minimized; include the capability to model multiple fluids (gas and liquid phase); design the code in a modular fashion so that new models can be added easily; provide the ability to include detailed or simplified component models; build on work previously done with other codes (RETRAN, RELAP, TRAC, CATHARE) and other code validation efforts (CSAU, CSNI SET and IET matrices)
Energy-Efficient Channel Coding Strategy for Underwater Acoustic Networks
Directory of Open Access Journals (Sweden)
Grasielli Barreto
2017-03-01
Full Text Available Underwater acoustic networks (UAN allow for efficiently exploiting and monitoring the sub-aquatic environment. These networks are characterized by long propagation delays, error-prone channels and half-duplex communication. In this paper, we address the problem of energy-efficient communication through the use of optimized channel coding parameters. We consider a two-layer encoding scheme employing forward error correction (FEC codes and fountain codes (FC for UAN scenarios without feedback channels. We model and evaluate the energy consumption of different channel coding schemes for a K-distributed multipath channel. The parameters of the FEC encoding layer are optimized by selecting the optimal error correction capability and the code block size. The results show the best parameter choice as a function of the link distance and received signal-to-noise ratio.
SAFE: A computer code for the steady-state and transient thermal analysis of LMR fuel elements
International Nuclear Information System (INIS)
Hayes, S.L.
1993-12-01
SAFE is a computer code developed for both the steady-state and transient thermal analysis of single LMR fuel elements. The code employs a two-dimensional control-volume based finite difference methodology with fully implicit time marching to calculate the temperatures throughout a fuel element and its associated coolant channel for both the steady-state and transient events. The code makes no structural calculations or predictions whatsoever. It does, however, accept as input structural parameters within the fuel such as the distributions of porosity and fuel composition, as well as heat generation, to allow a thermal analysis to be performed on a user-specified fuel structure. The code was developed with ease of use in mind. An interactive input file generator and material property correlations internal to the code are available to expedite analyses using SAFE. This report serves as a complete design description of the code as well as a user's manual. A sample calculation made with SAFE is included to highlight some of the code's features. Complete input and output files for the sample problem are provided
Metal hydride-based thermal energy storage systems
Vajo, John J.; Fang, Zhigang
2017-10-03
The invention provides a thermal energy storage system comprising a metal-containing first material with a thermal energy storage density of about 1300 kJ/kg to about 2200 kJ/kg based on hydrogenation; a metal-containing second material with a thermal energy storage density of about 200 kJ/kg to about 1000 kJ/kg based on hydrogenation; and a hydrogen conduit for reversibly transporting hydrogen between the first material and the second material. At a temperature of 20.degree. C. and in 1 hour, at least 90% of the metal is converted to the hydride. At a temperature of 0.degree. C. and in 1 hour, at least 90% of the metal hydride is converted to the metal and hydrogen. The disclosed metal hydride materials have a combination of thermodynamic energy storage densities and kinetic power capabilities that previously have not been demonstrated. This performance enables practical use of thermal energy storage systems for electric vehicle heating and cooling.
Theoretical Thermal Evaluation of Energy Recovery Incinerators
1985-12-01
Army Logistics Mgt Center, Fort Lee , VA DTIC Alexandria, VA DTNSRDC Code 4111 (R. Gierich), Bethesda MD; Code 4120, Annapolis, MD; Code 522 (Library...Washington. DC: Code (I6H4. Washington. DC NAVSECGRUACT PWO (Code .’^O.’^). Winter Harbor. IVIE ; PWO (Code 4(1). Edzell. Scotland; PWO. Adak AK...NEW YORK Fort Schuyler. NY (Longobardi) TEXAS A&M UNIVERSITY W.B. Ledbetter College Station. TX UNIVERSITY OF CALIFORNIA Energy Engineer. Davis CA
A code to study the water flow in a thermal test loop
International Nuclear Information System (INIS)
Saunier, Jean-Pierre; Duffourt, Nicole; Lago, Bernard
1965-01-01
A first part reports the theoretical and analytical formulation of a flow within a specific circuit used in a thermal test installation. Equations in the different parts of the circuit are developed, and their resolution for integration into a computation code is described, including boundary conditions, constants and input functions (cell characteristics, fluid characteristics, heat transfer, friction, time slicing). The second part reports an extension of this theoretical and analytical development and code development to a two-branch circuit
Qualification of code-Saturne for thermal-hydraulics single phase nuclear applications
International Nuclear Information System (INIS)
Archambeau, F.; Bechaud, C.; Gest, B.; Martin, A.; Sakiz, M.
2003-01-01
Code-Saturne is a general finite volume CFD (computational fluid dynamics) code developed by Electricite de France (EDF) under quality assurance for 2- and 3-dimensional simulations, laminar and turbulent flows, conjugate heat transfer (coupling with thermal code SYRTHES), including combustion modelling and a Lagrangian module. A very large range of meshes can be used. The solver relies on a finite volume method on arbitrary meshes (hybrid, with hanging nodes, any type of element). All variables are located at the cell centres. The solver is time marching, with a predictor-corrector scheme for Navier-Stokes equations. Standard Reynolds Average Navier-Stokes modelling (RANS) is included (k-epsilon, RSM). Code-Saturne is used by EDF in various industrial fields such as process engineering, aeraulics, combustion and nuclear applications. The present paper describes the qualification phase carried out during 2001 for single-phase nuclear applications. Indeed, once an industrial product has been released and validated, it is of major importance, especially in this particular field related to safety matters, to demonstrate the ability of the code to help engineers produce satisfactory conclusions to industrial problems. In coherence with analyses and best practice guidelines such as those published by the ERCOFTAC Special Interest Group, it seemed important to base the qualification phase on well defined and documented experimental facilities, sufficiently complex to be representative of industrial studies. Much attention has been devoted to evaluating sensitivity to numerical parameters such as grid refinement, time step... Moreover, the qualification studies have been carried out in real-life conditions, that is in limited time, with industrial limitations on the number of grid cells, and by the teams usually producing such studies, so as to integrate a real industrial process in the qualification phase. Two test cases chosen to assess certain types of flows in PWR
Khattab, K; Sulieman, I
2009-04-01
The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutron fluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method ((197)Au (n, gamma) (198)Au and (59)Co (n, gamma) (60)Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutron flux and the epithermal index in each site. The calculated and measured results agree well.
Energy efficient thermal management of data centers
Kumar, Pramod
2012-01-01
Energy Efficient Thermal Management of Data Centers examines energy flow in today's data centers. Particular focus is given to the state-of-the-art thermal management and thermal design approaches now being implemented across the multiple length scales involved. The impact of future trends in information technology hardware, and emerging software paradigms such as cloud computing and virtualization, on thermal management are also addressed. The book explores computational and experimental characterization approaches for determining temperature and air flow patterns within data centers. Thermodynamic analyses using the second law to improve energy efficiency are introduced and used in proposing improvements in cooling methodologies. Reduced-order modeling and robust multi-objective design of next generation data centers are discussed. This book also: Provides in-depth treatment of energy efficiency ideas based on fundamental heat transfer, fluid mechanics, thermodynamics, controls, and computer science Focus...
Policy Pathways: Modernising Building Energy Codes
Energy Technology Data Exchange (ETDEWEB)
NONE
2013-08-01
Buildings are the largest consumers of energy worldwide and will continue to be a source of increasing energy demand in the future. Globally, the sector’s final energy consumption doubled between 1971 and 2010 to reach 2 794 million tonnes of oil equivalent (Mtoe), driven primarily by population increase and economic growth. Under current policies, the global energy demand of buildings is projected by the IEA experts to grow by an additional 838 Mtoe by 2035 compared to 2010. The challenges of the projected increase of energy consumption due to the built environment vary by country. In IEA member countries, much of the future buildings stock is already in place, and so the main challenge is to renovate existing buildings stock. In non-IEA countries, more than half of the buildings stock needed by 2050 has yet to be built. The IEA and the UNDP partnered to analyse current practices in the design and implementation of building energy codes. The aim is to consolidate existing efforts and to encourage more attention to the role of the built environment in a low-carbon and climate-resilient world. This joint IEA-UNDP Policy Pathway aims to share lessons learned between IEA member countries and non-IEA countries. The objective is to spread best practices, limit pressures on global energy supply, improve energy security, and contribute to environmental sustainability. Part of the IEA Policy Pathway series, Modernising building energy codes to secure our global energy future sets out key steps in planning, implementation, monitoring and evaluation. The Policy Pathway series aims to help policy makers implement the IEA 25 Energy Efficiency Policy Recommendations endorsed by IEA Ministers (2011).
International Nuclear Information System (INIS)
Griggs, D.P.; Kazimi, M.S.; Henry, A.F.
1982-01-01
The initial development of TITAN, a three-dimensional coupled neutronics/thermal-hydraulics code for LWR safety analysis, has been completed. The transient neutronics code QUANDRY has been joined to the two-fluid thermal-hydraulics code THERMIT with the appropriate feedback mechanisms modeled. A detailed steady-state and transient coupling scheme based on the tandem technique was implemented in accordance with the important structural and operational characteristics of QUANDRY and THERMIT. A two channel sample problem formed the basis for steady-state and transient analyses performed with TITAN. TITAN steady-state results were compared with those obtained with MEKIN and showed good agreement. Null transients, simulated turbine trip transients, and a rod withdrawal transient were analyzed with TITAN and reasonable results were obtained
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
Energy Technology Data Exchange (ETDEWEB)
Matausek, M V [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)
1966-06-15
A combination of multigroup method and P{sub 3} approximation of spherical harmonics method was chosen for calculating space-energy distribution of thermal neutron flux in elementary reactor cell. Application of these methods reduced solution of complicated transport equation to the problem of solving an inhomogeneous system of six ordinary firs-order differential equations. A procedure is proposed which avoids numerical solution and enables analytical solution when applying certain approximations. Based on this approach, computer codes were written for ZUSE-Z-23 computer: SIGMA code for calculating group constants for a given material; MULTI code which uses results of SIGMA code as input and calculates spatial ana energy distribution of thermal neutron flux in a reactor cell. Calculations of thermal neutron spectra for a number of reactor cells were compared to results available from literature. Agreement was satisfactory in all the cases, which proved the correctness of the applied method. Some possibilities for improving the precision and acceleration of the calculation process were found during calculation. (author)
Energy Technology Data Exchange (ETDEWEB)
Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Escalante, Javier Jimenez; Espinoza, Victor Sanchez
2015-07-15
Highlights: • Simulation of BFBT-BWR steady-state and transient tests with ATHLET. • Validation of thermal-hydraulic models based on pressure drops and void fraction measurements. • TRACE system code is used for the comparative study. • Predictions result in a good agreement with the experiments. • Discrepancies are smaller or comparable with respect to the measurements uncertainty. - Abstract: Validation and qualification of thermal-hydraulic system codes based on separate effect tests are essential for the reliability of numerical tools when applied to nuclear power plant analyses. To this purpose, the Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is involved in various validation and qualification activities of different CFD, sub-channel and system codes. In this paper, the capabilities of the thermal-hydraulic code ATHLET are assessed based on the experimental results provided within the NUPEC BFBT benchmark related to key Boiling Water Reactors (BWR) phenomena. Void fraction and pressure drops measurements in the BFBT bundle performed under steady-state and transient conditions which are representative for e.g. turbine trip and recirculation pump trip events, are compared with the numerical results of ATHLET. The comparison of code predictions with the BFBT data has shown good agreement given the experimental uncertainty and the results are consistent with the trends obtained with similar thermal-hydraulic codes.
Concrete thermal energy storage for steam generation
DEFF Research Database (Denmark)
Singh, Shobhana; Sørensen, Kim
2017-01-01
Establishing enhancement methods to develop cost-effective thermal energy storage technology requires a detailed analysis. In this paper, a numerical investigation of the concrete based thermal energy storage system is carried out. The storage system consists of a heat transfer fluid flowing inside...
International Nuclear Information System (INIS)
Dickson, T.L.
1993-01-01
This report discusses probabilistic fracture mechanics (PFM) analysis which is a major element of the comprehensive probabilistic methodology endorsed by the NRC for evaluation of the integrity of Pressurized Water Reactor (PWR) pressure vessels subjected to pressurized-thermal-shock (PTS) transients. It is anticipated that there will be an increasing need for an improved and validated PTS PFM code which is accepted by the NRC and utilities, as more plants approach the PTS screening criteria and are required to perform plant-specific analyses. The NRC funded Heavy Section Steel Technology (HSST) Program at Oak Ridge National Laboratories is currently developing the FAVOR (Fracture Analysis of Vessels: Oak Ridge) PTS PFM code, which is intended to meet this need. The FAVOR code incorporates the most important features of both OCA-P and VISA-II and contains some new capabilities such as PFM global modeling methodology, the capability to approximate the effects of thermal streaming on circumferential flaws located inside a plume region created by fluid and thermal stratification, a library of stress intensity factor influence coefficients, generated by the NQA-1 certified ABAQUS computer code, for an adequate range of two and three dimensional inside surface flaws, the flexibility to generate a variety of output reports, and user friendliness
Energy Technology Data Exchange (ETDEWEB)
Peniguel, C [Electricite de France (EDF), 78 - Chatou (France). Direction des Etudes et Recherches; Rupp, I [Simulog, 78 (France)
1998-12-31
Thermal aspects take place in several industrial applications in which Electricite de France (EdF) is concerned. In most cases, several physical phenomena like conduction, radiation and convection are involved in thermal transfers. The aim of this paper is to present a numerical tool adapted to industrial configurations and which uses the coupling between fluid convection (resolved with ESTET in finite-volumes or with N3S in finite-elements) and radiant heat transfers between walls (resolved with SYRTHES using a radiosity method). SYRTHES manages the different thermal exchanges that can occur between fluid and solid domains thanks to an explicit iterative method. An extension of SYRTHES has been developed which allows to take into account simultaneously several fluid codes using `message passing` computer tools like Parallel Virtual Machine (PVM) and the code coupling software CALCIUM developed by the Direction of Studies and Researches (DER) of EdF. Various examples illustrate the interest of such a numerical tool. (J.S.) 12 refs.
Energy Technology Data Exchange (ETDEWEB)
Peniguel, C. [Electricite de France (EDF), 78 - Chatou (France). Direction des Etudes et Recherches; Rupp, I. [Simulog, 78 (France)
1997-12-31
Thermal aspects take place in several industrial applications in which Electricite de France (EdF) is concerned. In most cases, several physical phenomena like conduction, radiation and convection are involved in thermal transfers. The aim of this paper is to present a numerical tool adapted to industrial configurations and which uses the coupling between fluid convection (resolved with ESTET in finite-volumes or with N3S in finite-elements) and radiant heat transfers between walls (resolved with SYRTHES using a radiosity method). SYRTHES manages the different thermal exchanges that can occur between fluid and solid domains thanks to an explicit iterative method. An extension of SYRTHES has been developed which allows to take into account simultaneously several fluid codes using `message passing` computer tools like Parallel Virtual Machine (PVM) and the code coupling software CALCIUM developed by the Direction of Studies and Researches (DER) of EdF. Various examples illustrate the interest of such a numerical tool. (J.S.) 12 refs.
Current and anticipated uses of thermal-hydraulic codes in Spain
Energy Technology Data Exchange (ETDEWEB)
Pelayo, F.; Reventos, F. [Consejo de Seguridad Nuclear, Barcelona (Spain)
1997-07-01
Spanish activities in the field of Applied Thermal-Hydraulics are steadily increasing as the codes are becoming practicable enough to efficiently sustain engineering decision in the Nuclear Power industry. Before reaching this point, a lot of effort has been devoted to achieve this goal. This paper briefly describes this process, points at the current applications and draws conclusions on the limitations. Finally it establishes the applications where the use of T-H codes would be worth in the future, this in turn implies further development of the codes to widen the scope of application and improve the general performance. Due to the different uses of the codes, the applications mainly come from the authority, industry, universities and research institutions. The main conclusion derived from this paper establishes that further code development is justified if the following requisites are considered: (1) Safety relevance of scenarios not presently covered is established. (2) A substantial gain in margins or the capability to use realistic assumptions is obtained. (3) A general consensus on the licensability and methodology for application is reached. The role of Regulatory Body is stressed, as the most relevant outcome of the project may be related to the evolution of the licensing frame.
Experimental Determination of in Situ Utilization of Lunar Regolith for Thermal Energy Storage
Richter, Scott W.
1993-01-01
A Lunar Thermal Energy from Regolith (LUTHER) experiment has been designed and fabricated at the NASA Lewis Research Center to determine the feasibility of using lunar soil as thermal energy storage media. The experimental apparatus includes an alumina ceramic canister (25.4 cm diameter by 45.7 cm length) which contains simulated lunar regolith, a heater (either radiative or conductive), 9 heat shields, a heat transfer cold jacket, and 19 type B platinum rhodium thermocouples. The simulated lunar regolith is a basalt, mined and processed by the University of Minnesota, that closely resembles the lunar basalt returned to earth by the Apollo missions. The experiment will test the effects of vacuum, particle size, and density on the thermophysical properties of the regolith. The properties include melt temperature (range), specific heat, thermal conductivity, and latent heat of storage. Two separate tests, using two different heaters, will be performed to study the effect of heating the system using radiative and conductive heat transfer. The physical characteristics of the melt pattern, material compatibility of the molten regolith, and the volatile gas emission will be investigated by heating a portion of the lunar regolith to its melting temperature (1435 K) in a 10(exp -4) pascal vacuum chamber, equipped with a gas spectrum analyzer. A finite differencing SINDA model was developed at NASA Lewis Research Center to predict the performance of the LUTHER experiment. The analytical results of the code will be compared with the experimental data generated by the LUTHER experiment. The code will predict the effects of vacuum, particle size, and density has on the heat transfer to the simulated regolith.
Uncertainties in modeling and scaling in the prediction of fuel stored energy and thermal response
International Nuclear Information System (INIS)
Wulff, W.
1987-01-01
The steady-state temperature distribution and the stored energy in nuclear fuel elements are computed by analytical methods and used to rank, in the order of importance, the effects on stored energy from statistical uncertainties in modeling parameters, in boundary and in operating conditions. An integral technique is used to calculate the transient fuel temperature and to estimate the uncertainties in predicting the fuel thermal response and the peak clad temperature during a large-break loss of coolant accident. The uncertainty analysis presented here is an important part of evaluating the applicability, the uncertainties and the scaling capabilities of computer codes for nuclear reactor safety analyses. The methods employed in this analysis merit general attention because of their simplicity. It is shown that the blowdown peak is dominated by fuel stored energy alone or, equivalently, by linear heating rate. Gap conductance, peaking factors and fuel thermal conductivity are the three most important fuel modeling parameters affecting peak clad temperature uncertainty. 26 refs., 10 figs., 6 tabs
Implementation of CFD module in the KORSAR thermal-hydraulic system code
Energy Technology Data Exchange (ETDEWEB)
Yudov, Yury V.; Danilov, Ilia G.; Chepilko, Stepan S. [Alexandrov Research Inst. of Technology (NITI), Sosnovy Bor (Russian Federation)
2015-09-15
The Russian KORSAR/GP (hereinafter KORSAR) computer code was developed by a joint team from Alexandrov NITI and OKB ''Gidropress'' for VVER safety analysis and certified by the Rostechnadzor of Russia in 2009. The code functionality is based on a 1D two-fluid model for calculation of two-phase flows. A 3D CFD module in the KORSAR computer code is being developed by Alexandrov NITI for representing 3D effects in the downcomer and lower plenum during asymmetrical loop operation. The CFD module uses Cartesian grid method with cut cell approach. The paper presents a numerical algorithm for coupling 1D and 3D thermal- hydraulic modules in the KORSAR code. The combined pressure field is calculated by the multigrid method. The performance efficiency of the algorithm for coupling 1D and 3D modules was demonstrated by solving the benchmark problem of mixing cold and hot flows in a T-junction.
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)
A computerized energy systems code and information library at Soreq
Energy Technology Data Exchange (ETDEWEB)
Silverman, I; Shapira, M; Caner, D; Sapier, D [Israel Atomic Energy Commission, Yavne (Israel). Soreq Nuclear Research Center
1996-12-01
In the framework of the contractual agreement between the Ministry of Energy and Infrastructure and the Division of Nuclear Engineering of the Israel Atomic Energy Commission, both Soreq-NRC and Ben-Gurion University have agreed to establish, in 1991, a code center. This code center contains a library of computer codes and relevant data, with particular emphasis on nuclear power plant research and development support. The code center maintains existing computer codes and adapts them to the ever changing computing environment, keeps track of new code developments in the field of nuclear engineering, and acquires the most recent revisions of computer codes of interest. An attempt is made to collect relevant codes developed in Israel and to assure that proper documentation and application instructions are available. En addition to computer programs, the code center collects sample problems and international benchmarks to verify the codes and their applications to various areas of interest to nuclear power plant engineering and safety evaluation. Recently, the reactor simulation group at Soreq acquired, using funds provided by the Ministry of Energy and Infrastructure, a PC work station operating under a Linux operating system to give users of the library an easy on-line way to access resources available at the library. These resources include the computer codes and their documentation, reports published by the reactor simulation group, and other information databases available at Soreq. Registered users set a communication line, through a modem, between their computer and the new workstation at Soreq and use it to download codes and/or information or to solve their problems, using codes from the library, on the computer at Soreq (authors).
A computerized energy systems code and information library at Soreq
International Nuclear Information System (INIS)
Silverman, I.; Shapira, M.; Caner, D.; Sapier, D.
1996-01-01
In the framework of the contractual agreement between the Ministry of Energy and Infrastructure and the Division of Nuclear Engineering of the Israel Atomic Energy Commission, both Soreq-NRC and Ben-Gurion University have agreed to establish, in 1991, a code center. This code center contains a library of computer codes and relevant data, with particular emphasis on nuclear power plant research and development support. The code center maintains existing computer codes and adapts them to the ever changing computing environment, keeps track of new code developments in the field of nuclear engineering, and acquires the most recent revisions of computer codes of interest. An attempt is made to collect relevant codes developed in Israel and to assure that proper documentation and application instructions are available. En addition to computer programs, the code center collects sample problems and international benchmarks to verify the codes and their applications to various areas of interest to nuclear power plant engineering and safety evaluation. Recently, the reactor simulation group at Soreq acquired, using funds provided by the Ministry of Energy and Infrastructure, a PC work station operating under a Linux operating system to give users of the library an easy on-line way to access resources available at the library. These resources include the computer codes and their documentation, reports published by the reactor simulation group, and other information databases available at Soreq. Registered users set a communication line, through a modem, between their computer and the new workstation at Soreq and use it to download codes and/or information or to solve their problems, using codes from the library, on the computer at Soreq (authors)
Validation of a thermal-hydraulic system code on a simple example
International Nuclear Information System (INIS)
Kopecek, Vit; Zacha, Pavel
2014-01-01
A mathematical model of a U tube was set up and the analytical solution was calculated and used in the assessment of the numerical solutions obtained by using the RELAP5 mod3.3 and TRACE V5 thermal hydraulics codes. A good agreement between the 2 types of calculation was obtained.
Smart Building: Decision Making Architecture for Thermal Energy Management.
Uribe, Oscar Hernández; Martin, Juan Pablo San; Garcia-Alegre, María C; Santos, Matilde; Guinea, Domingo
2015-10-30
Smart applications of the Internet of Things are improving the performance of buildings, reducing energy demand. Local and smart networks, soft computing methodologies, machine intelligence algorithms and pervasive sensors are some of the basics of energy optimization strategies developed for the benefit of environmental sustainability and user comfort. This work presents a distributed sensor-processor-communication decision-making architecture to improve the acquisition, storage and transfer of thermal energy in buildings. The developed system is implemented in a near Zero-Energy Building (nZEB) prototype equipped with a built-in thermal solar collector, where optical properties are analysed; a low enthalpy geothermal accumulation system, segmented in different temperature zones; and an envelope that includes a dynamic thermal barrier. An intelligent control of this dynamic thermal barrier is applied to reduce the thermal energy demand (heating and cooling) caused by daily and seasonal weather variations. Simulations and experimental results are presented to highlight the nZEB thermal energy reduction.
Smart Building: Decision Making Architecture for Thermal Energy Management
Directory of Open Access Journals (Sweden)
Oscar Hernández Uribe
2015-10-01
Full Text Available Smart applications of the Internet of Things are improving the performance of buildings, reducing energy demand. Local and smart networks, soft computing methodologies, machine intelligence algorithms and pervasive sensors are some of the basics of energy optimization strategies developed for the benefit of environmental sustainability and user comfort. This work presents a distributed sensor-processor-communication decision-making architecture to improve the acquisition, storage and transfer of thermal energy in buildings. The developed system is implemented in a near Zero-Energy Building (nZEB prototype equipped with a built-in thermal solar collector, where optical properties are analysed; a low enthalpy geothermal accumulation system, segmented in different temperature zones; and an envelope that includes a dynamic thermal barrier. An intelligent control of this dynamic thermal barrier is applied to reduce the thermal energy demand (heating and cooling caused by daily and seasonal weather variations. Simulations and experimental results are presented to highlight the nZEB thermal energy reduction.
On the Non-Thermal Energy Content of Cosmic Structures
Directory of Open Access Journals (Sweden)
Franco Vazza
2016-11-01
Full Text Available (1 Background: the budget of non-thermal energy in galaxy clusters is not well constrained, owing to the observational and theoretical difficulties in studying these diluted plasmas on large scales; (2 Method: we use recent cosmological simulations with complex physics in order to connect the emergence of non-thermal energy to the underlying evolution of gas and dark matter; (3 Results: the impact of non-thermal energy (e.g., cosmic rays, magnetic fields and turbulent motions is found to increase in the outer region of galaxy clusters. Within numerical and theoretical uncertainties, turbulent motions dominate the budget of non-thermal energy in most of the cosmic volume; (4 Conclusion: assessing the distribution non-thermal energy in galaxy clusters is crucial to perform high-precision cosmology in the future. Constraining the level of non-thermal energy in cluster outskirts will improve our understanding of the acceleration of relativistic particles and of the origin of extragalactic magnetic fields.
Guide to Setting Thermal Comfort Criteria and Minimizing Energy Use in Delivering Thermal Comfort
Energy Technology Data Exchange (ETDEWEB)
Regnier, Cindy [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)
2012-08-01
Historically thermal comfort in buildings has been controlled by simple dry bulb temperature settings. As we move into more sophisticated low energy building systems that make use of alternate systems such as natural ventilation, mixed mode system and radiant thermal conditioning strategies, a more complete understanding of human comfort is needed for both design and control. This guide will support building designers, owners, operators and other stakeholders in defining quantifiable thermal comfort parameters?these can be used to support design, energy analysis and the evaluation of the thermal comfort benefits of design strategies. This guide also contains information that building owners and operators will find helpful for understanding the core concepts of thermal comfort. Whether for one building, or for a portfolio of buildings, this guide will also assist owners and designers in how to identify the mechanisms of thermal comfort and space conditioning strategies most important for their building and climate, and provide guidance towards low energy design options and operations that can successfully address thermal comfort. An example of low energy design options for thermal comfort is presented in some detail for cooling, while the fundamentals to follow a similar approach for heating are presented.
Photoswitchable Molecular Rings for Solar-Thermal Energy Storage.
Durgun, E; Grossman, Jeffrey C
2013-03-21
Solar-thermal fuels reversibly store solar energy in the chemical bonds of molecules by photoconversion, and can release this stored energy in the form of heat upon activation. Many conventional photoswichable molecules could be considered as solar thermal fuels, although they suffer from low energy density or short lifetime in the photoinduced high-energy metastable state, rendering their practical use unfeasible. We present a new approach to the design of chemistries for solar thermal fuel applications, wherein well-known photoswitchable molecules are connected by different linker agents to form molecular rings. This approach allows for a significant increase in both the amount of stored energy per molecule and the stability of the fuels. Our results suggest a range of possibilities for tuning the energy density and thermal stability as a function of the type of the photoswitchable molecule, the ring size, or the type of linkers.
Heat pipe solar receiver with thermal energy storage
Zimmerman, W. F.
1981-01-01
An HPSR Stirling engine generator system featuring latent heat thermal energy storge, excellent thermal stability and self regulating, effective thermal transport at low system delta T is described. The system was supported by component technology testing of heat pipes and of thermal storage and energy transport models which define the expected performance of the system. Preliminary and detailed design efforts were completed and manufacturing of HPSR components has begun.
Application of an accurate thermal hydraulics solver in VTT's reactor dynamics codes
International Nuclear Information System (INIS)
Rajamaeki, M.; Raety, H.; Kyrki-Rajamaeki, R.; Eskola, M.
1998-01-01
VTT's reactor dynamics codes are developed further and new more detailed models are created for tasks related to increased safety requirements. For thermal hydraulics calculations an accurate general flow model based on a new solution method PLIM has been developed. It has been applied in VTT's one-dimensional TRAB and three-dimensional HEXTRAN codes. Results of a demanding international boron dilution benchmark defined by VTT are given and compared against results of other codes with original or improved boron tracking. The new PLIM method not only allows the accurate modelling of a propagating boron dilution front, but also the tracking of a temperature front, which is missed by the special boron tracking models. (orig.)
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
User's manual for computer code RIBD-II, a fission product inventory code
International Nuclear Information System (INIS)
Marr, D.R.
1975-01-01
The computer code RIBD-II is used to calculate inventories, activities, decay powers, and energy releases for the fission products generated in a fuel irradiation. Changes from the earlier RIBD code are: the expansion to include up to 850 fission product isotopes, input in the user-oriented NAMELIST format, and run-time choice of fuels from an extensively enlarged library of nuclear data. The library that is included in the code package contains yield data for 818 fission product isotopes for each of fourteen different fissionable isotopes, together with fission product transmutation cross sections for fast and thermal systems. Calculational algorithms are little changed from those in RIBD. (U.S.)
Energy Technology Data Exchange (ETDEWEB)
Akerstream, T. [Manitoba Hydro, Winnipeg, MB (Canada); Allard, K. [City of Thompson, Thompson, MB (Canada); Anderson, N.; Beacham, D. [Manitoba Office of the Fire Commissioner, Winnipeg, MB (Canada); Andrich, R. [The Forks North Portage Partnership, MB (Canada); Auger, A. [Natural Resources Canada, Ottawa, ON (Canada). Office of Energy Efficiency; Downs, R.G. [Shindico Realty Inc., Winnipeg, MB (Canada); Eastwood, R. [Number Ten Architectural Group, Winnipeg, MB (Canada); Hewitt, C. [SMS Engineering Ltd., Winnipeg, MB (Canada); Joshi, D. [City of Winnipeg, Winnipeg, MB (Canada); Klassen, K. [Manitoba Dept. of Energy Science and Technology, Winnipeg, MB (Canada); Phillips, B. [Unies Ltd., Winnipeg, MB (Canada); Wiebe, R. [Ben Wiebe Construction Ltd., Winnipeg, MB (Canada); Woelk, D. [Bockstael Construction Ltd., Winnipeg, MB (Canada); Ziemski, S. [CREIT Management LLP, Winnipeg, MB (Canada)
2006-09-15
This report presented a strategy and a set of recommendations for the adoption, development and implementation of an energy code for new commercial construction in Manitoba. The report was compiled by an advisory committee comprised of industry representatives and government agency representatives. Recommendations were divided into 4 categories: (1) advisory committee recommendations; (2) code adoption recommendations; (3) code development recommendations; and (4) code implementation recommendations. It was suggested that Manitoba should adopt an amended version of the Model National Energy Code for Buildings (1997) as a regulation under the Buildings and Mobile Homes Act. Participation in a national initiative to update the Model National Energy Code for Buildings was also advised. It was suggested that the energy code should be considered as the first step in a longer-term process towards a sustainable commercial building code. However, the code should be adopted within the context of a complete market transformation approach. Other recommendations included: the establishment of a multi-stakeholder energy code task group; the provision of information and technical resources to help build industry capacity; the establishment of a process for energy code compliance; and an ongoing review of the energy code to assess impacts and progress. Supplemental recommendations for future discussion included the need for integrated design by building design teams in Manitoba; the development of a program to provide technical assistance to building design teams; and collaboration between post-secondary institutions to develop and deliver courses on integrated building design to students and professionals. 17 refs.
Parallel Computing Characteristics of Two-Phase Thermal-Hydraulics code, CUPID
International Nuclear Information System (INIS)
Lee, Jae Ryong; Yoon, Han Young
2013-01-01
Parallelized CUPID code has proved to be able to reproduce multi-dimensional thermal hydraulic analysis by validating with various conceptual problems and experimental data. In this paper, the characteristics of the parallelized CUPID code were investigated. Both single- and two phase simulation are taken into account. Since the scalability of a parallel simulation is known to be better for fine mesh system, two types of mesh system are considered. In addition, the dependency of the preconditioner for matrix solver was also compared. The scalability for the single-phase flow is better than that for two-phase flow due to the less numbers of iterations for solving pressure matrix. The CUPID code was investigated the parallel performance in terms of scalability. The CUPID code was parallelized with domain decomposition method. The MPI library was adopted to communicate the information at the interface cells. As increasing the number of mesh, the scalability is improved. For a given mesh, single-phase flow simulation with diagonal preconditioner shows the best speedup. However, for the two-phase flow simulation, the ILU preconditioner is recommended since it reduces the overall simulation time
International Nuclear Information System (INIS)
Mohr, C.L.; Lanning, D.D.; Panisko, F.E.
1979-01-01
The fuel performance code GAPCON-THERMAL-3 has been expanded to include recent transient material deformation constitutive relations and the FLECHT heat transfer correlation. The modifications make it possible to compute the thermal and mechanical response of nuclear fuel to postulated Loss of Coolant Accidents (LOCA). The numerical formulation has the capability of predicting both steady state and transient behavior of a fuel rod using a single analytical procedure. GAPCON-THERMAL-3 (G-T-3) uses a specialized finite element procedure for mechanics predictions and the method of weighted residuals and finite difference techniques to compute temperature and thermal behavior. Fuel behavior, gas release models, gas conductance models, and stored energy calculations are applicable to both steady state and transient conditions. The code has been used to perform scoping analysis for in-reactor LOCA simulation testing. (orig.)
Energy Savings Analysis of the Proposed Revision of the Washington D.C. Non-Residential Energy Code
Energy Technology Data Exchange (ETDEWEB)
Rosenberg, Michael I.; Athalye, Rahul A.; Hart, Philip R.
2017-12-01
This report presents the results of an assessment of savings for the proposed Washington D.C. energy code relative to ASHRAE Standard 90.1-2010. It includes annual and life cycle savings for site energy, source energy, energy cost, and carbon dioxide emissions that would result from adoption and enforcement of the proposed code for newly constructed buildings in Washington D.C. over a five year period.
A Comprehensive Review of Thermal Energy Storage
Directory of Open Access Journals (Sweden)
Ioan Sarbu
2018-01-01
Full Text Available Thermal energy storage (TES is a technology that stocks thermal energy by heating or cooling a storage medium so that the stored energy can be used at a later time for heating and cooling applications and power generation. TES systems are used particularly in buildings and in industrial processes. This paper is focused on TES technologies that provide a way of valorizing solar heat and reducing the energy demand of buildings. The principles of several energy storage methods and calculation of storage capacities are described. Sensible heat storage technologies, including water tank, underground, and packed-bed storage methods, are briefly reviewed. Additionally, latent-heat storage systems associated with phase-change materials for use in solar heating/cooling of buildings, solar water heating, heat-pump systems, and concentrating solar power plants as well as thermo-chemical storage are discussed. Finally, cool thermal energy storage is also briefly reviewed and outstanding information on the performance and costs of TES systems are included.
Energy Technology Data Exchange (ETDEWEB)
Laustsen, Jens
2008-03-15
The aim of this paper is to describe and analyse current approaches to encourage energy efficiency in building codes for new buildings. Based on this analysis the paper enumerates policy recommendations for enhancing how energy efficiency is addressed in building codes and other policies for new buildings. This paper forms part of the IEA work for the G8 Gleneagles Plan of Action. These recommendations reflect the study of different policy options for increasing energy efficiency in new buildings and examination of other energy efficiency requirements in standards or building codes, such as energy efficiency requirements by major renovation or refurbishment. In many countries, energy efficiency of buildings falls under the jurisdiction of the federal states. Different standards cover different regions or climatic conditions and different types of buildings, such as residential or simple buildings, commercial buildings and more complicated high-rise buildings. There are many different building codes in the world and the intention of this paper is not to cover all codes on each level in all countries. Instead, the paper details different regions of the world and different ways of standards. In this paper we also evaluate good practices based on local traditions. This project does not seek to identify one best practice amongst the building codes and standards. Instead, different types of codes and different parts of the regulation have been illustrated together with examples on how they have been successfully addressed. To complement this discussion of efficiency standards, this study illustrates how energy efficiency can be improved through such initiatives as efficiency labelling or certification, very best practice buildings with extremely low- or no-energy consumption and other policies to raise buildings' energy efficiency beyond minimum requirements. When referring to the energy saving potentials for buildings, this study uses the analysis of recent IEA
Stars with shell energy sources. Part 1. Special evolutionary code
International Nuclear Information System (INIS)
Rozyczka, M.
1977-01-01
A new version of the Henyey-type stellar evolution code is described and tested. It is shown, as a by-product of the tests, that the thermal time scale of the core of a red giant approaching the helium flash is of the order of the evolutionary time scale. The code itself appears to be a very efficient tool for investigations of the helium flash, carbon flash and the evolution of a white dwarf accreting mass. (author)
78 FR 55245 - Activities and Methodology for Assessing Compliance With Building Energy Codes
2013-09-10
...-0036] Activities and Methodology for Assessing Compliance With Building Energy Codes AGENCY: Office of... available to states to evaluate compliance with building energy codes and general approaches towards... building energy codes and general approaches towards compliance. The comment period ended on September 5...
Development of thermal hydraulic analysis code for IHX of FBR
International Nuclear Information System (INIS)
Kumagai, Hiromichi; Naohara, Nobuyuki
1991-01-01
In order to obtain flow resistance correlations for thermal-hydrauric analysis code concerned with an intermediate heat exchanger (IHX) of FBR, the hydraulic experiment by air was carried out through a bundle of tubes arranged in an in-line and staggard fashion. The main results are summarized as follows. (1) On pressure loss per unit length of a tube bundle, which is densely a regular triangle arrangement, the in-line fashion is almost the same as the staggard one. (2) In case of 30deg sector model for IHX tube bundle, pressure loss is 1/3 in comparison with the in-line or staggard arrangement. (3) By this experimental data, flow resistance correlations for thermalhydrauric analysis code are obtained. (author)
Validation of the Thermal-Hydraulic Model in the SACAP Code with the ISP Tests
Energy Technology Data Exchange (ETDEWEB)
Park, Soon-Ho; Kim, Dong-Min; Park, Chang-Hwan [FNC Technology Co., Yongin (Korea, Republic of)
2016-10-15
In safety viewpoint, the pressure of the containment is the important parameter, of course, the local hydrogen concentration is also the parameter of the major concern because of its flammability and the risk of the detonation. In Korea, there have been an extensive efforts to develop the computer code which can analyze the severe accident behavior of the pressurized water reactor. The development has been done in a modularized manner and SACAP(Severe Accident Containment Analysis Package) code is now under final stage of development. SACAP code adopts LP(Lumped Parameter) model and is applicable to analyze the synthetic behavior of the containment during severe accident occurred by thermal-hydraulic transient, combustible gas burn, direct containment heating by high pressure melt ejection, steam explosion and molten core-concrete interaction. The analyses of a number of ISP(International Standard Problem) experiments were done as a part of the SACAP code V and V(verification and validation). In this paper, the SACAP analysis results for ISP-35 NUPEC and ISP-47 TOSQAN are presented including comparison with other existing NPP simulation codes. In this paper, we selected and analyzed ISP-35 NUPEC, ISP-47 TOSQAN in order to confirm the computational performance of SACAP code currently under development. Now the multi-node analysis for the ISP-47 is under process. As a result of simulation, SACAP predicts well the thermal-hydraulic variables such as temperature, pressure, etc. Also, we verify that SACAP code is properly equipped to analyze the gas distribution and condensation.
International Nuclear Information System (INIS)
Mori, Takamasa; Nakagawa, Masayuki; Kaneko, Kunio.
1996-05-01
A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author)
Energy Technology Data Exchange (ETDEWEB)
Mori, Takamasa; Nakagawa, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio
1996-05-01
A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author).
International Nuclear Information System (INIS)
Spitz, P.B.; Lemoine, F.; Tirini, S.
1997-01-01
IPSN (Nuclear Protection and Safety Institute) and GRS (Gesellschaft fur Anlagen und Reaktorsicherheit Schwertnergasse 1) are developing the ESCADRE-ASTEC systems of codes devoted to the prediction of the behaviour of water-cooled reactors during a severe accident. The RALOC-mod 4 code belongs to this system and is specifically devoted to containment thermal-hydraulics studies. IPSN has designed a Thermal Hydraulic Containment Test Program in support to the Phebus Fission Product Test Program/2/. Evaporation tests have been recently performed in the Phebus containment test facility. The objective of this work is to assess against these tests the capability of the RALOC -mod 4 code to capture the phenomena observed in these experiments and more particularly the evaporation heat transfer and wall heat transfers. (DM)
International Nuclear Information System (INIS)
Guo, Y.; Wang, G.; Cheng, Y.; Peng, C.
2015-01-01
Water Cooled Blanket (WCB) is very important in the concept design and energy transfer in future fusion power plant. One concept design of WCB is under computational testing. RELAP5 and MELCOR codes, which are mature and often used in nuclear engineering, are selected as simulation tools. The complex inner flow channels and heat sources are simplified according to its thermal-hydraulic characteristics. Then the nodal models for RELAP5 and MELCOR are built for approximating the concept design. The superheated steam scheme is analyzed by two codes separately under different power levels. After some adjustments of the inlet flow resistance coefficients of some flow channels, the reasonable stable conditions can be obtained. The stable fluid and wall temperature distributions and pressure drops are studied. The results of two codes are compared and some advices are given. (authors)
Aquifer thermal energy storage in Finland
Energy Technology Data Exchange (ETDEWEB)
Iihola, H; Ala-Peijari, T; Seppaenen, H
1988-01-01
The rapid changes and crises in the field of energy during the 1970s and 1980s have forced us to examine the use of energy more critically and to look for new ideas. Seasonal aquifer thermal energy storage (T < 100/sup 0/C) on a large scale is one of the grey areas which have not yet been extensively explored. However, projects are currently underway in a dozen countries. In Finland there have been three demonstration projects from 1974 to 1987. International co-operation under the auspices of the International Energy Agency, Annex VI, 'Environmental and Chemical Aspects of Thermal Energy Storage in Aquifers and Research and Development of Water Treatment Methods' started in 1987. The research being undertaken in 8 countries includes several elements fundamental to hydrochemistry and biochemistry.
Needs of thermal-hydraulic codes for analyzing hydrogen behavior of future chinese NPPs
International Nuclear Information System (INIS)
Zhiwei Zhou; Jianjun Xiao; Mengjia Yang
2005-01-01
severe accident management guidelines are therefore needed for dealing with both the in-vessel and ex-vessel phenomena, including hydrogen generation, diffusion/convection and deflagration/detonation. To develop the sophisticated thermalhydraulic codes for analyzing severe accident related hydrogen behavior of a light water reactor system is quite expensive and rather unrealistic for China along to bear the cost. Therefore, the most effective way for China to establish the design capability of analyzing severe accident for new nuclear power plant projects is to participate the international or multi-national R and D program, such as EUROATOM cost-sharing program and GEN-IV program, etc. By international cooperation, China can not only gain in most extent the successful experience of the countries with advanced technology in developing nuclear power plants, but also contribute itself most effectively in keeping the momentum of enlarging the peaceful utilization of nuclear energy in the world. Certainly, the future Chinese nuclear power market will be a significant industrial driver for developing the-state-of-the-art thermal-hydraulic codes, including hydrogen behavior analysis codes. This paper also reports some computational study on hydrogen diffusion/convection behavior in the containment related to Daya Bay NPP severe accident analysis with CFD code GASFLOW. The code validation were largely carried out in past few years in Germany and had been applied to EPR and other German NPPs. (authors)
Thermal solar energy, towards a sunny interval?
International Nuclear Information System (INIS)
Anon.
2017-01-01
While its market results are continuously decreasing, the thermal solar sector regains confidence with the perspectives of a new thermal legislation in France, a higher carbon tax and the growing volume of installed equipment. This document contains 5 articles, which themes are: The renewal of the thermal solar energy sector in France, notably for the building market, due to a new regulation and a reduction in costs; Several companies are developing large capacity thermal solar plant for industrial facilities (one of them covers 10000 m 2 ) while another company is developing an all-in-one containerised system (less than 1 MW); Another example is given with a Caribbean chemical company which use thermal solar energy for its processes, with a reduction of the fuel consumption by a 2.5 factor; The return of experience show that hybrid solar panels present some limitations, especially in terms of performances and sizing; A collective building (35 apartments) in the West of France has 100 pc of its heating needs (hot water production and space heating) satisfied with solar energy
VIPRE-01: a thermal-hydraulic code for reactor cores. Volume 3: programmer's manual (Revision 2)
International Nuclear Information System (INIS)
Stewart, C.W.; Koontz, A.S.; Cuta, J.M.; Montgomery, S.D.
1985-07-01
The VIPRE thermal-hydraulic computer code for PWR and BWR core analysis has undergone a detailed design review by a committee of experts. A new version of the code, incorporating the committee's recommendations, has been submitted for NRC review and issuance of a safety evaluation report. The changes in the programmers's manual are given
Modeling energy flexibility of low energy buildings utilizing thermal mass
DEFF Research Database (Denmark)
Foteinaki, Kyriaki; Heller, Alfred; Rode, Carsten
2016-01-01
In the future energy system a considerable increase in the penetration of renewable energy is expected, challenging the stability of the system, as both production and consumption will have fluctuating patterns. Hence, the concept of energy flexibility will be necessary in order for the consumption...... to match the production patterns, shifting demand from on-peak hours to off-peak hours. Buildings could act as flexibility suppliers to the energy system, through load shifting potential, provided that the large thermal mass of the building stock could be utilized for energy storage. In the present study...... the load shifting potential of an apartment of a low energy building in Copenhagen is assessed, utilizing the heat storage capacity of the thermal mass when the heating system is switched off for relieving the energy system. It is shown that when using a 4-hour preheating period before switching off...
Energy exchange in thermal energy atom-surface scattering: impulsive models
International Nuclear Information System (INIS)
Barker, J.A.; Auerbach, D.J.
1979-01-01
Energy exchange in thermal energy atom surface collisions is studied using impulsive ('hard cube' and 'hard sphere') models. Both models reproduce the observed nearly linear relation between outgoing and incoming energies. In addition, the hard-sphere model accounts for the widths of the outcoming energy distributions. (Auth.)
Energy Storage Thermal Safety | Transportation Research | NREL
reaction/thermal runaway, internal short circuit, and electrical/chemical/thermal network models are used contributions to the U.S. Department of Energy's Computer-Aided Engineering of Batteries (CAEBAT) project Li-ion battery geometries. Chemical components in Li-ion batteries become thermally unstable when
Aquifer thermal-energy-storage costs with a seasonal-chill source
Brown, D. R.
1983-01-01
The cost of energy supplied by an aquifer thermal energy storage (ATES) ystem from a seasonal chill source was investigated. Costs were estimated for point demand and residential development ATES systems using the computer code AQUASTOR. AQUASTOR was developed at PNL specifically for the economic analysis of ATES systems. In this analysis the cost effect of varying a wide range of technical and economic parameters was examined. Those parameters exhibiting a substantial influence on the costs of ATES delivered chill were: system size; well flow rate; transmission distance; source temperature; well depth; and cost of capital. The effects of each parameter are discussed. Two primary constraints of ATES chill systems are the extremely low energy density of the storage fluid and the prohibitive costs of lengthly pipelines for delivering chill to residential users. This economic analysis concludes that ATES-delivered chill will not be competitive for residential cooling applications. The otherwise marginal attractiveness of ATES chill systems vanishes under the extremely low load factors characteristic of residential cooling systems. (LCL)
Novel Magnetic-to-Thermal Conversion and Thermal Energy Management Composite Phase Change Material
Directory of Open Access Journals (Sweden)
Xiaoqiao Fan
2018-05-01
Full Text Available Superparamagnetic materials have elicited increasing interest due to their high-efficiency magnetothermal conversion. However, it is difficult to effectively manage the magnetothermal energy due to the continuous magnetothermal effect at present. In this study, we designed and synthesized a novel Fe3O4/PEG/SiO2 composite phase change material (PCM that can simultaneously realize magnetic-to-thermal conversion and thermal energy management because of outstanding thermal energy storage ability of PCM. The composite was fabricated by in situ doping of superparamagnetic Fe3O4 nanoclusters through a simple sol–gel method. The synthesized Fe3O4/PEG/SiO2 PCM exhibited good thermal stability, high phase change enthalpy, and excellent shape-stabilized property. This study provides an additional promising route for application of the magnetothermal effect.
Romania needs a strategy for thermal energy
Directory of Open Access Journals (Sweden)
Leca Aureliu
2015-06-01
Full Text Available The energy sector in Romania consists of three sub-sectors: electricity, natural gas and heat. Among these, the sub-sector of thermal energy is in the most precarious situation because it has been neglected for a long time. This sub-sector is particularly important both due to the amount of final heat consumption (of over 50% of final energy consumption, and to the fact that it has a direct negative effect on the population, industry and services. This paper presents the main directions for developing a modern strategy of the thermal energy sub-sector, which would fit into Romania’s Energy Strategy that is still in preparation This is based on the author’s 50 years of experience in this field that includes knowledge about the processes and the equipment of thermal energy, expertise in the management and restructuring of energy companies and also knowledge of the specific legislation. It is therefore recommended, following the European regulations and practices, the promotion and upgrading of district heating systems using efficient cogeneration, using trigeneration in Romania, modernizing buildings in terms of energy use, using of renewable energy sources for heating, especially biomass, and modernizing the energy consumption of rural settlements.
Phase Change Materials for Thermal Energy Storage
Stiebra, L; Cabulis, U; Knite, M
2014-01-01
Phase change materials (PCMs) for thermal energy storage (TES) have become an important subject of research in recent years. Using PCMs for thermal energy storage provides a solution to increase the efficiency of the storage and use of energy in many domestic and industrial sectors. Phase change TES systems offer a number of advantages over other systems (e.g. chemical storage systems): particularly small temperature distance between the storage and retrieval cycles, small unit sizes and lo...
Using Nanoparticles for Enhance Thermal Conductivity of Latent Heat Thermal Energy Storage
Directory of Open Access Journals (Sweden)
Baydaa Jaber Nabhan
2015-06-01
Full Text Available Phase change materials (PCMs such as paraffin wax can be used to store or release large amount of energy at certain temperature at which their solid-liquid phase changes occurs. Paraffin wax that used in latent heat thermal energy storage (LHTES has low thermal conductivity. In this study, the thermal conductivity of paraffin wax has been enhanced by adding different mass concentration (1wt.%, 3wt.%, 5wt.% of (TiO2 nano-particles with about (10nm diameter. It is found that the phase change temperature varies with adding (TiO2 nanoparticles in to the paraffin wax. The thermal conductivity of the composites is found to decrease with increasing temperature. The increase in thermal conductivity has been found to increase by about (10% at nanoparticles loading (5wt.% and 15oC.
International Nuclear Information System (INIS)
Green, C.
1979-12-01
A set of FORTRAN subroutines is described for calculating water thermodynamic properties. These were written for use in a transient thermal-hydraulics program, where speed of execution is paramount. The choice of which subroutines to optimise depends on the primary variables in the thermal-hydraulics code. In this particular case the subroutine which has been optimised is the one which calculates pressure and specific enthalpy given the specific volume and the specific internal energy. Another two subroutines are described which complete a self-consistent set. These calculate the specific volume and the temperature given the pressure and the specific enthalpy, and the specific enthalpy and the specific volume given the pressure and the temperature (or the quality). The accuracy is high near the saturation lines, typically less than 1% relative error, and decreases as the fluid becomes more subcooled in the liquid region or more superheated in the steam region. This behaviour is inherent in the method which uses quantities defined on the saturation lines and assumes that certain derivatives are constant for excursions away from these saturation lines. The accuracy and speed of the subroutines are discussed in detail in this report. (author)
Efficient Solar-Thermal Energy Harvest Driven by Interfacial Plasmonic Heating-Assisted Evaporation.
Chang, Chao; Yang, Chao; Liu, Yanming; Tao, Peng; Song, Chengyi; Shang, Wen; Wu, Jianbo; Deng, Tao
2016-09-07
The plasmonic heating effect of noble nanoparticles has recently received tremendous attention for various important applications. Herein, we report the utilization of interfacial plasmonic heating-assisted evaporation for efficient and facile solar-thermal energy harvest. An airlaid paper-supported gold nanoparticle thin film was placed at the thermal energy conversion region within a sealed chamber to convert solar energy into thermal energy. The generated thermal energy instantly vaporizes the water underneath into hot vapors that quickly diffuse to the thermal energy release region of the chamber to condense into liquids and release the collected thermal energy. The condensed water automatically flows back to the thermal energy conversion region under the capillary force from the hydrophilic copper mesh. Such an approach simultaneously realizes efficient solar-to-thermal energy conversion and rapid transportation of converted thermal energy to target application terminals. Compared to conventional external photothermal conversion design, the solar-thermal harvesting device driven by the internal plasmonic heating effect has reduced the overall thermal resistance by more than 50% and has demonstrated more than 25% improvement of solar water heating efficiency.
International Nuclear Information System (INIS)
Guillermier, Pierre; Daniel, Lucile; Gauthier, Laurent
2009-01-01
To support AREVA NP in its design on HTR reactor and its HTR fuel R and D program, the Commissariat a l'Energie Atomique developed the ATLAS code (Advanced Thermal mechanicaL Analysis Software) with the objectives: - to quantify, with a statistical approach, the failed particle fraction and fission product release of a HTR fuel core under normal and accidental conditions (compact or pebble design). - to simulate irradiation tests or benchmark in order to compare measurements or others code results with ATLAS evaluation. These two objectives aim at qualifying the code in order to predict fuel behaviour and to design fuel according to core performance and safety requirements. A statistical calculation uses numerous deterministic calculations. The finite element method is used for these deterministic calculations, in order to be able to choose among three types of meshes, depending on what must be simulated: - One-dimensional calculation of one single particle, for intact particles or particles with fully debonded layers. - Two-dimensional calculations of one single particle, in the case of particles which are cracked, partially debonded or shaped in various ways. - Three-dimensional calculations of a whole compact slice, in order to simulate the interactions between the particles, the thermal gradient and the transport of fission products up to the coolant. - Some calculations of a whole pebble, using homogenization methods are being studied. The temperatures, displacements, stresses, strains and fission product concentrations are calculated on each mesh of the model. Statistical calculations are done using these results, taking into account ceramic failure mode, but also fabrication tolerances and material property uncertainties, variations of the loads (fluence, temperature, burn-up) and core data parameters. The statistical method used in ATLAS is the importance sampling. The model of migration of long-lived fission products in the coated particle and more
Fire-accident analysis code (FIRAC) verification
International Nuclear Information System (INIS)
Nichols, B.D.; Gregory, W.S.; Fenton, D.L.; Smith, P.R.
1986-01-01
The FIRAC computer code predicts fire-induced transients in nuclear fuel cycle facility ventilation systems. FIRAC calculates simultaneously the gas-dynamic, material transport, and heat transport transients that occur in any arbitrarily connected network system subjected to a fire. The network system may include ventilation components such as filters, dampers, ducts, and blowers. These components are connected to rooms and corridors to complete the network for moving air through the facility. An experimental ventilation system has been constructed to verify FIRAC and other accident analysis codes. The design emphasizes network system characteristics and includes multiple chambers, ducts, blowers, dampers, and filters. A larger industrial heater and a commercial dust feeder are used to inject thermal energy and aerosol mass. The facility is instrumented to measure volumetric flow rate, temperature, pressure, and aerosol concentration throughout the system. Aerosol release rates and mass accumulation on filters also are measured. We have performed a series of experiments in which a known rate of thermal energy is injected into the system. We then simulated this experiment with the FIRAC code. This paper compares and discusses the gas-dynamic and heat transport data obtained from the ventilation system experiments with those predicted by the FIRAC code. The numerically predicted data generally are within 10% of the experimental data
DELOCA, a code for simulation of CANDU fuel channel in thermal transients
International Nuclear Information System (INIS)
Mihalache, M.; Florea, Silviu; Ionescu, V.; Pavelescu, M.
2005-01-01
Full text: In certain LOCA scenarios into the CANDU fuel channel, the ballooning of the pressure tube and the contact with the calandria tube can occur. After the contact moment, a radial heat transfer from cooling fluid to moderator arises through the contact area. If the temperature of channel walls increases, the contact area is drying, the heat transfer becomes inefficiently and the fuel channel could lose its integrity. DELOCA code was developed to simulate the mechanical behaviour of pressure tube during pre-contact transition, and mechanical and thermal behaviour of pressure tube and calandria tube after the contact between the two tubes. The code contains a few models: the creep of Zr-2.5%Nb alloy, the heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. This code was systematically verified by Contact1 and Cathena codes. This paper presents the results obtained at different temperature increasing rates. In addition, the contact moment for a RIH 5% postulated accident was calculated. The Cathena thermo-hydraulic code provided the input data. (authors)
DELOCA, a code for simulation of CANDU fuel channel in thermal transients
International Nuclear Information System (INIS)
Mihalache, M.; Florea, Silviu; Ionescu, V.; Pavelescu, M.
2005-01-01
In certain LOCA scenarios into the CANDU fuel channel, the ballooning of the pressure tube and the contact with the calandria tube can occur. After the contact moment, a radial heat transfer from cooling fluid to moderator arises through the contact area. If the temperature of channel walls increases, the contact area is drying, the heat transfer becomes inefficiently and the fuel channel could lose its integrity. DELOCA code was developed to simulate the mechanical behaviour of pressure tube during pre-contact transition, and mechanical and thermal behaviour of pressure tube and calandria tube after the contact between the two tubes. The code contains a few models: the creep of Zr-2.5%Nb alloy, the heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. This code was systematically verified by Contact1 and Cathena codes. This paper presents the results obtained at different temperature increasing rates. In addition, the contact moment for a RIH 5% postulated accident was calculated. The Cathena thermo-hydraulic code provided the input data. (authors)
Thermal energy and charge currents in multi-terminal nanorings
Energy Technology Data Exchange (ETDEWEB)
Kramer, Tobias [Novel Materials Group, Institut für Physik, Humboldt-Universität zu Berlin, 12489 Berlin (Germany); Konrad-Zuse-Zentrum für Informationstechnik Berlin, 14195 Berlin (Germany); Kreisbeck, Christoph; Riha, Christian, E-mail: riha@physik.hu-berlin.de; Chiatti, Olivio; Buchholz, Sven S.; Fischer, Saskia F. [Novel Materials Group, Institut für Physik, Humboldt-Universität zu Berlin, 12489 Berlin (Germany); Wieck, Andreas D. [Angewandte Festkörperphysik, Ruhr-Universität Bochum, 44780 Bochum (Germany); Reuter, Dirk [Optoelektronische Materialien und Bauelemente, Universität Paderborn, 33098 Paderborn (Germany)
2016-06-15
We study in experiment and theory thermal energy and charge transfer close to the quantum limit in a ballistic nanodevice, consisting of multiply connected one-dimensional electron waveguides. The fabricated device is based on an AlGaAs/GaAs heterostructure and is covered by a global top-gate to steer the thermal energy and charge transfer in the presence of a temperature gradient, which is established by a heating current. The estimate of the heat transfer by means of thermal noise measurements shows the device acting as a switch for charge and thermal energy transfer. The wave-packet simulations are based on the multi-terminal Landauer-Büttiker approach and confirm the experimental finding of a mode-dependent redistribution of the thermal energy current, if a scatterer breaks the device symmetry.
SWAT2: The improved SWAT code system by incorporating the continuous energy Monte Carlo code MVP
International Nuclear Information System (INIS)
Mochizuki, Hiroki; Suyama, Kenya; Okuno, Hiroshi
2003-01-01
SWAT is a code system, which performs the burnup calculation by the combination of the neutronics calculation code, SRAC95 and the one group burnup calculation code, ORIGEN2.1. The SWAT code system can deal with the cell geometry in SRAC95. However, a precise treatment of resonance absorptions by the SRAC95 code using the ultra-fine group cross section library is not directly applicable to two- or three-dimensional geometry models, because of restrictions in SRAC95. To overcome this problem, SWAT2 which newly introduced the continuous energy Monte Carlo code, MVP into SWAT was developed. Thereby, the burnup calculation by the continuous energy in any geometry became possible. Moreover, using the 147 group cross section library called SWAT library, the reactions which are not dealt with by SRAC95 and MVP can be treated. OECD/NEA burnup credit criticality safety benchmark problems Phase-IB (PWR, a single pin cell model) and Phase-IIIB (BWR, fuel assembly model) were calculated as a verification of SWAT2, and the results were compared with the average values of calculation results of burnup calculation code of each organization. Through two benchmark problems, it was confirmed that SWAT2 was applicable to the burnup calculation of the complicated geometry. (author)
Energy Technology Data Exchange (ETDEWEB)
Bartzis, J G; Megaritou, A; Belessiotis, V
1987-09-01
THEAP-I is a computer code developed in NRCPS `DEMOCRITUS` with the aim to contribute to the safety analysis of the open pool research reactors. THEAP-I is designed for three dimensional, transient thermal/hydraulic analysis of a thermally interacting channel bundle totally immersed into water or air, such as the reactor core. In the present report the mathematical and physical models and methods of the solution are given as well as the code description and the input data. A sample problem is also included, refering to the Greek Research Reactor analysis, under an hypothetical severe loss of coolant accident.
Improvement of multi-dimensional realistic thermal-hydraulic system analysis code, MARS 1.3
International Nuclear Information System (INIS)
Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok
1998-09-01
The MARS (Multi-dimensional Analysis of Reactor Safety) code is a multi-dimensional, best-estimate thermal-hydraulic system analysis code. This report describes the new features that have been improved in the MARS 1.3 code since the release of MARS 1.3 in July 1998. The new features include: - implementation of point kinetics model into the 3D module - unification of the heat structure model - extension of the control function to the 3D module variables - improvement of the 3D module input check function. Each of the items has been implemented in the developmental version of the MARS 1.3.1 code and, then, independently verified and assessed. The effectiveness of the new features is well verified and it is shown that these improvements greatly extend the code capability and enhance the user friendliness. Relevant input data changes are also described. In addition to the improvements, this report briefly summarizes the future code developmental activities that are being carried out or planned, such as coupling of MARS 1.3 with the containment code CONTEMPT and the three-dimensional reactor kinetics code MASTER 2.0. (author). 8 refs
Improvement of multi-dimensional realistic thermal-hydraulic system analysis code, MARS 1.3
Energy Technology Data Exchange (ETDEWEB)
Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok
1998-09-01
The MARS (Multi-dimensional Analysis of Reactor Safety) code is a multi-dimensional, best-estimate thermal-hydraulic system analysis code. This report describes the new features that have been improved in the MARS 1.3 code since the release of MARS 1.3 in July 1998. The new features include: - implementation of point kinetics model into the 3D module - unification of the heat structure model - extension of the control function to the 3D module variables - improvement of the 3D module input check function. Each of the items has been implemented in the developmental version of the MARS 1.3.1 code and, then, independently verified and assessed. The effectiveness of the new features is well verified and it is shown that these improvements greatly extend the code capability and enhance the user friendliness. Relevant input data changes are also described. In addition to the improvements, this report briefly summarizes the future code developmental activities that are being carried out or planned, such as coupling of MARS 1.3 with the containment code CONTEMPT and the three-dimensional reactor kinetics code MASTER 2.0. (author). 8 refs.
Theory of the space-dependent fuel management computer code ''UAFCC''
International Nuclear Information System (INIS)
El-Meshad, Y.; Morsy, S.; El-Osery, I.A.
1981-01-01
This report displays the theory of the spatial burnup computer code ''UAFCC'' which has been constructed as a part of an integrated reactor calculation scheme proposed at the Reactors Department of the ARE Atomic Energy Authority. The ''UAFCC'' is a single energy-one-dimensional diffusion burnup FORTRAN computer code for well moderated, multiregion, cylindrical thermal reactors. The effect of reactivity variation with burnup is introduced in the steady state diffusion equation by a fictitious neutron source. The infinite multiplication factor, the total migration area, and the power density per unit thermal flux are calculated from the point model burnup code ''UABUC'' fitted to polynomials of suitable degree in the flux-time, and then used as an input data to the ''UAFCC'' code. The proposed burnup spatial model has been used to study the different stratogemes of the incore fuel management schemes. The conclusions of this study will be presented in a future publication. (author)
An improved thermal-hydraulic modeling of the Jules Horowitz Reactor using the CATHARE2 system code
Energy Technology Data Exchange (ETDEWEB)
Pegonen, R., E-mail: pegonen@kth.se [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden); Bourdon, S.; Gonnier, C. [CEA, DEN, DER, SRJH, CEA Cadarache, 13108 Saint-Paul-lez-Durance Cedex (France); Anglart, H. [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden)
2017-01-15
Highlights: • An improved thermal-hydraulic modeling of the JHR reactor is described. • Thermal-hydraulics of the JHR is analyzed during loss of flow accident. • The heat exchanger approach gives more realistic and less conservative results. - Abstract: The newest European high performance material testing reactor, the Jules Horowitz Reactor, will support current and future nuclear reactor designs. The reactor is under construction at the CEA Cadarache research center in southern France and is expected to achieve first criticality at the end of this decade. This paper presents an improved thermal-hydraulic modeling of the reactor using solely CATHARE2 system code. Up to now, the CATHARE2 code was simulating the full reactor with a simplified approach for the core and the boundary conditions were transferred into the three-dimensional FLICA4 core simulation. A new more realistic methodology is utilized to analyze the thermal-hydraulic simulation of the reactor during a loss of flow accident.
Recycled Thermal Energy from High Power Light Emitting Diode Light Source.
Ji, Jae-Hoon; Jo, GaeHun; Ha, Jae-Geun; Koo, Sang-Mo; Kamiko, Masao; Hong, JunHee; Koh, Jung-Hyuk
2018-09-01
In this research, the recycled electrical energy from wasted thermal energy in high power Light Emitting Diode (LED) system will be investigated. The luminous efficiency of lights has been improved in recent years by employing the high power LED system, therefore energy efficiency was improved compared with that of typical lighting sources. To increase energy efficiency of high power LED system further, wasted thermal energy should be re-considered. Therefore, wasted thermal energy was collected and re-used them as electrical energy. The increased electrical efficiency of high power LED devices was accomplished by considering the recycled heat energy, which is wasted thermal energy from the LED. In this work, increased electrical efficiency will be considered and investigated by employing the high power LED system, which has high thermal loss during the operating time. For this research, well designed thermoelement with heat radiation system was employed to enhance the collecting thermal energy from the LED system, and then convert it as recycled electrical energy.
LiH thermal energy storage device
Olszewski, M.; Morris, D.G.
1994-06-28
A thermal energy storage device for use in a pulsed power supply to store waste heat produced in a high-power burst operation utilizes lithium hydride as the phase change thermal energy storage material. The device includes an outer container encapsulating the lithium hydride and an inner container supporting a hydrogen sorbing sponge material such as activated carbon. The inner container is in communication with the interior of the outer container to receive hydrogen dissociated from the lithium hydride at elevated temperatures. 5 figures.
77 FR 29322 - Updating State Residential Building Energy Efficiency Codes
2012-05-17
... Glass Company North America (AGC). However, DOE notes that PNA/AGC's comment was received late. Although... following. Steel-framed wall insulation Air barrier location Changes whose effect is unclear: Fenestration... code's primary regulation of a home's envelope thermal resistance, or the resistance of the ceilings...
Metal hydrides based high energy density thermal battery
International Nuclear Information System (INIS)
Fang, Zhigang Zak; Zhou, Chengshang; Fan, Peng; Udell, Kent S.; Bowman, Robert C.; Vajo, John J.; Purewal, Justin J.; Kekelia, Bidzina
2015-01-01
Highlights: • The principle of the thermal battery using advanced metal hydrides was demonstrated. • The thermal battery used MgH 2 and TiMnV as a working pair. • High energy density can be achieved by the use of MgH 2 to store thermal energy. - Abstract: A concept of thermal battery based on advanced metal hydrides was studied for heating and cooling of cabins in electric vehicles. The system utilized a pair of thermodynamically matched metal hydrides as energy storage media. The pair of hydrides that was identified and developed was: (1) catalyzed MgH 2 as the high temperature hydride material, due to its high energy density and enhanced kinetics; and (2) TiV 0.62 Mn 1.5 alloy as the matching low temperature hydride. Further, a proof-of-concept prototype was built and tested, demonstrating the potential of the system as HVAC for transportation vehicles
Development of best estimate auditing code for CANDU thermal-hydraulic safety analysis
Energy Technology Data Exchange (ETDEWEB)
Chung, Bub Dong; Lee, Won Jae; Hwang, Moon Kyu; Lim, Hong Sik [Korea Atomic Energy Research Institute, Taejeon (Korea)
2002-04-01
The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3.The study was performed by reconsideration of the previous code assessment works and phenomena identification for essential accident scenario. Improvement areas of model development for auditing tool were identified based on the code comparison and PIRT results. Nine models have been improved significantly for the analysis of LOCA and Mon LOCA event. Conceptual problem or separate effect assessment have been performed to verify the model improvement. The linking calculation with CONTAIN 2.0 has been also enabled to establish the unified auditing code system. Analysis for the CANDU plant real transient and hypothetical LOCA bas been performed using the improved version. It has been concluded that the developed version can be utilized for the auditing analysis of LOCA and non-LOCA event for the CANDU reactor. 25 refs., 84 figs., 36 tabs. (Author)
ANTEO: An optimised PC computer code for the steady state thermal hydraulic analysis of rod bundles
International Nuclear Information System (INIS)
Cevolani, S.
1996-07-01
The paper deals with the description of a Personal Computer oriented subchannel code, devoted to the steady state thermal hydraulic analysis of nuclear reactor fuel bundles. The development of a such code was made possible by two facts: first, the increase the computing power of the desk machines; secondly, the fact several years of experience into operate subchannels codes have shown how to simplify many of the physical models without a sensible loss of accuracy. For sake of validation, the developed code was compared with a traditional subchannel code, the COBRA one. The results of the comparison show a very good agreement between the two codes
Azobenzene-functionalized carbon nanotubes as high-energy density solar thermal fuels.
Kolpak, Alexie M; Grossman, Jeffrey C
2011-08-10
Solar thermal fuels, which reversibly store solar energy in molecular bonds, are a tantalizing prospect for clean, renewable, and transportable energy conversion/storage. However, large-scale adoption requires enhanced energy storage capacity and thermal stability. Here we present a novel solar thermal fuel, composed of azobenzene-functionalized carbon nanotubes, with the volumetric energy density of Li-ion batteries. Our work also demonstrates that the inclusion of nanoscale templates is an effective strategy for design of highly cyclable, thermally stable, and energy-dense solar thermal fuels.
Design and installation manual for thermal energy storage
Energy Technology Data Exchange (ETDEWEB)
Cole, R L; Nield, K J; Rohde, R R; Wolosewicz, R M [eds.
1979-02-01
The purpose for this manual is to provide information on the design and installation of thermal energy storage in solar heating systems. It is intended for contractors, installers, solar system designers, engineers, architects, and manufacturers who intend to enter the solar energy business. The reader should have general knowledge of how solar heating systems operate and knowledge of construction methods and building codes. Knowledge of solar analysis methods such as f-chart, SOLCOST, DOE-1, or TRNSYS would be helpful. The information contained in the manual includes sizing storage, choosing a location for the storage device, and insulation requirements. Both air-based and liquid-based systems are covered with topics on designing rock beds, tank types, pump and fan selection, installation, costs, and operation and maintenance. Topics relevant to heating domestic water include safety, single- and dual-tank systems, domestic water heating with air- and liquid-based space heating system, and stand-alone domestic hot water systems. Several appendices present common problems with storage systems and their solutions, heat transfer fluid properties, heat exchanger sizing, and sample specifications for heat exchangers, wooden rock bins, steel tanks, concrete tanks, and fiberglass-reinforced plastic tanks.
The KFA-Version of the high-energy transport code HETC and the generalized evaluation code SIMPEL
International Nuclear Information System (INIS)
Cloth, P.; Filges, D.; Sterzenbach, G.; Armstrong, T.W.; Colborn, B.L.
1983-03-01
This document describes the updates that have been made to the high-energy transport code HETC for use in the German spallation-neutron source project SNQ. Performance and purpose of the subsidiary code SIMPEL that has been written for general analysis of the HETC output are also described. In addition means of coupling to low energy transport programs, such as the Monte-Carlo code MORSE is provided. As complete input descriptions for HETC and SIMPEL are given together with a sample problem, this document can serve as a user's manual for these two codes. The document is also an answer to the demand that has been issued by a greater community of HETC users on the ICANS-IV meeting, Oct 20-24 1980, Tsukuba-gun, Japan for a complete description of at least one single version of HETC among the many different versions that exist. (orig.)
Thermal Properties of Cement-Based Composites for Geothermal Energy Applications
Bao, Xiaohua; Memon, Shazim Ali; Yang, Haibin; Dong, Zhijun; Cui, Hongzhi
2017-01-01
Geothermal energy piles are a quite recent renewable energy technique where geothermal energy in the foundation of a building is used to transport and store geothermal energy. In this paper, a structural–functional integrated cement-based composite, which can be used for energy piles, was developed using expanded graphite and graphite nanoplatelet-based composite phase change materials (CPCMs). Its mechanical properties, thermal-regulatory performance, and heat of hydration were evaluated. Test results showed that the compressive strength of GNP-Paraffin cement-based composites at 28 days was more than 25 MPa. The flexural strength and density of thermal energy storage cement paste composite decreased with increases in the percentage of CPCM in the cement paste. The infrared thermal image analysis results showed superior thermal control capability of cement based materials with CPCMs. Hence, the carbon-based CPCMs are promising thermal energy storage materials and can be used to improve the durability of energy piles. PMID:28772823
Thermal Properties of Cement-Based Composites for Geothermal Energy Applications
Directory of Open Access Journals (Sweden)
Xiaohua Bao
2017-04-01
Full Text Available Geothermal energy piles are a quite recent renewable energy technique where geothermal energy in the foundation of a building is used to transport and store geothermal energy. In this paper, a structural–functional integrated cement-based composite, which can be used for energy piles, was developed using expanded graphite and graphite nanoplatelet-based composite phase change materials (CPCMs. Its mechanical properties, thermal-regulatory performance, and heat of hydration were evaluated. Test results showed that the compressive strength of GNP-Paraffin cement-based composites at 28 days was more than 25 MPa. The flexural strength and density of thermal energy storage cement paste composite decreased with increases in the percentage of CPCM in the cement paste. The infrared thermal image analysis results showed superior thermal control capability of cement based materials with CPCMs. Hence, the carbon-based CPCMs are promising thermal energy storage materials and can be used to improve the durability of energy piles.
Thermal Properties of Cement-Based Composites for Geothermal Energy Applications.
Bao, Xiaohua; Memon, Shazim Ali; Yang, Haibin; Dong, Zhijun; Cui, Hongzhi
2017-04-27
Geothermal energy piles are a quite recent renewable energy technique where geothermal energy in the foundation of a building is used to transport and store geothermal energy. In this paper, a structural-functional integrated cement-based composite, which can be used for energy piles, was developed using expanded graphite and graphite nanoplatelet-based composite phase change materials (CPCMs). Its mechanical properties, thermal-regulatory performance, and heat of hydration were evaluated. Test results showed that the compressive strength of GNP-Paraffin cement-based composites at 28 days was more than 25 MPa. The flexural strength and density of thermal energy storage cement paste composite decreased with increases in the percentage of CPCM in the cement paste. The infrared thermal image analysis results showed superior thermal control capability of cement based materials with CPCMs. Hence, the carbon-based CPCMs are promising thermal energy storage materials and can be used to improve the durability of energy piles.
Proceedings of the General Committee for solar thermal energy 2015
International Nuclear Information System (INIS)
Gibert, Francois; Loyen, Richard; Khebchache, Bouzid; Cholin, Xavier; Leicher, David; Mozas, Kevin; Leclercq, Martine; Laugier, Patrick; Dias, Pedro; Kuczer, Eric; Benabdelkarim, Mohamed; Brottier, Laetitia; Soussana, Max; Cheze, David; Mugnier, Daniel; Laplagne, Valerie; Mykieta, Frederic; Ducloux, Antoine; Egret, Dominique; Noisette, Nadege; Peneau, Yvan; Seguis, Anne-Sophie; Gerard, Roland
2017-10-01
After an introducing contribution which discussed the difficult evolution of the solar thermal energy sector in 2015, contributions addressed development plans for SOCOL (a plan for collective solar thermal and solar heat) which aims at reviving the market and at opening new markets. A next set of contributions discussed how solar thermal energy can be at the service of energy transition. Following sessions addressed issues like innovation at the service of solar thermal energy, energetic display of solar systems and application of the Ecodesign and Labelling directives, and the reduction of carbon footprint and the energy dependence of territories
Enhancing radiative energy transfer through thermal extraction
Directory of Open Access Journals (Sweden)
Tan Yixuan
2016-06-01
Full Text Available Thermal radiation plays an increasingly important role in many emerging energy technologies, such as thermophotovoltaics, passive radiative cooling and wearable cooling clothes [1]. One of the fundamental constraints in thermal radiation is the Stefan-Boltzmann law, which limits the maximum power of far-field radiation to P0 = σT4S, where σ is the Boltzmann constant, S and T are the area and the temperature of the emitter, respectively (Fig. 1a. In order to overcome this limit, it has been shown that near-field radiations could have an energy density that is orders of magnitude greater than the Stefan-Boltzmann law [2-7]. Unfortunately, such near-field radiation transfer is spatially confined and cannot carry radiative heat to the far field. Recently, a new concept of thermal extraction was proposed [8] to enhance far-field thermal emission, which, conceptually, operates on a principle similar to oil immersion lenses and light extraction in light-emitting diodes using solid immersion lens to increase light output [62].Thermal extraction allows a blackbody to radiate more energy to the far field than the apparent limit of the Stefan-Boltzmann law without breaking the second law of thermodynamics.
International Nuclear Information System (INIS)
Raymond, P.; Caruge, D.; Paik, H.J.
1994-01-01
The French CEA has recently developed a set of new computer codes for reactor physics computations called the Saphir system which includes CRONOS-2, a three-dimensional neutronic code, FLICA-4, a three-dimensional core thermal hydraulic code, and FLICA-S, a primary loops thermal-hydraulic transient computation code, which are coupled and applied to analyze a severe reactivity accident induced by a thermal hydraulic transient: the Steamline Break accident for a pressurized water reactor until soluble boron begins to accumulate in the core. The coupling of these codes has proved to be numerically stable. 15 figs., 7 refs
International Nuclear Information System (INIS)
Avramova, M.; Ivanov, K.; Arenas, C.
2013-01-01
The principles that support the risk-informed regulation are to be considered in an integrated decision-making process. Thus, any evaluation of licensing issues supported by a safety analysis would take into account both deterministic and probabilistic aspects of the problem. The deterministic aspects will be addressed using Best Estimate code calculations and considering the associated uncertainties i.e. Plus Uncertainty (BEPU) calculations. In recent years there has been an increasing demand from nuclear research, industry, safety and regulation for best estimate predictions to be provided with their confidence bounds. This applies also to the sub-channel thermal-hydraulic codes, which are used to evaluate local safety parameters. The paper discusses the extension of BEPU methods to the sub-channel thermal-hydraulic codes on the example of the Pennsylvania State University (PSU) version of COBRA-TF (CTF). The use of coupled codes supplemented with uncertainty analysis allows to avoid unnecessary penalties due to incoherent approximations in the traditional decoupled calculations, and to obtain more accurate evaluation of margins regarding licensing limit. This becomes important for licensing power upgrades, improved fuel assembly and control rod designs, higher burn-up and others issues related to operating LWRs as well as to the new Generation 3+ designs being licensed now (ESBWR, AP-1000, EPR-1600 and etc.). The paper presents the application of Generalized Perturbation Theory (GPT) to generate uncertainties associated with the few-group assembly homogenized neutron cross-section data used as input in coupled reactor core calculations. This is followed by a discussion of uncertainty propagation methodologies, being implemented by PSU in cooperation of Technical University of Catalonia (UPC) for reactor core calculations and for comprehensive multi-physics simulations. (authors)
Transient analysis and thermal hydraulic margins of GHARR-1 using the PARET/NAL code
International Nuclear Information System (INIS)
Adoo, N.A.
2009-06-01
The PARET code has been adapted by the IAEA for testing transient behaviour in research reactors. The PARET code provides a coupled thermal hydrodynamic and point kinetics capability with a continuous reactivity feedback and an optional voiding model that estimates the voiding produced by the subcooled boiling. The present version of the PARET/ANL 73 code provides a convenient means of assessing the various models and correlations proposed for the use in the analysis of research reactor behaviour. The Monte Carlo N-Particle code (MCNP) has been used to obtain power peaking profile for a two channel PARET/ANL model. A PARET model with the corresponding neutronics and thermal hydraulic characteristics for the miniature neutron source reactor (MNSR) has been used to simulate reactivity accidents for the Ghana Research Reactor - 1(GHARR-1) under the MNSR operation conditions of natural circulation, normal operation and reactivity insertion accidents. The simulation results via the insertion of large reactivity demonstrated the high inherent safety features of the MNSR for which the high negative reactivity feedback of moderator temperature limits power excursion and avoids consequently the escalation of clad temperature to the level of onset of sub-cooled void formation. The hot channel peaking factors for both radial and axial were found to be 1.17 and 1.44 respectively. Thermal hydraulic performance characteristics were investigated and the safety margins determined. The peak clad and coolant temperatures ranged from 59.18 0 C to 106.75 0 C and 42.95 0 C to 178.44 0 C respectively at which nucleate boiling will occur within the flow channels of the core. (au)
International Nuclear Information System (INIS)
Podlazov, L. N.
1998-01-01
Two realistic benchmark problems are defined and used to assess the performance of coupled thermal-hydraulic and neutronic codes used in simulating dynamic processes in VVER-1000 and RBMK reactor systems. One of the problems simulates a design basis accident involving the ejection of three control and protection system rods from a VVER-1000 reactor. The other is based on a postulated rod withdrawal from an operating RBMK reactor. Preliminary results calculated by various codes are compared. While these results show significant differences, the intercomparisons performed so far provide a basis for further evaluation of code limitations and modeling assumptions
Thermal power plant efficiency enhancement with Ocean Thermal Energy Conversion
International Nuclear Information System (INIS)
Soto, Rodrigo; Vergara, Julio
2014-01-01
In addition to greenhouse gas emissions, coastal thermal power plants would gain further opposition due to their heat rejection distressing the local ecosystem. Therefore, these plants need to enhance their thermal efficiency while reducing their environmental offense. In this study, a hybrid plant based on the principle of Ocean Thermal Energy Conversion was coupled to a 740 MW coal-fired power plant project located at latitude 28°S where the surface to deepwater temperature difference would not suffice for regular OTEC plants. This paper presents the thermodynamical model to assess the overall efficiency gained by adopting an ammonia Rankine cycle plus a desalinating unit, heated by the power plant condenser discharge and refrigerated by cold deep seawater. The simulation allowed us to optimize a system that would finally enhance the plant power output by 25–37 MW, depending on the season, without added emissions while reducing dramatically the water temperature at discharge and also desalinating up to 5.8 million tons per year. The supplemental equipment was sized and the specific emissions reduction was estimated. We believe that this approach would improve the acceptability of thermal and nuclear power plant projects regardless of the plant location. -- Highlights: • An Ocean Thermal Energy Conversion hybrid plant was designed. • The waste heat of a power plant was delivered as an OTEC heat source. • The effect of size and operating conditions on plant efficiency were studied. • The OTEC implementation in a Chilean thermal power plant was evaluated. • The net efficiency of the thermal power plant was increased by 1.3%
Energy and Data Throughput for Asymmetric Inter-Session Network Coding
DEFF Research Database (Denmark)
Paramanathan, Achuthan; Heide, Janus; Pahlevani, Peyman
2012-01-01
on commercial platforms. The outcome of this paper confirms the analytical expression, and the results shows that even with a large asymmetric data rate there is a gain in terms of energy consumption and throughput when network coding is applied in compare to the case when network coding is not applied.......In this paper we investigate the impact of asymmetric traffic patterns on the energy consumption and throughput in a wireless multi hop network. Network coding is a novel technique for communication systems and a viable solution for wireless multi hop networks. State-of-the-art research is mainly...
Benchmark study of some thermal and structural computer codes for nuclear shipping casks
International Nuclear Information System (INIS)
Ikushima, Takeshi; Kanae, Yoshioki; Shimada, Hirohisa; Shimoda, Atsumu; Halliquist, J.O.
1984-01-01
There are many computer codes which could be applied to the design and analysis of nuclear material shipping casks. One of problems which the designer of shipping cask faces is the decision regarding the choice of the computer codes to be used. For this situation, the thermal and structural benchmark tests for nuclear shipping casks are carried out to clarify adequacy of the calculation results. The calculation results are compared with the experimental ones. This report describes the results and discussion of the benchmark test. (author)
New kinds of energy-storing building composite PCMs for thermal energy storage
International Nuclear Information System (INIS)
Biçer, Alper; Sarı, Ahmet
2013-01-01
Graphical abstract: In this work, 10 new kinds of BCPCMs were prepared by blending of liquid xylitol pentalaurate (XPL) and xylitol pentamyristate (XPM) esters into gypsum, cement, diatomite, perlite and vermiculite. DSC results showed that the melting temperatures and energy storage capacities of the prepared BCPCMs are in range of about 40–55 °C and 31–126 J/g, respectively. TG investigations and thermal cycling test showed that the BCPCMs had good thermal endurance and thermal reliability. It can be also concluded that among the prepared 10 kinds materials, especially the BCPCMs including perlite, vermiculite, diatomite were found to better candidates for thermal energy storage applications in buildings due to the fact that they have relatively high heat storage ability. Highlights: ► New kinds BCPCMs were prepared by blending of liquid XPL and XPM esters with some building materials. ► The BCPCMs had suitable melting temperatures and energy storage capacities. ► Especially, the BCPCMs including perlite, vermiculite, diatomite were found to better candidates for thermal energy storage. - Abstract: Energy storing-composite phase change materials (PCMs) are significant means of thermal energy storage in buildings. Although several building composite PCMs (BCPCMs) have been developed in recent years, the additional investigations are still required to enrich the diversity of BCPCMs for solar heating and energy conservation applications in buildings. For this purpose, the present work is focused the preparation, characterization and determination of 10 new kinds of BCPCMs. The BCPCMs were prepared by blending of liquid xylitol pentalaurate (XPL) and xylitol pentamyristate (XPM) esters with gypsum, cement, diatomite, perlite and vermiculite as supporting matrices. The scanning electron microscopy (SEM) and Fourier Transform Infrared (FT-IR) analysis showed that the ester compounds were adsorbed uniformly into the building materials due to capillary forces
ARCADIAR - A New Generation of Coupled Neutronics / Core Thermal- Hydraulics Code System at AREVA NP
International Nuclear Information System (INIS)
Curca-Tivig, Florin; Merk, Stephan; Pautz, Andreas; Thareau, Sebastien
2007-01-01
Anticipating future needs of our customers and willing to concentrate synergies and competences existing in the company for the benefit of our customers, AREVA NP decided in 2002 to develop the next generation of coupled neutronics/ core thermal-hydraulic (TH) code systems for fuel assembly and core design calculations for both, PWR and BWR applications. The global CONVERGENCE project was born: after a feasibility study of one year (2002) and a conceptual phase of another year (2003), development was started at the beginning of 2004. The present paper introduces the CONVERGENCE project, presents the main feature of the new code system ARCADIA R and concludes on customer benefits. ARCADIA R is designed to meet AREVA NP market and customers' requirements worldwide. Besides state-of-the-art physical modeling, numerical performance and industrial functionality, the ARCADIA R system is featuring state-of-the-art software engineering. The new code system will bring a series of benefits for our customers: e.g. improved accuracy for heterogeneous cores (MOX/ UOX, Gd...), better description of nuclide chains, and access to local neutronics/ thermal-hydraulics and possibly thermal-mechanical information (3D pin by pin full core modeling). ARCADIA is a registered trademark of AREVA NP. (authors)
Thermal Energy Storage with Phase Change Material
Directory of Open Access Journals (Sweden)
Lavinia Gabriela SOCACIU
2012-08-01
Full Text Available Thermal energy storage (TES systems provide several alternatives for efficient energy use and conservation. Phase change materials (PCMs for TES are materials supplying thermal regulation at particular phase change temperatures by absorbing and emitting the heat of the medium. TES in general and PCMs in particular, have been a main topic in research for the last 30 years, but although the information is quantitatively enormous, it is also spread widely in the literature, and difficult to find. PCMs absorb energy during the heating process as phase change takes place and release energy to the environment in the phase change range during a reverse cooling process. PCMs possesses the ability of latent thermal energy change their state with a certain temperature. PCMs for TES are generally solid-liquid phase change materials and therefore they need encapsulation. TES systems using PCMs as a storage medium offers advantages such as high TES capacity, small unit size and isothermal behaviour during charging and discharging when compared to the sensible TES.
Energy-Efficient Cluster Based Routing Protocol in Mobile Ad Hoc Networks Using Network Coding
Directory of Open Access Journals (Sweden)
Srinivas Kanakala
2014-01-01
Full Text Available In mobile ad hoc networks, all nodes are energy constrained. In such situations, it is important to reduce energy consumption. In this paper, we consider the issues of energy efficient communication in MANETs using network coding. Network coding is an effective method to improve the performance of wireless networks. COPE protocol implements network coding concept to reduce number of transmissions by mixing the packets at intermediate nodes. We incorporate COPE into cluster based routing protocol to further reduce the energy consumption. The proposed energy-efficient coding-aware cluster based routing protocol (ECCRP scheme applies network coding at cluster heads to reduce number of transmissions. We also modify the queue management procedure of COPE protocol to further improve coding opportunities. We also use an energy efficient scheme while selecting the cluster head. It helps to increase the life time of the network. We evaluate the performance of proposed energy efficient cluster based protocol using simulation. Simulation results show that the proposed ECCRP algorithm reduces energy consumption and increases life time of the network.
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)
Metal hydrides based high energy density thermal battery
Energy Technology Data Exchange (ETDEWEB)
Fang, Zhigang Zak, E-mail: zak.fang@utah.edu [Department of Metallurgical Engineering, The University of Utah, 135 South 1460 East, Room 412, Salt Lake City, UT 84112-0114 (United States); Zhou, Chengshang; Fan, Peng [Department of Metallurgical Engineering, The University of Utah, 135 South 1460 East, Room 412, Salt Lake City, UT 84112-0114 (United States); Udell, Kent S. [Department of Metallurgical Engineering, The University of Utah, 50 S. Central Campus Dr., Room 2110, Salt Lake City, UT 84112-0114 (United States); Bowman, Robert C. [Department of Metallurgical Engineering, The University of Utah, 135 South 1460 East, Room 412, Salt Lake City, UT 84112-0114 (United States); Vajo, John J.; Purewal, Justin J. [HRL Laboratories, LLC, 3011 Malibu Canyon Road, Malibu, CA 90265 (United States); Kekelia, Bidzina [Department of Metallurgical Engineering, The University of Utah, 50 S. Central Campus Dr., Room 2110, Salt Lake City, UT 84112-0114 (United States)
2015-10-05
Highlights: • The principle of the thermal battery using advanced metal hydrides was demonstrated. • The thermal battery used MgH{sub 2} and TiMnV as a working pair. • High energy density can be achieved by the use of MgH{sub 2} to store thermal energy. - Abstract: A concept of thermal battery based on advanced metal hydrides was studied for heating and cooling of cabins in electric vehicles. The system utilized a pair of thermodynamically matched metal hydrides as energy storage media. The pair of hydrides that was identified and developed was: (1) catalyzed MgH{sub 2} as the high temperature hydride material, due to its high energy density and enhanced kinetics; and (2) TiV{sub 0.62}Mn{sub 1.5} alloy as the matching low temperature hydride. Further, a proof-of-concept prototype was built and tested, demonstrating the potential of the system as HVAC for transportation vehicles.
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.)
Directory of Open Access Journals (Sweden)
Ahmad Hasan
2016-09-01
Full Text Available Photovoltaic (PV panels convert a certain amount of incident solar radiation into electricity, while the rest is converted to heat, leading to a temperature rise in the PV. This elevated temperature deteriorates the power output and induces structural degradation, resulting in reduced PV lifespan. One potential solution entails PV thermal management employing active and passive means. The traditional passive means are found to be largely ineffective, while active means are considered to be energy intensive. A passive thermal management system using phase change materials (PCMs can effectively limit PV temperature rises. The PCM-based approach however is cost inefficient unless the stored thermal energy is recovered effectively. The current article investigates a way to utilize the thermal energy stored in the PCM behind the PV for domestic water heating applications. The system is evaluated in the winter conditions of UAE to deliver heat during water heating demand periods. The proposed system achieved a ~1.3% increase in PV electrical conversion efficiency, along with the recovery of ~41% of the thermal energy compared to the incident solar radiation.
Energy Technology Data Exchange (ETDEWEB)
None
1977-12-01
Under the sponsorship of the United States Department of Energy, a Model Code for Energy Conservation in New Building Construction has been developed by those national organizations primarily concerned with the development and promulgation of model codes. The technical provisions are based on ASHRAE Standard 90-75 and are intended for use by state and local officials. The subject of regulation of new building construction to assure energy conservation is recognized as one in which code officials have not had previous exposure. It was also determined that application of the model code would be made at varying levels by officials with both a specific requirement for knowledge and a differing degree of prior training in the state-of-the-art. Therefore, a training program and instructional materials were developed for code officials to assist them in the implementation and enforcement of energy efficient standards and codes. The training program for Energy Conservation Tehnology for Plan Examiners and Code Administrators (ECT Series 200) is presented.
Thermal-hydraulic analysis of water cooled breeding blanket of K-DEMO using MARS-KS code
Energy Technology Data Exchange (ETDEWEB)
Lee, Jeong-Hun; Park, Il Woong; Kim, Geon-Woo; Park, Goon-Cherl [Seoul National University, Seoul (Korea, Republic of); Cho, Hyoung-Kyu, E-mail: chohk@snu.ac.kr [Seoul National University, Seoul (Korea, Republic of); Im, Kihak [National Fusion Research Institute, Daejeon (Korea, Republic of)
2015-10-15
Highlights: • The thermal design of breeding blanket for the K-DEMO is evaluated using MARS-KS. • To confirm the prediction capability of MARS, the results were compared with the CFD. • The results of MARS-KS calculation and CFD prediction are in good agreement. • A transient simulation was carried out so as to show the applicability of MARS-KS. • A methodology to simulate the entire blanket system is proposed. - Abstract: The thermal design of a breeding blanket for the Korean Fusion DEMOnstration reactor (K-DEMO) is evaluated using the Multidimensional Analysis of Reactor Safety (MARS-KS) code in this study. The MARS-KS code has advantages in simulating transient two-phase flow over computational fluid dynamics (CFD) codes. In order to confirm the prediction capability of the code for the present blanket system, the calculation results were compared with the CFD prediction. The results of MARS-KS calculation and CFD prediction are in good agreement. Afterwards, a transient simulation for a conceptual problem was carried out so as to show the applicability of MARS-KS for a transient or accident condition. Finally, a methodology to simulate the multiple blanket modules is proposed.
Potential Job Creation in Nevada as a Result of Adopting New Residential Building Energy Codes
Energy Technology Data Exchange (ETDEWEB)
Scott, Michael J.; Niemeyer, Jackie M.
2013-09-01
Are there advantages to states that adopt the most recent model building energy codes other than saving energy? For example, can the construction activity and energy savings associated with code-compliant housing units become significant sources of job creation for states if new building energy codes are adopted to cover residential construction? , The U.S. Department of Energy (DOE) Building Energy Codes Program (BECP) asked Pacific Northwest National Laboratory (PNNL) to research and ascertain whether jobs would be created in individual states based on their adoption of model building energy codes. Each state in the country is dealing with high levels of unemployment, so job creation has become a top priority. Many programs have been created to combat unemployment with various degrees of failure and success. At the same time, many states still have not yet adopted the most current versions of the International Energy Conservation Code (IECC) model building energy code, when doing so could be a very effective tool in creating jobs to assist states in recovering from this economic downturn.
Potential Job Creation in Tennessee as a Result of Adopting New Residential Building Energy Codes
Energy Technology Data Exchange (ETDEWEB)
Scott, Michael J.; Niemeyer, Jackie M.
2013-09-01
Are there advantages to states that adopt the most recent model building energy codes other than saving energy? For example, can the construction activity and energy savings associated with code-compliant housing units become significant sources of job creation for states if new building energy codes are adopted to cover residential construction? , The U.S. Department of Energy (DOE) Building Energy Codes Program (BECP) asked Pacific Northwest National Laboratory (PNNL) to research and ascertain whether jobs would be created in individual states based on their adoption of model building energy codes. Each state in the country is dealing with high levels of unemployment, so job creation has become a top priority. Many programs have been created to combat unemployment with various degrees of failure and success. At the same time, many states still have not yet adopted the most current versions of the International Energy Conservation Code (IECC) model building energy code, when doing so could be a very effective tool in creating jobs to assist states in recovering from this economic downturn.
Potential Job Creation in Minnesota as a Result of Adopting New Residential Building Energy Codes
Energy Technology Data Exchange (ETDEWEB)
Scott, Michael J.; Niemeyer, Jackie M.
2013-09-01
Are there advantages to states that adopt the most recent model building energy codes other than saving energy? For example, can the construction activity and energy savings associated with code-compliant housing units become significant sources of job creation for states if new building energy codes are adopted to cover residential construction? , The U.S. Department of Energy (DOE) Building Energy Codes Program (BECP) asked Pacific Northwest National Laboratory (PNNL) to research and ascertain whether jobs would be created in individual states based on their adoption of model building energy codes. Each state in the country is dealing with high levels of unemployment, so job creation has become a top priority. Many programs have been created to combat unemployment with various degrees of failure and success. At the same time, many states still have not yet adopted the most current versions of the International Energy Conservation Code (IECC) model building energy code, when doing so could be a very effective tool in creating jobs to assist states in recovering from this economic downturn.
GOTHIC code simulation of thermal stratification in POOLEX facility
International Nuclear Information System (INIS)
Li, H.; Kudinov, P.
2009-07-01
Pressure suppression pool is an important element of BWR containment. It serves as a heat sink and steam condenser to prevent containment pressure buildup during loss of coolant accident or safety relief valve opening during normal operations of a BWR. Insufficient mixing in the pool, in case of low mass flow rate of steam, can cause development of thermal stratification and reduction of pressure suppression pool capacity. For reliable prediction of mixing and stratification phenomena validation of simulation tools has to be performed. Data produced in POOLEX/PPOOLEX facility at Lappeenranta University of Technology about development of thermal stratification in a large scale model of a pressure suppression pool is used for GOTHIC lumped and distributed parameter validation. Sensitivity of GOTHIC solution to different boundary conditions and grid convergence study for 2D simulations of POOLEX STB-20 experiment are performed in the present study. CFD simulation was carried out with FLUENT code in order to get additional insights into physics of stratification phenomena. In order to support development of experimental procedures for new tests in the PPOOLEX facility lumped parameter pre-test GOTHIC simulations were performed. Simulations show that drywell and wetwell pressures can be kept within safety margins during a long transient necessary for development of thermal stratification. (au)
GOTHIC code simulation of thermal stratification in POOLEX facility
Energy Technology Data Exchange (ETDEWEB)
Li, H.; Kudinov, P. (Royal Institute of Technology (KTH) (Sweden))
2009-07-15
Pressure suppression pool is an important element of BWR containment. It serves as a heat sink and steam condenser to prevent containment pressure buildup during loss of coolant accident or safety relief valve opening during normal operations of a BWR. Insufficient mixing in the pool, in case of low mass flow rate of steam, can cause development of thermal stratification and reduction of pressure suppression pool capacity. For reliable prediction of mixing and stratification phenomena validation of simulation tools has to be performed. Data produced in POOLEX/PPOOLEX facility at Lappeenranta University of Technology about development of thermal stratification in a large scale model of a pressure suppression pool is used for GOTHIC lumped and distributed parameter validation. Sensitivity of GOTHIC solution to different boundary conditions and grid convergence study for 2D simulations of POOLEX STB-20 experiment are performed in the present study. CFD simulation was carried out with FLUENT code in order to get additional insights into physics of stratification phenomena. In order to support development of experimental procedures for new tests in the PPOOLEX facility lumped parameter pre-test GOTHIC simulations were performed. Simulations show that drywell and wetwell pressures can be kept within safety margins during a long transient necessary for development of thermal stratification. (au)
Controlling Energy Radiations of Electromagnetic Waves via Frequency Coding Metamaterials.
Wu, Haotian; Liu, Shuo; Wan, Xiang; Zhang, Lei; Wang, Dan; Li, Lianlin; Cui, Tie Jun
2017-09-01
Metamaterials are artificial structures composed of subwavelength unit cells to control electromagnetic (EM) waves. The spatial coding representation of metamaterial has the ability to describe the material in a digital way. The spatial coding metamaterials are typically constructed by unit cells that have similar shapes with fixed functionality. Here, the concept of frequency coding metamaterial is proposed, which achieves different controls of EM energy radiations with a fixed spatial coding pattern when the frequency changes. In this case, not only different phase responses of the unit cells are considered, but also different phase sensitivities are also required. Due to different frequency sensitivities of unit cells, two units with the same phase response at the initial frequency may have different phase responses at higher frequency. To describe the frequency coding property of unit cell, digitalized frequency sensitivity is proposed, in which the units are encoded with digits "0" and "1" to represent the low and high phase sensitivities, respectively. By this merit, two degrees of freedom, spatial coding and frequency coding, are obtained to control the EM energy radiations by a new class of frequency-spatial coding metamaterials. The above concepts and physical phenomena are confirmed by numerical simulations and experiments.
High-fidelity plasma codes for burn physics
Energy Technology Data Exchange (ETDEWEB)
Cooley, James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Graziani, Frank [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marinak, Marty [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Murillo, Michael [Michigan State Univ., East Lansing, MI (United States)
2016-10-19
Accurate predictions of equation of state (EOS), ionic and electronic transport properties are of critical importance for high-energy-density plasma science. Transport coefficients inform radiation-hydrodynamic codes and impact diagnostic interpretation, which in turn impacts our understanding of the development of instabilities, the overall energy balance of burning plasmas, and the efficacy of self-heating from charged-particle stopping. Important processes include thermal and electrical conduction, electron-ion coupling, inter-diffusion, ion viscosity, and charged particle stopping. However, uncertainties in these coefficients are not well established. Fundamental plasma science codes, also called high-fidelity plasma codes, are a relatively recent computational tool that augments both experimental data and theoretical foundations of transport coefficients. This paper addresses the current status of HFPC codes and their future development, and the potential impact they play in improving the predictive capability of the multi-physics hydrodynamic codes used in HED design.
Energy Technology Data Exchange (ETDEWEB)
Falikov, A A; Vakhrushev, V V; Kuul, V S; Samoilov, O B; Tarasov, G I [OKBM, Nizhny Novgorod (Russian Federation)
1997-09-01
The paper briefly reviews the specific thermal-hydraulic problems for AST-type NHRs, the experimental investigations that have been carried out in the RF, and the design procedures and computer codes used for AST-500 thermohydraulic characteristics and safety validation. (author). 13 refs, 10 figs, 1 tab.
Transport code and nuclear data in intermediate energy region
Energy Technology Data Exchange (ETDEWEB)
Hasegawa, Akira; Odama, Naomitsu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Maekawa, F.; Ueki, K.; Kosaka, K.; Oyama, Y.
1998-11-01
We briefly reviewed the problems of intermediate energy nuclear data file and transport codes in connection with processing of the data. This is a summary of our group in the task force on JENDL High Energy File Integral Evaluation (JHEFIE). In this article we stress the necessity of the production of intermediate evaluated nuclear data file up to 3 GeV for the application of accelerator driven transmutation (ADT) system. And also we state the necessity of having our own transport code system to calculate the radiation fields using these evaluated files from the strategic points of view to keep our development of the ADT technology completely free from other conditions outside of our own such as imported codes and data with poor maintenance or unknown accuracy. (author)
Transport code and nuclear data in intermediate energy region
International Nuclear Information System (INIS)
Hasegawa, Akira; Odama, Naomitsu; Maekawa, F.; Ueki, K.; Kosaka, K.; Oyama, Y.
1998-01-01
We briefly reviewed the problems of intermediate energy nuclear data file and transport codes in connection with processing of the data. This is a summary of our group in the task force on JENDL High Energy File Integral Evaluation (JHEFIE). In this article we stress the necessity of the production of intermediate evaluated nuclear data file up to 3 GeV for the application of accelerator driven transmutation (ADT) system. And also we state the necessity of having our own transport code system to calculate the radiation fields using these evaluated files from the strategic points of view to keep our development of the ADT technology completely free from other conditions outside of our own such as imported codes and data with poor maintenance or unknown accuracy. (author)
Seasonal thermal energy storage in shallow geothermal systems: thermal equilibrium stage
Directory of Open Access Journals (Sweden)
Nowamooz Hossein
2016-01-01
Full Text Available This paper is dedicated to the study of seasonal heat storage in shallow geothermal installations in unsaturated soils for which hydrothermal properties such as degree of saturation and thermal conductivity vary with time throughout the profile. In the model, a semi-analytical model which estimates time-spatial thermal conductivity is coupled with a 2D cylindrical heat transfer modeling using finite difference method. The variation of temperature was obtained after 3 heating and cooling cycles for the different types of loads with maximum thermal load of qmax = 15 W.m−1 with variable angular frequency (8 months of heating and 4 months of cooling.and constant angular frequency (6 months of heating and 6 months of cooling to estimate the necessary number of cycles to reach the thermal equilibrium stage. The results show that we approach a thermal equilibrium stage where the same variation of temperature can be observed in soils after several heating and cooling cycles. Based on these simulations, the necessary number of cycles can be related to the total applied energy on the system and the minimum number of cycles is for a system with the total applied energy of 1.9qmax.
Solar Thermal Energy Storage in a Photochromic Macrocycle.
Vlasceanu, Alexandru; Broman, Søren L; Hansen, Anne S; Skov, Anders B; Cacciarini, Martina; Kadziola, Anders; Kjaergaard, Henrik G; Mikkelsen, Kurt V; Nielsen, Mogens Brøndsted
2016-07-25
The conversion and efficient storage of solar energy is recognized to hold significant potential with regard to future energy solutions. Molecular solar thermal batteries based on photochromic systems exemplify one possible technology able to harness and apply this potential. Herein is described the synthesis of a macrocycle based on a dimer of the dihydroazulene/vinylheptafulvene (DHA/VHF) photo/thermal couple. By taking advantage of conformational strain, this DHA-DHA macrocycle presents an improved ability to absorb and store incident light energy in chemical bonds (VHF-VHF). A stepwise energy release over two sequential ring-closing reactions (VHF→DHA) combines the advantages of an initially fast discharge, hypothetically addressing immediate energy consumption needs, followed by a slow process for consistent, long-term use. This exemplifies another step forward in the molecular engineering and design of functional organic materials towards solar thermal energy storage and release. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.
One-dimensional arrays of oscillators: Energy localization in thermal equilibrium
International Nuclear Information System (INIS)
Reigada, R.; Romero, A.H.; Sarmiento, A.; Lindenberg, K.
1999-01-01
All systems in thermal equilibrium exhibit a spatially variable energy landscape due to thermal fluctuations. Thus at any instant there is naturally a thermodynamically driven localization of energy in parts of the system relative to other parts of the system. The specific characteristics of the spatial landscape such as, for example, the energy variance, depend on the thermodynamic properties of the system and vary from one system to another. The temporal persistence of a given energy landscape, that is, the way in which energy fluctuations (high or low) decay toward the thermal mean, depends on the dynamical features of the system. We discuss the spatial and temporal characteristics of spontaneous energy localization in 1D anharmonic chains in thermal equilibrium. copyright 1999 American Institute of Physics
Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes
International Nuclear Information System (INIS)
Baratta, A.J.
1997-01-01
To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together
Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes
Energy Technology Data Exchange (ETDEWEB)
Baratta, A.J.
1997-07-01
To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.
Energy Technology Data Exchange (ETDEWEB)
Dominguez, R.; Moreno, J.
2008-07-01
The take in effect in 2006 of the new Technical Code of Edification (CTE) augured a remarkable impact in the market of the thermal solar energy in Spain. Parallelly, in april of this year was published in Portugal the Regulation of the Characteristics of thermal Behaviours of the Buildings (RCCTE) that also generated promising expectations. A year and a half later, the market does not match to the initial forecasts made by the government agencies and the solar industry. The purpose of this article is to make an analysis of the causes of ralentization in the definitive takeoff of both markets. (Author)
Energy and Environment Guide to Action - Chapter 4.3: Building Codes for Energy Efficiency
Provides guidance and recommendations for establishing, implementing, and evaluating state building codes for energy efficiency, which improve energy efficiency in new construction and major renovations. State success stories are included for reference.
Aquifer thermal energy storage. International symposium: Proceedings
Energy Technology Data Exchange (ETDEWEB)
NONE
1995-05-01
Aquifers have been used to store large quantities of thermal energy to supply process cooling, space cooling, space heating, and ventilation air preheating, and can be used with or without heat pumps. Aquifers are used as energy sinks and sources when supply and demand for energy do not coincide. Aquifer thermal energy storage may be used on a short-term or long-term basis; as the sole source of energy or as a partial storage; at a temperature useful for direct application or needing upgrade. The sources of energy used for aquifer storage are ambient air, usually cold winter air; waste or by-product energy; and renewable energy such as solar. The present technical, financial and environmental status of ATES is promising. Numerous projects are operating and under development in several countries. These projects are listed and results from Canada and elsewhere are used to illustrate the present status of ATES. Technical obstacles have been addressed and have largely been overcome. Cold storage in aquifers can be seen as a standard design option in the near future as it presently is in some countries. The cost-effectiveness of aquifer thermal energy storage is based on the capital cost avoidance of conventional chilling equipment and energy savings. ATES is one of many developments in energy efficient building technology and its success depends on relating it to important building market and environmental trends. This paper attempts to provide guidance for the future implementation of ATES. Individual projects have been processed separately for entry onto the Department of Energy databases.
Low temperature desalination using solar collectors augmented by thermal energy storage
International Nuclear Information System (INIS)
Gude, Veera Gnaneswar; Nirmalakhandan, Nagamany; Deng, Shuguang; Maganti, Anand
2012-01-01
Highlights: ► A new low temperature desalination process using solar collectors was investigated. ► A thermal energy storage tank (TES) was included for continuous process operation. ► Solar collector area and TES volumes were optimized by theoretical simulations. ► Economic analysis for the entire process was compared with and without TES tank. ► Energy and emission payback periods for the solar collector system were reported. -- Abstract: A low temperature desalination process capable of producing 100 L/d freshwater was designed to utilize solar energy harvested from flat plate solar collectors. Since solar insolation is intermittent, a thermal energy storage system was incorporated to run the desalination process round the clock. The requirements for solar collector area as well as thermal energy storage volume were estimated based on the variations in solar insolation. Results from this theoretical study confirm that thermal energy storage is a useful component of the system for conserving thermal energy to meet the energy demand when direct solar energy resource is not available. Thermodynamic advantages of the low temperature desalination using thermal energy storage, as well as energy and environmental emissions payback period of the system powered by flat plate solar collectors are presented. It has been determined that a solar collector area of 18 m 2 with a thermal energy storage volume of 3 m 3 is adequate to produce 100 L/d of freshwater round the clock considering fluctuations in the weather conditions. An economic analysis on the desalination system with thermal energy storage is also presented.
Atucha II NPP full scope simulator modelling with the thermal hydraulic code TRACRT
International Nuclear Information System (INIS)
Alonso, Pablo Rey; Ruiz, Jose Antonio; Rivero, Norberto
2011-01-01
In February 2010 NA-SA (Nucleoelectrica Argentina S.A.) awarded Tecnatom the Atucha II full scope simulator project. NA-SA is a public company owner of the Argentinean nuclear power plants. Atucha II is due to enter in operation shortly. Atucha II NPP is a PHWR type plant cooled by the water of the Parana River and has the same design as the Atucha I unit, doubling its power capacity. Atucha II will produce 745 MWe utilizing heavy water as coolant and moderator, and natural uranium as fuel. A plant singular feature is the permanent core refueling. TRAC R T is the first real time thermal hydraulic six-equations code used in the training simulation industry for NSSS modeling. It is the result from adapting to real time the best estimate code TRACG. TRAC R T is based on first principle conservation equations for mass, energy and momentum for liquid and steam phases, with two phase flows under non homogeneous and non equilibrium conditions. At present, it has been successfully implemented in twelve full scope replica simulators in different training centers throughout the world. To ease the modeling task, TRAC R T includes a graphical pre-processing tool designed to optimize this process and alleviate the burden of entering alpha numerical data in an input file. (author)
Numerical modeling of aquifer thermal energy storage system
Energy Technology Data Exchange (ETDEWEB)
Kim, Jongchan [Korea Institute of Geoscience and Mineral Resources, Geothermal Resources Department, 92 Gwahang-no, Yuseong-gu, Daejeon 305-350 (Korea, Republic of); Kongju National University, Department of Geoenvironmental Sciences, 182 Singwan-dong, Gongju-si, Chungnam 314-701 (Korea, Republic of); Lee, Youngmin [Korea Institute of Geoscience and Mineral Resources, Geothermal Resources Department, 92 Gwahang-no, Yuseong-gu, Daejeon 305-350 (Korea, Republic of); Yoon, Woon Sang; Jeon, Jae Soo [nexGeo Inc., 134-1 Garak 2-dong, Songpa-gu, Seoul 138-807 (Korea, Republic of); Koo, Min-Ho; Keehm, Youngseuk [Kongju National University, Department of Geoenvironmental Sciences, 182 Singwan-dong, Gongju-si, Chungnam 314-701 (Korea, Republic of)
2010-12-15
The performance of the ATES (aquifer thermal energy storage) system primarily depends on the thermal interference between warm and cold thermal energy stored in an aquifer. Additionally the thermal interference is mainly affected by the borehole distance, the hydraulic conductivity, and the pumping/injection rate. Thermo-hydraulic modeling was performed to identify the thermal interference by three parameters and to estimate the system performance change by the thermal interference. Modeling results indicate that the thermal interference grows as the borehole distance decreases, as the hydraulic conductivity increases, and as the pumping/injection rate increases. The system performance analysis indicates that if {eta} (the ratio of the length of the thermal front to the distance between two boreholes) is lower than unity, the system performance is not significantly affected, but if {eta} is equal to unity, the system performance falls up to {proportional_to}22%. Long term modeling for a factory in Anseong was conducted to test the applicability of the ATES system. When the pumping/injection rate is 100 m{sup 3}/day, system performances during the summer and winter after 3 years of operation are estimated to be {proportional_to}125 kW and {proportional_to}110 kW, respectively. Therefore, 100 m{sup 3}/day of the pumping/injection rate satisfies the energy requirements ({proportional_to}70 kW) for the factory. (author)
Quality Assurance for Thermal Hydraulic Analysis Code, TASS/SMR-S
International Nuclear Information System (INIS)
Kim, Hee Kyung; Kim, Soo Hyoung; Chung, Young Jong; Kim, Hyeon Soo
2012-01-01
Safety analysis for a System-integrated Modular Advanced Reactor (SMART), a computer code called TASS/SMR-S has been developed by Korea Atomic Energy Research Institute (KAERI). To guarantee the quality of the software, a series of software Quality Assurance (QA) procedures has been developed for the TASS/SMR-S code. These procedures are described herein, from the requirement phase to the Verification and Validation (V and V) phase, and representative results of the TASS/SMR-S QA are presented
Progress on China nuclear data processing code system
Liu, Ping; Wu, Xiaofei; Ge, Zhigang; Li, Songyang; Wu, Haicheng; Wen, Lili; Wang, Wenming; Zhang, Huanyu
2017-09-01
China is developing the nuclear data processing code Ruler, which can be used for producing multi-group cross sections and related quantities from evaluated nuclear data in the ENDF format [1]. The Ruler includes modules for reconstructing cross sections in all energy range, generating Doppler-broadened cross sections for given temperature, producing effective self-shielded cross sections in unresolved energy range, calculating scattering cross sections in thermal energy range, generating group cross sections and matrices, preparing WIMS-D format data files for the reactor physics code WIMS-D [2]. Programming language of the Ruler is Fortran-90. The Ruler is tested for 32-bit computers with Windows-XP and Linux operating systems. The verification of Ruler has been performed by comparison with calculation results obtained by the NJOY99 [3] processing code. The validation of Ruler has been performed by using WIMSD5B code.
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
Flexible hybrid energy cell for simultaneously harvesting thermal, mechanical, and solar energies.
Yang, Ya; Zhang, Hulin; Zhu, Guang; Lee, Sangmin; Lin, Zong-Hong; Wang, Zhong Lin
2013-01-22
We report the first flexible hybrid energy cell that is capable of simultaneously or individually harvesting thermal, mechanical, and solar energies to power some electronic devices. For having both the pyroelectric and piezoelectric properties, a polarized poly(vinylidene fluoride) (PVDF) film-based nanogenerator (NG) was used to harvest thermal and mechanical energies. Using aligned ZnO nanowire arrays grown on the flexible polyester (PET) substrate, a ZnO-poly(3-hexylthiophene) (P3HT) heterojunction solar cell was designed for harvesting solar energy. By integrating the NGs and the solar cells, a hybrid energy cell was fabricated to simultaneously harvest three different types of energies. With the use of a Li-ion battery as the energy storage, the harvested energy can drive four red light-emitting diodes (LEDs).
A Perspective of Energy Codes and Regulations for the Buildings of the Future
Energy Technology Data Exchange (ETDEWEB)
Rosenberg, Michael [Pacific Northwest National Laboratory,2032 Todd Street,Eugene, OR 97405e-mail: michael.rosenberg@pnnl.gov; Jonlin, Duane [Seattle Department ofConstruction and Inspections,P.O. Box 34019,Seattle, WA 98124e-mail: duane.jonlin@seattle.gov; Nadel, Steven [American Council for anEnergy-Efficient Economy,529 14th Street NW #600,Washington, DC 20045e-mail: snadel@aceee.org
2016-10-13
Today’s building energy codes focus on prescriptive requirements for features of buildings that are directly controlled by the design and construction teams and verifiable by municipal inspectors. Although these code requirements have had a significant impact, they fail to influence a large slice of the building energy use pie – including not only miscellaneous plug loads, cooking equipment and commercial/industrial processes, but the maintenance and optimization of the code-mandated systems as well. Currently, code compliance is verified only through the end of construction, and there are no limits or consequences for the actual energy use in an occupied building. In the future, our suite of energy regulations will likely expand to include building efficiency, energy use or carbon emission budgets over their full life cycle. Intelligent building systems, extensive renewable energy, and a transition from fossil fuel to electric heating systems will likely be required to meet ultra-low-energy targets. This paper lays out the authors’ perspectives on how buildings may evolve over the course of the 21st century and the roles that codes and regulations will play in shaping those buildings of the future.
Thermal-spectrum recriticality energetics
International Nuclear Information System (INIS)
Schwinkendorf, K.N.
1993-12-01
Large computer codes have been created in the past to predict the energy release in hypothetical core disruptive accidents (CDA), postulated to occur in liquid metal reactors (LMR). These codes, such as SIMMER, are highly specific to LMR designs. More recent attention has focused on thermal-spectrum criticality accidents, such as for fuel storage basins and waste tanks containing fissile material. This paper resents results from recent one-dimensional kinetics simulations, performed for a recriticality accident in a thermal spectrum. Reactivity insertion rates generally are smaller than in LMR CDAs, and the energetics generally are more benign. Parametric variation of input was performed, including reactivity insertion and initial temperature
ANTEO+: A subchannel code for thermal-hydraulic analysis of liquid metal cooled systems
Energy Technology Data Exchange (ETDEWEB)
Lodi, F., E-mail: francesco.lodi5@unibo.it [DIN – Laboratory of Montecuccolino, University of Bologna, Via dei Colli 16, 40136 Bologna (Italy); Grasso, G., E-mail: giacomo.grasso@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Mattioli, D., E-mail: davide.mattioli@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Sumini, M., E-mail: marco.sumini@unibo.it [DIN – Laboratory of Montecuccolino, University of Bologna, Via dei Colli 16, 40136 Bologna (Italy)
2016-05-15
Highlights: • The code structure is presented in detail. • The performed validation is outlined. • Results are critically discussed assessing code accuracy. • Conclusions are drawn and ground for future work identified. - Abstract: Liquid metal cooled fast reactors are promising options for achieving the high degrees of safety and sustainability demanded by the Generation IV paradigm. Among the critical aspects to be addressed in the design process, thermal-hydraulics is one of the most challenging; in order to embed safety in the core conceptualization, these aspects are to be considered at the very beginning of the design process, and translated in a design perspective. For achieving these objectives the subchannel code ANTEO+ has been conceived, able to simulate pin bundle arrangements cooled by liquid metals. The main purposes of ANTEO+ are simplifying the problem description maintaining the required accuracy, enabling a more transparent interface with the user, and having a clear and identifiable application domain, in order to help the user interpreting the results and, mostly, defining their confidence. Since ANTEO+ relies on empirical correlations, the validation phase is of paramount importance along with a clear discussion on the simplifications adopted in modeling the conservation equations. In the present work a detailed description of ANTEO+ structure is given along with a thorough validation of the main models implemented for flow split, pressure drops and subchannel temperatures. The analysis confirmed the ability of ANTEO+ in reproducing experimental data in its anticipated validity domain, with a relatively high degree of accuracy when compared to other classical subchannel tools like ENERGY-II, COBRA-IV-I-MIT and BRS-TVS.
Use of ETOG and ETOT computer codes for preparating the Library of LEOPARD with data from ENDFIB-IV
International Nuclear Information System (INIS)
Cunha Menezes Filho, A. da.
1983-01-01
The modifications carried out in the ETOT-3 and ETOG-3 computer codes used for preparating the thermal (172 energy groups) and epithermal (54 energy groups) libraries, respectivelly, of LEOPARD computer code, are presented. (M.C.K.) [pt
International Nuclear Information System (INIS)
Christensen, M.E.; Willoughby, O.H.
2009-01-01
EnergySolutions LLC is permitted by the State of Utah to treat organically-contaminated Mixed Waste by a vacuum thermal desorption (VTD) treatment process at its Clive, Utah treatment, storage, and disposal facility. The VTD process separates organics from organically-contaminated waste by heating the material in an inert atmosphere, and captures them as concentrated liquid by condensation. The majority of the radioactive materials present in the feed to the VTD are retained with the treated solids; the recovered aqueous and organic condensates are not radioactive. This is generally true when the radioactivity is present in solid form such as inorganic salts, metals or metallic oxides. The exception is when volatile radioactive materials are present such as radon gas, tritium, or carbon-14 organic chemicals. Volatile radioactive materials are a small fraction of the feed material. On August 28, 2006, EnergySolutions submitted a request to the USEPA for a variance to the Land Disposal Restrictions (LDR) standards for wastes designated with the combustion treatment code (CMBST). The final rule granting a site specific treatment variance was effective June 13, 2008. This variance is an alternative treatment standard to treatment by CMBST required for these wastes under USEPA's rules. The State of Utah provides oversight of the VTD processing operations. A demonstration test for treating CMBST-coded wastes was performed on April 29, 2008 through May 1, 2008. Three separate process cycles were conducted during this test. Both solid/liquid samples and emission samples were collected each day during the demonstration test. To adequately challenge the unit, feed material was spiked with trichloroethylene, o-cresol, dibenzofuran, and coal tar. Emission testing was conducted by EnergySolutions' emissions test contractor and sampling for radioactivity within the off-gas was completed by EnergySolutions' Health Physics department. This report discusses the emission testing
Directory of Open Access Journals (Sweden)
Sudarmono Sudarmono
2015-03-01
Full Text Available The failure of heat removal system of water-cooled reactor such as PWR in Three Mile Islands and Fukushima Daiichi BWR makes nuclear society starting to consider the use of high temperature gas-cooled reactor (HTGR. Reactor Physics and Technology Division – Center for Nuclear Reactor Safety and Technology (PTRKN has tasks to perform research and development on the conceptual design of cogeneration gas cooled reactor with medium power level of 200 MWt. HTGR is one of nuclear energy generation system, which has high energy efficiency, and has high and clean inherent safety level. The geometry and structure of the HTGR200 core are designed to produce the output of helium gas coolant temperature as high as 950 °C to be used for hydrogen production and other industrial processes in co-generative way. The output of very high temperature helium gas will cause thermal stress on the fuel pebble that threats the integrity of fission product confinement. Therefore, it is necessary to perform thermal-flow evaluation to determine the temperature distribution in the graphite and fuel pebble in the HTGR core. The evaluation was carried out by Thermix-Konvek module code that has been already integrated into VSOP'94 code. The HTGR core geometry was done using BIRGIT module code for 2-D model (RZ model with 5 channels of pebble flow in active core in the radial direction. The evaluation results showed that the highest and lowest temperatures in the reactor core are 999.3 °C and 886.5 °C, while the highest temperature of TRISO UO2 is 1510.20 °C in the position (z= 335.51 cm; r=0 cm. The analysis done based on reactor condition of 120 kg/s of coolant mass flow rate, 7 MPa of pressure and 200 MWth of power. Compared to the temperature distribution resulted between VSOP’94 code and fuel temperature limitation as high as 1600 oC, there is enough safety margin from melting or disintegrating. Keywords: Thermal-Flow, VSOP’94, Thermix-Konvek, HTGR, temperature
Thermal energy creation and transport and X-ray/EUV emission in a thermodynamic MHD CME simulation
Reeves, K.; Mikic, Z.; Torok, T.; Linker, J.; Murphy, N. A.
2017-12-01
We model a CME using the PSI 3D numerical MHD code that includes coronal heating, thermal conduction and radiative cooling in the energy equation. The magnetic flux distribution at 1 Rs is produced by a localized subsurface dipole superimposed on a global dipole field, mimicking the presence of an active region within the global corona. We introduce transverse electric fields near the neutral line in the active region to form a flux rope, then a converging flow is imposed that causes the eruption. We follow the formation and evolution of the current sheet and find that instabilities set in soon after the reconnection commences. We simulate XRT and AIA EUV emission and find that the instabilities manifest as bright features emanating from the reconnection region. We examine the quantities responsible for plasma heating and cooling during the eruption, including thermal conduction, radiation, adiabatic compression and expansion, coronal heating and ohmic heating due to dissipation of currents. We find that the adiabatic compression plays an important role in heating the plasma around the current sheet, especially in the later stages of the eruption when the instabilities are present. Thermal conduction also plays an important role in the transport of thermal energy away from the current sheet region throughout the reconnection process.
Baker, Karl W. (Inventor); Dustin, Miles O. (Inventor)
1992-01-01
A plurality of heat pipes in a shell receive concentrated solar energy and transfer the energy to a heat activated system. To provide for even distribution of the energy despite uneven impingement of solar energy on the heat pipes, absence of solar energy at times, or failure of one or more of the heat pipes, energy storage means are disposed on the heat pipes which extend through a heat pipe thermal coupling means into the heat activated device. To enhance energy transfer to the heat activated device, the heat pipe coupling cavity means may be provided with extensions into the device. For use with a Stirling engine having passages for working gas, heat transfer members may be positioned to contact the gas and the heat pipes. The shell may be divided into sections by transverse walls. To prevent cavity working fluid from collecting in the extensions, a porous body is positioned in the cavity.
Thermal energy storage using thermo-chemical heat pump
International Nuclear Information System (INIS)
Hamdan, M.A.; Rossides, S.D.; Haj Khalil, R.
2013-01-01
Highlights: ► Understanding of the performance of thermo chemical heat pump. ► Tool for storing thermal energy. ► Parameters that affect the amount of thermal stored energy. ► Lithium chloride has better effect on storing thermal energy. - Abstract: A theoretical study was performed to investigate the potential of storing thermal energy using a heat pump which is a thermo-chemical storage system consisting of water as sorbet, and sodium chloride as the sorbent. The effect of different parameters namely; the amount of vaporized water from the evaporator, the system initial temperature and the type of salt on the increase in temperature of the salt was investigated and hence on the performance of the thermo chemical heat pump. It was found that the performance of the heat pump improves with the initial system temperature, with the amount of water vaporized and with the water remaining in the system. Finally it was also found that lithium chloride salt has higher effect on the performance of the heat pump that of sodium chloride.
Thermal-Hydraulic Analysis of SWAMUP Facility Using ATHLET-SC Code
Energy Technology Data Exchange (ETDEWEB)
Wang, Zidi; Cao, Zhen; Liu, Xiaojing, E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shanghai (China)
2015-03-16
During the loss of coolant accident (LOCA) of supercritical water-cooled reactor (SCWR), the pressure in the reactor system will undergo a rapid decrease from the supercritical pressure to the subcritical condition. This process is called trans-critical transients, which is of crucial importance for the LOCA analysis of SCWR. In order to simulate the trans-critical transient, a number of system codes for SCWR have been developed up to date. However, the validation work for the trans-critical models in these codes is still missing. The test facility Supercritical WAter MUltiPurpose loop (SWAMUP) with 2 × 2 rod bundle in Shanghai Jiao Tong University (SJTU) will be applied to provide test data for code validation. Some pre-test calculations are important and necessary to show the feasibility of the experiment. In this study, trans-critical transient analysis is performed for the SWAMUP facility with the system code ATHLET-SC, which is modified in SJTU, for supercritical water system. This paper presents the system behavior, e.g., system pressure, coolant mass flow, cladding temperature during the depressurization. The effects of some important parameters such as heating power, depressurization rate on the system characteristics are also investigated in this paper. Additionally, some sensitivities study of the code models, e.g., heat transfer coefficient, critical heat flux correlation are analyzed and discussed. The results indicate that the revised system code ATHLET-SC is capable of simulating thermal-hydraulic behavior during the trans-critical transient. According to the results, the cladding temperature during the transient is kept at a low value. However, the pressure difference of the heat exchanger after depressurization could reach 6 MPa, which should be considered in the experiment.
International Nuclear Information System (INIS)
Auder, Benjamin
2011-01-01
This research thesis has been made within the frame of a project on nuclear reactor vessel life. It deals with the use of numerical codes aimed at estimating probability densities for every input parameter in order to calculate probability margins at the output level. More precisely, it deals with codes with one-dimensional functional responses. The author studies the numerical simulation of a pressurized thermal shock on a nuclear reactor vessel, i.e. one of the possible accident types. The study of the vessel integrity relies on a thermal-hydraulic analysis and on a mechanical analysis. Algorithms are developed and proposed for each of them. Input-output data are classified using a clustering technique and a graph-based representation. A method for output dimension reduction is proposed, and a regression is applied between inputs and reduced representations. Applications are discussed in the case of modelling and sensitivity analysis for the CATHARE code (a code used at the CEA for the thermal-hydraulic analysis)
International Nuclear Information System (INIS)
Tzika, F.; Stamatelatos, I.E.
2004-01-01
Thermal neutron self-shielding within large samples was studied using the Monte Carlo neutron transport code MCNP. The code enabled a three-dimensional modeling of the actual source and geometry configuration including reactor core, graphite pile and sample. Neutron flux self-shielding correction factors derived for a set of materials of interest for large sample neutron activation analysis are presented and evaluated. Simulations were experimentally verified by measurements performed using activation foils. The results of this study can be applied in order to determine neutron self-shielding factors of unknown samples from the thermal neutron fluxes measured at the surface of the sample
Research opportunities in salt hydrates for thermal energy storage
Braunstein, J.
1983-11-01
The state of the art of salt hydrates as phase change materials for low temperature thermal energy storage is reviewed. Phase equilibria, nucleation behavior and melting kinetics of the commonly used hydrate are summarized. The development of efficient, reliable inexpensive systems based on phase change materials, especially salt hydrates for the storage (and retrieval) of thermal energy for residential heating is outlined. The use of phase change material thermal energy storage systems is not yet widespread. Additional basic research is needed in the areas of crystallization and melting kinetics, prediction of phase behavior in ternary systems, thermal diffusion in salt hydrate systems, and in the physical properties pertinent to nonequilibrium and equilibrium transformations in these systems.
Survey of solar thermal energy storage subsystems for thermal/electric applications
Energy Technology Data Exchange (ETDEWEB)
Segaser, C. L.
1978-08-01
A survey of the current technology and estimated costs of subsystems for storing the thermal energy produced by solar collectors is presented. The systems considered were capable of producing both electricity and space conditioning for three types of loads: a single-family detached residence, an apartment complex of 100 units, and a city of 30,000 residents, containing both single-family residences and apartments. Collector temperatures will be in four ranges: (1) 100 to 250/sup 0/F (used for space heating and single-cycle air conditioners and organic Rankine low-temperature turbines); (2) 300 to 400/sup 0/F (used for dual-cycle air conditioners and low-temperature turbines); (3) 400 to 600/sup 0/F (using fluids from parabolic trough collectors to run Rankine turbines); (4) 800 to 1000/sup 0/F (using fluids from heliostats to run closed-cycle gas turbines and steam Rankine turbines). The solar thermal energy subsystems will require from 60 to 36 x 10/sup 5/ kWhr (2.05 x 10/sup 5/ to 1.23 x 10/sup 10/ Btu) of thermal storage capacity. In addition to sensible heat and latent heat storage materials, several other media were investigated as potential thermal energy storage materials, including the clathrate and semiclathrate hydrates, various metal hydrides, and heat storage based on inorganic chemical reactions.
Energy and exergy analyses of an ice-on-coil thermal energy storage system
International Nuclear Information System (INIS)
Ezan, Mehmet Akif; Erek, Aytunç; Dincer, Ibrahim
2011-01-01
In this study, energy and exergy analyses are carried out for the charging period of an ice-on-coil thermal energy storage system. The present model is developed using a thermal resistance network technique. First, the time-dependent variations of the predicted total stored energy, mass of ice, and outlet temperature of the heat transfer fluid from a storage tank are compared with the experimental data. Afterward, performance of an ice-on-coil type latent heat thermal energy storage system is investigated for several working and design parameters. The results of a comparative study are presented in terms of the variations of the heat transfer rate, total stored energy, dimensionless energetic/exergetic effectiveness and energy/exergy efficiency. The results indicate that working and design parameters of the ice-on-coil thermal storage tank should be determined by considering both energetic and exergetic behavior of the system. For the current parameters, storage capacity and energy efficiency of the system increases with decreasing the inlet temperature of the heat transfer fluid and increasing the length of the tube. Besides, the exergy efficiency increases with increasing the inlet temperature of the heat transfer fluid and increasing the length of the tube. -- Highlights: ► A comprehensive study on energy and exergy analyses of an ice-on-coil TES system. ► Determination of irreversibilities and their potential sources. ► Evaluation of both energy and exergy efficiencies and their comparisons.
International Nuclear Information System (INIS)
Delbecq, J.M.
1999-01-01
The Aster code is a 2D or 3D finite-element calculation code for structures developed by the R and D direction of Electricite de France (EdF). This dossier presents a complete overview of the characteristics and uses of the Aster code: introduction of version 4; the context of Aster (organisation of the code development, versions, systems and interfaces, development tools, quality assurance, independent validation); static mechanics (linear thermo-elasticity, Euler buckling, cables, Zarka-Casier method); non-linear mechanics (materials behaviour, big deformations, specific loads, unloading and loss of load proportionality indicators, global algorithm, contact and friction); rupture mechanics (G energy restitution level, restitution level in thermo-elasto-plasticity, 3D local energy restitution level, KI and KII stress intensity factors, calculation of limit loads for structures), specific treatments (fatigue, rupture, wear, error estimation); meshes and models (mesh generation, modeling, loads and boundary conditions, links between different modeling processes, resolution of linear systems, display of results etc..); vibration mechanics (modal and harmonic analysis, dynamics with shocks, direct transient dynamics, seismic analysis and aleatory dynamics, non-linear dynamics, dynamical sub-structuring); fluid-structure interactions (internal acoustics, mass, rigidity and damping); linear and non-linear thermal analysis; steels and metal industry (structure transformations); coupled problems (internal chaining, internal thermo-hydro-mechanical coupling, chaining with other codes); products and services. (J.S.)
Thermal Comfort and Strategies for Energy Conservation.
Rohles, Frederick H., Jr.
1981-01-01
Discusses studies in thermal comfort which served as the basis for the comfort standard. Examines seven variables in the human response to the thermal environment in terms of the ways in which they can be modified to conserve energy. (Author/MK)
Investigation of Solar and Solar-Gas Thermal Energy Sources
Ivan Herec; Jan Zupa
2003-01-01
The article deals with the investigation of solar thermal sources of electrical and heat energy as well as the investigation of hybrid solar-gas thermal sources of electrical and heat energy (so called photothermal sources). Photothermal sources presented here utilize computer-controlled injection of the conversion fluid into special capillary porous substance that is adjusted to direct temperature treatment by the concentrated thermal radiation absorption.
DEFF Research Database (Denmark)
Kazanci, Ongun Berk; Olesen, Bjarne W.
2015-01-01
indoor environment. For the energy consumption of the HVAC system, air-to-brine heat pump, mixing station and controller of the radiant floor, and the air handling unit were considered. The measurements were analyzed based on the achieved indoor environment category (according to EN 15251...... the floor cooling system) and increasing the ventilation rate provided a better thermal indoor environment but with increased energy consumption. The thermal indoor environment and energy performance of the house can be improved with decreased glazing area, increased thermal mass, installation of solar...
Characteristics Study of Photovoltaic Thermal System with Emphasis on Energy Efficiency
Directory of Open Access Journals (Sweden)
Yong Chuah Yee
2018-01-01
Full Text Available Solar energy is typically collected through photovoltaic (PV to generate electricity or through thermal collectors as heat energy, they are generally utilised separately. This project is done with the purpose of integrating the two systems to improve the energy efficiency. The idea of this photovoltaic-thermal (PVT setup design is to simultaneously cool the PV panel so it can operate at a lower temperature thus higher electrical efficiency and also store the thermal energy. The experimental data shows that the PVT setup increased the electrical efficiency of the standard PV setup from 1.64% to 2.15%. The integration of the thermal collector also allowed 37.25% of solar energy to be stored as thermal energy. The standard PV setup harnessed only 1.64% of the solar energy, whereas the PVT setup achieved 39.4%. Different flowrates were tested to determine its effects on the PVT setup’s electrical and thermal efficiency. The various flowrate does not significantly impact the electrical efficiency since it did not significantly impact the cooling of the panel. The various flowrates resulted in fluctuating thermal efficiencies, the relation between the two is inconclusive in this project.
Energy Technology Data Exchange (ETDEWEB)
Alkan, Cemil; Sari, Ahmet; Karaipekli, Ali [Department of Chemistry, Gaziosmanpasa University, 60240 Tokat (Turkey); Uzun, Orhan [Department of Physics, Gaziosmanpasa University, 60240 Tokat (Turkey)
2009-01-15
This study is focused on the preparation, characterization, and determination of thermal properties of microencapsulated docosane with polymethylmethacrylate (PMMA) as phase change material for thermal energy storage. Microencapsulation of docosane has been carried out by emulsion polymerization. The microencapsulated phase change material (MEPCM) was characterized using scanning electron microscopy (SEM) and Fourier transform infrared (FT-IR) spectroscopy. Thermal properties and thermal stability of MEPCM were measured by differential scanning calorimetry (DSC) and thermal gravimetric analysis (TGA). DSC analysis indicated that the docosane in the microcapsules melts at 41.0 C and crystallizes at 40.6 C. It has latent heats of 54.6 and -48.7 J/g for melting and crystallization, respectively. TGA showed that the MEPCM degraded in three distinguishable steps and had good chemical stability. Accelerated thermal cycling tests also indicated that the MEPCM had good thermal reliability. Based on all these results, it can be concluded that the microencapsulated docosane as MEPCMs have good potential for thermal energy storage purposes such as solar space heating applications. (author)
Energy Technology Data Exchange (ETDEWEB)
Chung, Ji Bum [Institute for Advanced Engineering, Yongin (Korea, Republic of); Park, Jong Woon [Korea Electric Power Research Institute, Taejon (Korea, Republic of)
1998-12-31
In order to enhance the dynamic and interactive simulation capability of a system thermal hydraulic code for nuclear power plant, applicability of flow network models in SINDA/FLUINT{sup TM} has been tested by modeling feedwater system and coupling to DSNP which is one of a system thermal hydraulic simulation code for a pressurized heavy water reactor. The feedwater system is selected since it is one of the most important balance of plant systems with a potential to greatly affect the behavior of nuclear steam supply system. The flow network model of this feedwater system consists of condenser, condensate pumps, low and high pressure heaters, deaerator, feedwater pumps, and control valves. This complicated flow network is modeled and coupled to DSNP and it is tested for several normal and abnormal transient conditions such turbine load maneuvering, turbine trip, and loss of class IV power. The results show reasonable behavior of the coupled code and also gives a good dynamic and interactive simulation capabilities for the several mild transient conditions. It has been found that coupling system thermal hydraulic code with a flow network code is a proper way of upgrading simulation capability of DSNP to mature nuclear plant analyzer (NPA). 5 refs., 10 figs. (Author)
Development of An Automatic Verification Program for Thermal-hydraulic System Codes
Energy Technology Data Exchange (ETDEWEB)
Lee, J. Y.; Ahn, K. T.; Ko, S. H.; Kim, Y. S.; Kim, D. W. [Pusan National University, Busan (Korea, Republic of); Suh, J. S.; Cho, Y. S.; Jeong, J. J. [System Engineering and Technology Co., Daejeon (Korea, Republic of)
2012-05-15
As a project activity of the capstone design competitive exhibition, supported by the Education Center for Green Industry-friendly Fusion Technology (GIFT), we have developed a computer program which can automatically perform non-regression test, which is needed repeatedly during a developmental process of a thermal-hydraulic system code, such as the SPACE code. A non-regression test (NRT) is an approach to software testing. The purpose of the non-regression testing is to verify whether, after updating a given software application (in this case, the code), previous software functions have not been compromised. The goal is to prevent software regression, whereby adding new features results in software bugs. As the NRT is performed repeatedly, a lot of time and human resources will be needed during the development period of a code. It may cause development period delay. To reduce the cost and the human resources and to prevent wasting time, non-regression tests need to be automatized. As a tool to develop an automatic verification program, we have used Visual Basic for Application (VBA). VBA is an implementation of Microsoft's event-driven programming language Visual Basic 6 and its associated integrated development environment, which are built into most Microsoft Office applications (In this case, Excel)
Development of An Automatic Verification Program for Thermal-hydraulic System Codes
International Nuclear Information System (INIS)
Lee, J. Y.; Ahn, K. T.; Ko, S. H.; Kim, Y. S.; Kim, D. W.; Suh, J. S.; Cho, Y. S.; Jeong, J. J.
2012-01-01
As a project activity of the capstone design competitive exhibition, supported by the Education Center for Green Industry-friendly Fusion Technology (GIFT), we have developed a computer program which can automatically perform non-regression test, which is needed repeatedly during a developmental process of a thermal-hydraulic system code, such as the SPACE code. A non-regression test (NRT) is an approach to software testing. The purpose of the non-regression testing is to verify whether, after updating a given software application (in this case, the code), previous software functions have not been compromised. The goal is to prevent software regression, whereby adding new features results in software bugs. As the NRT is performed repeatedly, a lot of time and human resources will be needed during the development period of a code. It may cause development period delay. To reduce the cost and the human resources and to prevent wasting time, non-regression tests need to be automatized. As a tool to develop an automatic verification program, we have used Visual Basic for Application (VBA). VBA is an implementation of Microsoft's event-driven programming language Visual Basic 6 and its associated integrated development environment, which are built into most Microsoft Office applications (In this case, Excel)
Development of best estimate auditing code for CANDU thermal hydraulic safety analysis
International Nuclear Information System (INIS)
Hwnag, M.
2001-04-01
The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a fourth step of the whole project, applying the RELAP5/MOD3/CANDU+ version for the real CANDU plant LOCA Analysis and D2O leakage incident. There are three main models under investigation, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs, especially when CANDU LOCA is tested. Also, for Wolsung unit 1 D2O leakage incident analysis, the plant behavior is predicited with the newly developed version for the first 1000 seconds after onset of the incident, with the main interest aiming for system pressure, level control system, and thermal hydraulic transient behavior of the secondary system. The model applided for this particular application includes heat transfer model of nuclear fuel assembly, decay heat model, and MOV (Motor Operated Valve) model. Finally, the code maintenance work, mainly correcting the known errors, is presented
Kinetic---a system code for analyzing nuclear thermal propulsion rocket engine transients
International Nuclear Information System (INIS)
Schmidt, E.; Lazareth, O.; Ludewig, H.
1993-01-01
A system code suitable for analyzing Nuclear Thermal Propulsion (NTP) rocket engines is described in this paper. The code consists of a point reactor model and nodes to describe the fluid dynamics and heat transfer mechanism. Feedback from the fuel, coolant, moderator and reflector are allowed for, and the control of the reactor is by motion of controls element (drums or rods). The worth of the control element and feedback coefficients are predetermined. Separate models for the turbo-pump assembly (TPA) and nozzle are also included. The model to be described in this paper is specific for the Particle Bed Reactor (PBR). An illustrative problem is solved. This problem consists of a PBR operating in a blowdown mode
Kinetic—a system code for analyzing nuclear thermal propulsion rocket engine transients
Schmidt, Eldon; Lazareth, Otto; Ludewig, Hans
1993-01-01
A system code suitable for analyzing Nuclear Thermal Propulsion (NTP) rocket engines is described in this paper. The code consists of a point reactor model and nodes to describe the fluid dynamics and heat transfer mechanism. Feedback from the fuel, coolant, moderator and reflector are allowed for, and the control of the reactor is by motion of controls element (drums or rods). The worth of the control element and feedback coefficients are predetermined. Separate models for the turbo-pump assembly (TPA) and nozzle are also included. The model to be described in this paper is specific for the Particle Bed Reactor (PBR). An illustrative problem is solved. This problem consists of a PBR operating in a blowdown mode.
KINETIC: A system code for analyzing Nuclear thermal propulsion rocket engine transients
Schmidt, E.; Lazareth, O.; Ludewig, H.
1993-07-01
A system code suitable for analyzing Nuclear Thermal Propulsion (NTP) rocket engines is described in this paper. The code consists of a point reactor model and nodes to describe the fluid dynamics and heat transfer mechanism. Feedback from the fuel coolant, moderator and reflector are allowed for, and the control of the reactor is by motion of control elements (drums or rods). The worth of the control clement and feedback coefficients are predetermined. Separate models for the turbo-pump assembly (TPA) and nozzle are also included. The model to be described in this paper is specific for the Particle Bed Reactor (PBR). An illustrative problem is solved. This problem consists of a PBR operating in a blowdown mode.
CONTRANS 2 code conversion from Apollo to HP
International Nuclear Information System (INIS)
Lee, Hae Cho
1996-01-01
CONTRANS2 computer code is used to calculate transient thermal hydraulic responses of containment building to loss of coolant and main steam line break accident. Mass and energy release to the containment following an accident are code inputs. This report firstly describes detailed work carried out for installation of CONTRANS2 on Apollo DN10000 and code validation results after installation. Secondly, A series of work is also describes in relation to installation of CONTRANS2 on HP 9000/700 series as well as relevant code validation results. Attached is a report on software verification and validation results. 7 refs. (Author) .new
Mother code specifications (Appendix to CEA report 2472)
International Nuclear Information System (INIS)
Pillard, Denise; Soule, Jean-Louis
1964-12-01
The Mother code (written in Fortran for IBM 7094) computes the integral cross section and the first two moments of energy transfer of a thermalizer. Computation organisation and methods are presented in an other document. This document presents code specifications, i.e. input data (for spectrum description, printing options, input record formats, conditions to be met by values), and results (printing formats and options, writing and punching options and formats)
Fluctuation and thermal energy balance for drift-wave turbulence
International Nuclear Information System (INIS)
Kim, Chang-Bae; Horton, W.
1990-05-01
Energy conservation for the drift-wave system is shown to be separated into the wave-energy power balance equation and an ambient thermal-energy transport equation containing the anomalous transport fluxes produced by the fluctuations. The wave energy equation relates the wave energy density and wave energy flux to the anomalous transport flux and the dissipation of the fluctuations. The thermal balance equation determines the evolution of the temperature profiles from the divergence of the anomalous heat flux, the collisional heating and cooling mechanisms and the toroidal pumping effect. 16 refs., 1 tab
Fluctuation and thermal energy balance for drift-wave turbulence
International Nuclear Information System (INIS)
Changbae Kim; Horton, W.
1991-01-01
Energy conservation for the drift-wave system is shown to be separated into the wave-energy power balance equation and an ambient thermal-energy transport equation containing the anomalous transport fluxes produced by the fluctuations. The wave energy equation relates the wave energy density and wave energy flux to the anomalous transport flux and the dissipation of the fluctuations. The thermal balance equation determines the evolution of the temperature profiles from the divergence of the anomalous heat flux, the collisional heating and cooling mechanisms and the toroidal pumping effect. (author)
International Nuclear Information System (INIS)
Khattab, K.; Omar, H.; Ghazi, N.
2009-01-01
A 3-D (R, θ , Z) neutronic model for the Miniature Neutron Source Reactor (MNSR) was developed earlier to conduct the reactor neutronic analysis. The group constants for all the reactor components were generated using the WIMSD4 code. The reactor excess reactivity and the four group neutron flux distributions were calculated using the CITATION code. This model is used in this paper to calculate the point wise four energy group neutron flux distributions in the MNSR versus the radius, angle and reactor axial directions. Good agreement is noticed between the measured and the calculated thermal neutron flux in the inner and the outer irradiation site with relative difference less than 7% and 5% respectively. (author)
Potential Job Creation in Rhode Island as a Result of Adopting New Residential Building Energy Codes
Energy Technology Data Exchange (ETDEWEB)
Scott, Michael J.; Niemeyer, Jackie M.
2013-09-01
Are there advantages to states that adopt the most recent model building energy codes other than saving energy? For example, can the construction activity and energy savings associated with code-compliant housing units become significant sources of job creation for states if new building energy codes are adopted to cover residential construction? , The U.S. Department of Energy (DOE) Building Energy Codes Program (BECP) asked Pacific Northwest National Laboratory (PNNL) to research and ascertain whether jobs would be created in individual states based on their adoption of model building energy codes. Each state in the country is dealing with high levels of unemployment, so job creation has become a top priority. Many programs have been created to combat unemployment with various degrees of failure and success. At the same time, many states still have not yet adopted the most current versions of the International Energy Conservation Code (IECC) model building energy code, when doing so could be a very effective tool in creating jobs to assist states in recovering from this economic downturn.
Energy Technology Data Exchange (ETDEWEB)
Costa, J [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Grenoble, Grenoble (France)
1983-07-01
The purpose of the meeting was to discuss and exchange views on thermal stratification existing in sodium of the main vessel, secondary circuits and large components of LMFBRs under various operating conditions. The meeting was divided into four sessions: national position presentations; fundamental studies on theory and application of stratification problems, numerical and experimental investigations applied to stratified flow phenomena; computer codes for evaluation of thermal stratification; applied studies covering the computer codes and experimental studies for prediction of temperature velocity field.
International Nuclear Information System (INIS)
Costa, J.
1983-07-01
The purpose of the meeting was to discuss and exchange views on thermal stratification existing in sodium of the main vessel, secondary circuits and large components of LMFBRs under various operating conditions. The meeting was divided into four sessions: national position presentations; fundamental studies on theory and application of stratification problems, numerical and experimental investigations applied to stratified flow phenomena; computer codes for evaluation of thermal stratification; applied studies covering the computer codes and experimental studies for prediction of temperature velocity field
The role of Solar thermal in Future Energy Systems
DEFF Research Database (Denmark)
Mathiesen, Brian Vad; Hansen, Kenneth
This report deals with solar thermal technologies and investigates possible roles for solar thermal in future energy systems for four national energy systems; Germany, Austria, Italy and Denmark. The project period started in January 2014 and finished by October 2017. This report is based...
International Nuclear Information System (INIS)
Pleskunas, R.J.
2015-01-01
In response to the Fukushima Dai-ichi beyond design basis accident in March 2011, the Nuclear Regulatory Commission (NRC) issued Order EA-12-049, 'Issuance of Order to Modify Licenses with Regard to Requirements for Mitigation Strategies Beyond-Design-Basis-External-Events'. To outline the process to be used by individual licensees to define and implement site-specific diverse and flexible mitigation strategies (FLEX) that reduce the risks associated with beyond design basis conditions, Nuclear Energy Institute document NEI 12-06, 'Diverse and Flexible Coping Strategies (FLEX) Implementation Guide', was issued. A beyond design basis external event (BDBEE) is postulated to cause an Extended Loss of AC Power (ELAP), which will result in a loss of ventilation which has the potential to impact room habitability and equipment operability. During the ELAP, portable FLEX equipment will be used to achieve and maintain safe shutdown, and only a minimal set of instruments and controls will be available. Given these circumstances, analysis is required to determine the environmental conditions in several vital areas of the Nuclear Power Plant. The BDBEE mitigating strategies require certain room environments to be maintained such that they can support the occupancy of personnel and the functionality of equipment located therein, which is required to support the strategies associated with compliance to NRC Order EA-12-049. Three thermal-hydraulic analyses of vital areas during an extended loss of AC power using the GOTHIC computer code will be presented: 1) Safety-related pump and instrument room transient analysis; 2) Control Room transient analysis; and 3) Auxiliary/Control Building transient analysis. GOTHIC (Generation of Thermal-Hydraulic Information for Containment) is a general purpose thermal-hydraulics software package for the analysis of nuclear power plant containments, confinement buildings, and system components. It is a volume/path/heat sink
HEATING6 analysis of international thermal benchmark problem sets 1 and 2
International Nuclear Information System (INIS)
Childs, K.W.; Bryan, C.B.
1986-10-01
In order to assess the heat transfer computer codes used in the analysis of nuclear fuel shipping casks, the Nuclear Energy Agency Committee on Reactor Physics has defined seven problems for benchmarking thermal codes. All seven of these problems have been solved using the HEATING6 heat transfer code. This report presents the results of five of the problems. The remaining two problems were used in a previous benchmarking of thermal codes used in the United States, and their solutions have been previously published
One-dimensional modeling of thermal energy produced in a seismic fault
Konga, Guy Pascal; Koumetio, Fidèle; Yemele, David; Olivier Djiogang, Francis
2017-12-01
Generally, one observes an anomaly of temperature before a big earthquake. In this paper, we established the expression of thermal energy produced by friction forces between the walls of a seismic fault while considering the dynamic of a one-dimensional spring-block model. It is noted that, before the rupture of a seismic fault, displacements are caused by microseisms. The curves of variation of this thermal energy with time show that, for oscillatory and aperiodic displacement, the thermal energy is accumulated in the same way. The study reveals that thermal energy as well as temperature increases abruptly after a certain amount of time. We suggest that the corresponding time is the start of the anomaly of temperature observed which can be considered as precursory effect of a big seism. We suggest that the thermal energy can heat gases and dilate rocks until they crack. The warm gases can then pass through the cracks towards the surface. The cracks created by thermal energy can also contribute to the rupture of the seismic fault. We also suggest that the theoretical model of thermal energy, produced in seismic fault, associated with a large quantity of experimental data may help in the prediction of earthquakes.
International Nuclear Information System (INIS)
Vahid-Pakdel, M.J.; Nojavan, Sayyad; Mohammadi-ivatloo, B.; Zare, Kazem
2017-01-01
Highlights: • Studying heating market impact on energy hub operation considering price uncertainty. • Investigating impact of implementation of heat demand response on hub operation. • Presenting stochastic method to consider wind generation and prices uncertainties. - Abstract: Multi carrier energy systems or energy hubs has provided more flexibility for energy management systems. On the other hand, due to mutual impact of different energy carriers in energy hubs, energy management studies become more challengeable. The initial patterns of energy demands from grids point of view can be modified by optimal scheduling of energy hubs. In this work, optimal operation of multi carrier energy system has been studied in the presence of wind farm, electrical and thermal storage systems, electrical and thermal demand response programs, electricity market and thermal energy market. Stochastic programming is implemented for modeling the system uncertainties such as demands, market prices and wind speed. It is shown that adding new source of heat energy for providing demand of consumers with market mechanism changes the optimal operation point of multi carrier energy system. Presented mixed integer linear formulation for the problem has been solved by executing CPLEX solver of GAMS optimization software. Simulation results shows that hub’s operation cost reduces up to 4.8% by enabling the option of using thermal energy market for meeting heat demand.
PERCON: A flexible computer code for detailed thermal performance studies
International Nuclear Information System (INIS)
Boardman, F.B.; Collier, W.D.
1975-07-01
PERCON is a computer code which evaluates temperatures in three dimensions for a block containing heat sources and having coolant flow in one dimension. The solution is obtained at successive planes perpendicular to the coolant flow and the progression from one plane to the next occurs by the heat to the coolant determining convective boundary conditions at the next plane after due allowance being made for any lateral mixing or mass transfer between coolants. It is also possible to calculate the diametral change along a radius as a function of irradiation shrinkage and thermal expansion. This is used in a 'through life' calculation which evalates interaction pressure in tubular fuel elements. Physical property data used by the code may be specified as functions of temperature. The coolant flow may be specified, or alternatively derived by the program to satisfy either a specified overall pressure drop or mixed mean temperature rise. The pressure drop through each coolant is calculated and the flow modified, followed by a repeat of the temperature calculation, until the pressure imbalance between chosen flow channels at chosen axial positions is less than the specified convergence limit. A detailed description of the facilities in the code is given and some cases which have been studied are discussed. (U.K.)
Steady state thermal hydraulic analysis of LMR core using COBRA-K code
Energy Technology Data Exchange (ETDEWEB)
Kim, Eui Kwang; Kim, Young Gyun; Kim Young In; Kim Young Cheol
1997-02-01
A thermal hydraulics analysis code COBRA-K is being developed by the KAERI LMR core design technology development team. COBRA-K is a part of the integrated computation system for LMR core design and analysis, the K-CORE system. COBRA-K is supposed to predict the flow and temperature distributions in LMR core. COBRA-K is an extension of the previously published COBRA-IV-I code with several functional improvements. Specially COBRA-K has been improved to analyze single and multi-assembly, and whole-core in the transient condition. This report describes the overall features of COBRA-K and gives general input descriptions. The 19 pin assembly experimental data of ORNL were used to verify the accuracy of this code for the steady state analysis. The comparative results show good agreements between the calculated and the measured data. And COBRA-K can be used to predict flow and temperature distributions for the LMR core design. (author). 7 refs., 6 tabs., 13 figs.
International Nuclear Information System (INIS)
2012-01-01
Whereas thermal renewable energies are to become inescapable, and notably 'green heat' which is acclaimed by real estate professionals as well as by industries, their market is foreseen to grow at a rate of 6 per cent a year until 2015. This high rate is notably due to the soaring price of conventional energies like electricity, gas or oil fuel, but also to environmental constraints related to the reduction of greenhouse gas emissions. A first part proposes an overview of the French market of new sources of thermal renewable energies for a domestic use in 2011, and discusses perspectives by 2015. A detailed analysis of the three main technologies (heat pumps, thermal solar devices, wood fuelled domestic heating devices) is proposed and challenges faced by involved enterprises and possible answers provided by professionals are also detailed. A second part gathers and comments data related to thermal energy production for industrial and collective use (in collective housing and office building): energy production level, legal and regulatory framework, evolution of demand, predictions for the different energy sources (wood energy, geothermal, waste energetic valorisation). It also proposes an analysis of stakes related to these applications. The third part proposes an assessment of the size of the different sectors by presenting key economic figures (turnover, staff, etc.). While the fourth part proposes an overview of leaders for each sector (thermal solar, biomass, and heat pump) and a more detailed presentation of 14 important actors, the fifth and last part proposes a large set of financial and economic indicators of 200 involved operators
Energy loss and thermalization of low-energy electrons
International Nuclear Information System (INIS)
LaVerne, J.A.; Mozumder, A.; Notre Dame Univ., IN
1984-01-01
Various processes involved in the moderation of low-energy electrons (< 10 keV in energy) have been delineated in gaseous and liquid media. The discussion proceeds in two stages. The first stage ends and the second stage begins when the electron energy equals the first excitation potential of the medium. The second stage ends with thermalization. Cross sections for electronic excitation and for the excitation (and de-excitation) of sub-electronic processes have been evaluated and incorporated in suitable stopping power and transport theories. Comparison between experiment and theory and intercomparisons between theories and experiments have been provided where possible. (author)
Energy and Power Measurements for Network Coding in the Context of Green Mobile Clouds
DEFF Research Database (Denmark)
Paramanathan, Achuthan; Pedersen, Morten Videbæk; Roetter, Daniel Enrique Lucani
2013-01-01
results for inter-session network coding in Open-Mesh routers underline that the energy invested in performing network coding pays off by dramatically reducing the total energy for the transmission of data over wireless links. We also show measurements for intra-session network coding in three different......This paper presents an in-depth power and energy measurement campaign for inter- and intra-session network coding enabled communication in mobile clouds. The measurements are carried out on different commercial platforms with focus on routers and mobile phones with different CPU capabilities. Our...
Myo-inositol based nano-PCM for solar thermal energy storage
International Nuclear Information System (INIS)
Singh, D.K.; Suresh, S.; Singh, H.; Rose, B.A.J.; Tassou, S.; Anantharaman, N.
2017-01-01
Highlights: • Properties of Myo-Inositol laden with Al_2O_3 and CuO nanoparticles was studied. • The melting point was found to increase for MI-A and decrease for MI-C. • MI interacted only physically on addition of NPs. • Mass changes were <3% after thermal cycling of MI-A and MI-C. • MI-A is more suited for thermal energy storage than MI-C. - Abstract: The thermo-physical behavior of Myo-Inositol (MI), (a sugar alcohol), was investigated as a potential material for developing more compact solar thermal energy storage systems than those currently available. This latent heat storage medium could be utilized for commercial and industrial applications using solar thermal energy storage in the temperature range of 160–260 °C, if its thermal performance was modified. The objective of this investigation was to determine via experimentation, if Al_2O_3 and CuO nanoparticles dispersed in pure MI for mixtures of 1, 2 and 3% (by weight) improved the thermal performance of MI for solar thermal energy systems. Nanoparticles only physically interacted with MI, and not chemically, even after 50 thermal cycles. The distribution of CuO nanoparticles in the nano-PCM was found to be more uniform than alumina nanoparticles. After cycling, nano-MIs studied here suffered a lower decrease in heat of fusion than pure MI, which makes nano-MIs more suitable for solar thermal storage applications at 160–260 °C. Between CuO and Al_2O_3 nanoparticles, latter was found to be more suitable for compact solar thermal energy storage owing to an increase in melting point observed.
Energy Technology Data Exchange (ETDEWEB)
Hainoun, A., E-mail: pscientific2@aec.org.sy [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Doval, A. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400 S.C de Bariloche, Rio Negro (Argentina); Umbehaun, P. [Centro de Engenharia Nuclear – CEN, IPEN-CNEN/SP, Av. Lineu Prestes 2242-Cidade Universitaria, CEP-05508-000 São Paulo, SP (Brazil); Chatzidakis, S. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Ghazi, N. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Park, S. [Research Reactor Design and Engineering Division, Basic Science Project Operation Dept., Korea Atomic Energy Research Institute (Korea, Republic of); Mladin, M. [Institute for Nuclear Research, Campului Street No. 1, P.O. Box 78, 115400 Mioveni, Arges (Romania); Shokr, A. [Division of Nuclear Installation Safety, Research Reactor Safety Section, International Atomic Energy Agency, A-1400 Vienna (Austria)
2014-12-15
Highlights: • A set of advanced system thermal hydraulic codes are benchmarked against IFA of IEA-R1. • Comparative safety analysis of IEA-R1 reactor during LOFA by 7 working teams. • This work covers both experimental and calculation effort and presents new out findings on TH of RR that have not been reported before. • LOFA results discrepancies from 7% to 20% for coolant and peak clad temperatures are predicted conservatively. - Abstract: In the framework of the IAEA Coordination Research Project on “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal hydraulic computational methods and tools for operation and safety analysis of research reactors” the Brazilian research reactor IEA-R1 has been selected as reference facility to perform benchmark calculations for a set of thermal hydraulic codes being widely used by international teams in the field of research reactor (RR) deterministic safety analysis. The goal of the conducted benchmark is to demonstrate the application of innovative reactor analysis tools in the research reactor community, validation of the applied codes and application of the validated codes to perform comprehensive safety analysis of RR. The IEA-R1 is equipped with an Instrumented Fuel Assembly (IFA) which provided measurements for normal operation and loss of flow transient. The measurements comprised coolant and cladding temperatures, reactor power and flow rate. Temperatures are measured at three different radial and axial positions of IFA summing up to 12 measuring points in addition to the coolant inlet and outlet temperatures. The considered benchmark deals with the loss of reactor flow and the subsequent flow reversal from downward forced to upward natural circulation and presents therefore relevant phenomena for the RR safety analysis. The benchmark calculations were performed independently by the participating teams using different thermal hydraulic and safety
Applications of the thermit code to 3D thermal hydraulic analysis of LWR cores
International Nuclear Information System (INIS)
Reed, W.H.
1979-01-01
The THERMIT code calculates the three-dimensional transient thermal hydraulic behavior of light water reactor cores. Its two-fluid dynamics equations for two-phase flow offer improved physical modelling capability needed in the context of calculation coupled to neutron kinetics for feedback. The numerical fluid dynamics method was chosen for reliability over a wider range of transients. An improved heat transfer numerical method is presented which gives better numerical stability and accuracy. A number of example calculations are discussed which give an idea of the power and flexibility of the code
Thermal energy storage for smart grid applications
Al-Hallaj, Said; Khateeb, Siddique; Aljehani, Ahmed; Pintar, Mike
2018-01-01
Energy consumption for commercial building cooling accounts for 15% of all commercial building's electricity usage [1]. Electric utility companies charge their customers time of use consumption charges (/kWh) and additionally demand usage charges (/kW) to limit peak energy consumption and offset their high operating costs. Thus, there is an economic incentive to reduce both the electricity consumption charges and demand charges by developing new energy efficient technologies. Thermal energy storage (TES) systems using a phase change material (PCM) is one such technology that can reduce demand charges and shift the demand from on-peak to off-peak rates. Ice and chilled water have been used in thermal storage systems for many decades, but they have certain limitations, which include a phase change temperature of 0 degrees Celsius and relatively low thermal conductivity in comparison to other materials, which limit their applications as a storage medium. To overcome these limitations, a novel phase change composite (PCC) TES material was developed that has much higher thermal conductivity that significantly improves the charge / discharge rate and a customizable phase change temperature to allow for better integration with HVAC systems. Compared to ice storage, the PCC TES system is capable of very high heat transfer rate and has lower system and operational costs. Economic analysis was performed to compare the PCC TES system with ice system and favorable economics was proven. A 4.5 kWh PCC TES prototype system was also designed for testing and validation purpose.
International Nuclear Information System (INIS)
2005-01-01
A - Description of program or function: (1) Problems to be solved: MVP/GMVP can solve eigenvalue and fixed-source problems. The multigroup code GMVP can solve forward and adjoint problems for neutron, photon and neutron-photon coupled transport. The continuous-energy code MVP can solve only the forward problems. Both codes can also perform time-dependent calculations. (2) Geometry description: MVP/GMVP employs combinatorial geometry to describe the calculation geometry. It describes spatial regions by the combination of the 3-dimensional objects (BODIes). Currently, the following objects (BODIes) can be used. - BODIes with linear surfaces: half space, parallelepiped, right parallelepiped, wedge, right hexagonal prism; - BODIes with quadratic surface and linear surfaces: cylinder, sphere, truncated right cone, truncated elliptic cone, ellipsoid by rotation, general ellipsoid; - Arbitrary quadratic surface and torus. The rectangular and hexagonal lattice geometry can be used to describe the repeated geometry. Furthermore, the statistical geometry model is available to treat coated fuel particles or pebbles for high temperature reactors. (3) Particle sources: The various forms of energy-, angle-, space- and time-dependent distribution functions can be specified. (4) Cross sections: The ANISN-type PL cross sections or the double-differential cross sections can be used in the multigroup code GMVP. On the other hand, the specific cross section libraries are used in the continuous-energy code MVP. The libraries are generated from the evaluated nuclear data (JENDL-3.3, ENDF/B-VI, JEF-3.0 etc.) by using the LICEM code. The neutron cross sections in the unresolved resonance region are described by the probability table method. The neutron cross sections at arbitrary temperatures are available for MVP by just specifying the temperatures in the input data. (5) Boundary conditions: Vacuum, perfect reflective, isotropic reflective (white), periodic boundary conditions can be
Thermoelectric cooling in combination with photovoltaics and thermal energy storage
Directory of Open Access Journals (Sweden)
Skovajsa Jan
2017-01-01
Full Text Available The article deals with the use of modern technologies that can improve the thermal comfort in buildings. The article describes the usage of thermal energy storage device based on the phase change material (PCM. The technology improves the thermal capacity of the building and it is possible to use it for active heating and cooling. It is designed as a “green technology” so it is able to use renewable energy sources, e.g., photovoltaic panels, solar thermal collectors, and heat pump. Moreover, an interesting possibility is the ability to use thermal energy storage in combination with a photovoltaic system and thermoelectric coolers. In the research, there were made measurements of the different operating modes and the results are presented in the text.
Buffer thermal energy storage for an air Brayton solar engine
Strumpf, H. J.; Barr, K. P.
1981-01-01
The application of latent-heat buffer thermal energy storage to a point-focusing solar receiver equipped with an air Brayton engine was studied. To demonstrate the effect of buffer thermal energy storage on engine operation, a computer program was written which models the recuperator, receiver, and thermal storage device as finite-element thermal masses. Actual operating or predicted performance data are used for all components, including the rotating equipment. Based on insolation input and a specified control scheme, the program predicts the Brayton engine operation, including flows, temperatures, and pressures for the various components, along with the engine output power. An economic parametric study indicates that the economic viability of buffer thermal energy storage is largely a function of the achievable engine life.
Thermal dynamic simulation of wall for building energy efficiency under varied climate environment
Wang, Xuejin; Zhang, Yujin; Hong, Jing
2017-08-01
Aiming at different kind of walls in five cities of different zoning for thermal design, using thermal instantaneous response factors method, the author develops software to calculation air conditioning cooling load temperature, thermal response factors, and periodic response factors. On the basis of the data, the author gives the net work analysis about the influence of dynamic thermal of wall on air-conditioning load and thermal environment in building of different zoning for thermal design regional, and put forward the strategy how to design thermal insulation and heat preservation wall base on dynamic thermal characteristic of wall under different zoning for thermal design regional. And then provide the theory basis and the technical references for the further study on the heat preservation with the insulation are in the service of energy saving wall design. All-year thermal dynamic load simulating and energy consumption analysis for new energy-saving building is very important in building environment. This software will provide the referable scientific foundation for all-year new thermal dynamic load simulation, energy consumption analysis, building environment systems control, carrying through farther research on thermal particularity and general particularity evaluation for new energy -saving walls building. Based on which, we will not only expediently design system of building energy, but also analyze building energy consumption and carry through scientific energy management. The study will provide the referable scientific foundation for carrying through farther research on thermal particularity and general particularity evaluation for new energy saving walls building.
Validation and applicability of the 3D core kinetics and thermal hydraulics coupled code SPARKLE
International Nuclear Information System (INIS)
Miyata, Manabu; Maruyama, Manabu; Ogawa, Junto; Otake, Yukihiko; Miyake, Shuhei; Tabuse, Shigehiko; Tanaka, Hirohisa
2009-01-01
The SPARKLE code is a coupled code system based on three individual codes whose physical models have already been verified and validated. Mitsubishi Heavy Industries (MHI) confirmed the coupling calculation, including data transfer and the total reactor coolant system (RCS) behavior of the SPARKLE code. The confirmation uses the OECD/NEA MSLB benchmark problem, which is based on Three Mile Island Unit 1 (TMI-1) nuclear power plant data. This benchmark problem has been used to verify coupled codes developed and used by many organizations. Objectives of the benchmark program are as follows. Phase 1 is to compare the results of the system transient code using point kinetics. Phase 2 is to compare the results of the coupled three-dimensional (3D) core kinetics code and 3D core thermal-hydraulics (T/H) code, and Phase 3 is to compare the results of the combined coupled system transient code, 3D core kinetics code, and 3D core T/H code as a total validation of the coupled calculation. The calculation results of the SPARKLE code indicate good agreement with other benchmark participants' results. Therefore, the SPARKLE code is validated through these benchmark problems. In anticipation of applying the SPARKLE code to licensing analyses, MHI and Japanese PWR utilities have established a safety analysis method regarding the calculation conditions such as power distributions, reactivity coefficients, and event-specific features. (author)
Development of input data to energy code for analysis of reactor fuel bundles
International Nuclear Information System (INIS)
Carre, F.O.; Todreas, N.E.
1975-05-01
The ENERGY 1 code is a semi-empirical method for predicting temperature distributions in wire wrapped rod bundles of a LMFBR. A comparison of ENERGY 1 and MISTRAL 2 is presented. The predictions of ENERGY 1 for special sets of data taken under geometric conditions at the limits of the code are analyzed. 14 references
Aquifer thermal energy (heat and chill) storage
Energy Technology Data Exchange (ETDEWEB)
Jenne, E.A. (ed.)
1992-11-01
As part of the 1992 Intersociety Conversion Engineering Conference, held in San Diego, California, August 3--7, 1992, the Seasonal Thermal Energy Storage Program coordinated five sessions dealing specifically with aquifer thermal energy storage technologies (ATES). Researchers from Sweden, The Netherlands, Germany, Switzerland, Denmark, Canada, and the United States presented papers on a variety of ATES related topics. With special permission from the Society of Automotive Engineers, host society for the 1992 IECEC, these papers are being republished here as a standalone summary of ATES technology status. Individual papers are indexed separately.
Lower Bounds on the Maximum Energy Benefit of Network Coding for Wireless Multiple Unicast
Goseling, J.; Matsumoto, R.; Uyematsu, T.; Weber, J.H.
2010-01-01
We consider the energy savings that can be obtained by employing network coding instead of plain routing in wireless multiple unicast problems. We establish lower bounds on the benefit of network coding, defined as the maximum of the ratio of the minimum energy required by routing and network coding
Lower bounds on the maximum energy benefit of network coding for wireless multiple unicast
Goseling, Jasper; Matsumoto, Ryutaroh; Uyematsu, Tomohiko; Weber, Jos H.
2010-01-01
We consider the energy savings that can be obtained by employing network coding instead of plain routing in wireless multiple unicast problems. We establish lower bounds on the benefit of network coding, defined as the maximum of the ratio of the minimum energy required by routing and network coding
Lower Bounds on the Maximum Energy Benefit of Network Coding for Wireless Multiple Unicast
Directory of Open Access Journals (Sweden)
Matsumoto Ryutaroh
2010-01-01
Full Text Available We consider the energy savings that can be obtained by employing network coding instead of plain routing in wireless multiple unicast problems. We establish lower bounds on the benefit of network coding, defined as the maximum of the ratio of the minimum energy required by routing and network coding solutions, where the maximum is over all configurations. It is shown that if coding and routing solutions are using the same transmission range, the benefit in d-dimensional networks is at least . Moreover, it is shown that if the transmission range can be optimized for routing and coding individually, the benefit in 2-dimensional networks is at least 3. Our results imply that codes following a decode-and-recombine strategy are not always optimal regarding energy efficiency.
Statistical physics inspired energy-efficient coded-modulation for optical communications.
Djordjevic, Ivan B; Xu, Lei; Wang, Ting
2012-04-15
Because Shannon's entropy can be obtained by Stirling's approximation of thermodynamics entropy, the statistical physics energy minimization methods are directly applicable to the signal constellation design. We demonstrate that statistical physics inspired energy-efficient (EE) signal constellation designs, in combination with large-girth low-density parity-check (LDPC) codes, significantly outperform conventional LDPC-coded polarization-division multiplexed quadrature amplitude modulation schemes. We also describe an EE signal constellation design algorithm. Finally, we propose the discrete-time implementation of D-dimensional transceiver and corresponding EE polarization-division multiplexed system. © 2012 Optical Society of America
Validation studies of thermal-hydraulic code for safety analysis of nuclear power plants
International Nuclear Information System (INIS)
Haapalehto, T.
1995-01-01
The thesis gives an overview of the validation process for thermal-hydraulic system codes and it presents in more detail the assessment and validation of the French code CATHARE for VVER calculations. Three assessment cases are presented: loop seal clearing, core reflooding and flow in a horizontal steam generator. The experience gained during these assessment and validation calculations has been used to analyze the behavior of the horizontal steam generator and the natural circulation in the geometry of the Loviisa nuclear power plant. Large part of the work has been performed in cooperation with the CATHARE-team in Grenoble, France. (41 refs., 11 figs., 8 tabs.)
Microwavable thermal energy storage material
Salyer, I.O.
1998-09-08
A microwavable thermal energy storage material is provided which includes a mixture of a phase change material and silica, and a carbon black additive in the form of a conformable dry powder of phase change material/silica/carbon black, or solid pellets, films, fibers, moldings or strands of phase change material/high density polyethylene/ethylene vinyl acetate/silica/carbon black which allows the phase change material to be rapidly heated in a microwave oven. The carbon black additive, which is preferably an electrically conductive carbon black, may be added in low concentrations of from 0.5 to 15% by weight, and may be used to tailor the heating times of the phase change material as desired. The microwavable thermal energy storage material can be used in food serving applications such as tableware items or pizza warmers, and in medical wraps and garments. 3 figs.
Dynamic tuning of optical absorbers for accelerated solar-thermal energy storage.
Wang, Zhongyong; Tong, Zhen; Ye, Qinxian; Hu, Hang; Nie, Xiao; Yan, Chen; Shang, Wen; Song, Chengyi; Wu, Jianbo; Wang, Jun; Bao, Hua; Tao, Peng; Deng, Tao
2017-11-14
Currently, solar-thermal energy storage within phase-change materials relies on adding high thermal-conductivity fillers to improve the thermal-diffusion-based charging rate, which often leads to limited enhancement of charging speed and sacrificed energy storage capacity. Here we report the exploration of a magnetically enhanced photon-transport-based charging approach, which enables the dynamic tuning of the distribution of optical absorbers dispersed within phase-change materials, to simultaneously achieve fast charging rates, large phase-change enthalpy, and high solar-thermal energy conversion efficiency. Compared with conventional thermal charging, the optical charging strategy improves the charging rate by more than 270% and triples the amount of overall stored thermal energy. This superior performance results from the distinct step-by-step photon-transport charging mechanism and the increased latent heat storage through magnetic manipulation of the dynamic distribution of optical absorbers.
Zhang, Fang
2015-02-13
© 2015 WILEY-VCH Verlag GmbH & Co. KGaA. A thermally regenerative ammonia battery (TRAB) is a new approach for converting low-grade thermal energy into electricity by using an ammonia electrolyte and copper electrodes. TRAB operation at 72°C produced a power density of 236±8 Wm-2, with a linear decrease in power to 95±5 Wm-2 at 23°C. The improved power at higher temperatures was due to reduced electrode overpotentials and more favorable thermodynamics for the anode reaction (copper oxidation). The energy density varied with temperature and discharge rates, with a maximum of 650 Whm-3 at a discharge energy efficiency of 54% and a temperature of 37°C. The energy efficiency calculated with chemical process simulation software indicated a Carnot-based efficiency of up to 13% and an overall thermal energy recovery of 0.5%. It should be possible to substantially improve these energy recoveries through optimization of electrolyte concentrations and by using improved ion-selective membranes and energy recovery systems such as heat exchangers.
SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis
Energy Technology Data Exchange (ETDEWEB)
Basehore, K.L.; Todreas, N.E.
1980-08-01
Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries.
SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis
International Nuclear Information System (INIS)
Basehore, K.L.; Todreas, N.E.
1980-08-01
Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries
Effect of Set-point Variation on Thermal Comfort and Energy Use in a Plus-energy Dwelling
DEFF Research Database (Denmark)
Toftum, Jørn; Kazanci, Ongun Berk; Olesen, Bjarne W.
2016-01-01
When designing buildings and space conditioning systems, the occupant thermal comfort, health, and productivity are the main criteria to satisfy. However, this should be achieved with the most energy-efficient space conditioning systems (heating, cooling, and ventilation). Control strategy, set......-points, and control dead-bands have a direct effect on the thermal environment in and the energy use of a building. The thermal environment in and the energy use of a building are associated with the thermal mass of the building and the control strategy, including set-points and control dead-bands. With thermally...... active building systems (TABS), temperatures are allowed to drift within the comfort zone, while in spaces with air-conditioning, temperatures in a narrower interval typically are aimed at. This behavior of radiant systems provides certain advantages regarding energy use, since the temperatures...
Hierarchical Graphene Foam for Efficient Omnidirectional Solar-Thermal Energy Conversion.
Ren, Huaying; Tang, Miao; Guan, Baolu; Wang, Kexin; Yang, Jiawei; Wang, Feifan; Wang, Mingzhan; Shan, Jingyuan; Chen, Zhaolong; Wei, Di; Peng, Hailin; Liu, Zhongfan
2017-10-01
Efficient solar-thermal energy conversion is essential for the harvesting and transformation of abundant solar energy, leading to the exploration and design of efficient solar-thermal materials. Carbon-based materials, especially graphene, have the advantages of broadband absorption and excellent photothermal properties, and hold promise for solar-thermal energy conversion. However, to date, graphene-based solar-thermal materials with superior omnidirectional light harvesting performances remain elusive. Herein, hierarchical graphene foam (h-G foam) with continuous porosity grown via plasma-enhanced chemical vapor deposition is reported, showing dramatic enhancement of broadband and omnidirectional absorption of sunlight, which thereby can enable a considerable elevation of temperature. Used as a heating material, the external solar-thermal energy conversion efficiency of the h-G foam impressively reaches up to ≈93.4%, and the solar-vapor conversion efficiency exceeds 90% for seawater desalination with high endurance. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.
Validation of coupled neutronic / thermal-hydraulic codes for VVER reactors. Final report
Energy Technology Data Exchange (ETDEWEB)
Mittag, S.; Grundmann, U.; Kliem, S.; Kozmenkov, Y.; Rindelhardt, U.; Rohde, U.; Weiss, F.-P.; Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K.-D.; Vanttola, T.; Haemaelaeinen, A.; Kaloinen, E.; Kereszturi, A.; Hegyi, G.; Panka, I.; Hadek, J.; Strmensky, C.; Darilek, P.; Petkov, P.; Stefanova, S.; Kuchin, A.; Khalimonchuk, V.; Hlbocky, P.; Sico, D.; Danilin, S.; Ionov, V.; Nikonov, S.; Powney, D.
2004-08-01
In recent years, the simulation methods for the safety analysis of nuclear power plants have been continuously improved to perform realistic calculations. Therefore in VALCO work package 2 (WP 2), the usual application of coupled neutron-kinetic / thermal-hydraulic codes to VVER has been supplemented by systematic uncertainty and sensitivity analyses. A comprehensive uncertainty analysis has been carried out. The GRS uncertainty and sensitivity method based on the statistical code package SUSA was applied to the two transients studied earlier in SRR-1/95: A load drop of one turbo-generator in Loviisa-1 (VVER-440), and a switch-off of one feed water pump in Balakovo-4 (VVER-1000). The main steps of these analyses and the results obtained by applying different coupled code systems (SMABRE - HEXTRAN, ATHLET - DYN3D, ATHLET - KIKO3D, ATHLET - BIPR-8) are described in this report. The application of this method is only based on variations of input parameter values. No internal code adjustments are needed. An essential result of the analysis using the GRS SUSA methodology is the identification of the input parameters, such as the secondary-circuit pressure, the control-assembly position (as a function of time), and the control-assembly efficiency, that most sensitively affect safety-relevant output parameters, like reactor power, coolant heat-up, and primary pressure. Uncertainty bands for these output parameters have been derived. The variation of potentially uncertain input parameter values as a consequence of uncertain knowledge can activate system actions causing quite different transient evolutions. This gives indications about possible plant conditions that might be reached from the initiating event assuming only small disturbances. In this way, the uncertainty and sensitivity analysis reveals the spectrum of possible transient evolutions. Deviations of SRR-1/95 coupled code calculations from measurements also led to the objective to separate neutron kinetics from
Survey of particle codes in the Magnetic Fusion Energy Program
International Nuclear Information System (INIS)
1977-12-01
In the spring of 1976, the Fusion Plasma Theory Branch of the Division of Magnetic Fusion Energy conducted a survey of all the physics computer codes being supported at that time. The purpose of that survey was to allow DMFE to prepare a description of the codes for distribution to the plasma physics community. This document is the first of several planned and covers those types of codes which treat the plasma as a group of particles
Energy-Efficient Cluster Based Routing Protocol in Mobile Ad Hoc Networks Using Network Coding
Srinivas Kanakala; Venugopal Reddy Ananthula; Prashanthi Vempaty
2014-01-01
In mobile ad hoc networks, all nodes are energy constrained. In such situations, it is important to reduce energy consumption. In this paper, we consider the issues of energy efficient communication in MANETs using network coding. Network coding is an effective method to improve the performance of wireless networks. COPE protocol implements network coding concept to reduce number of transmissions by mixing the packets at intermediate nodes. We incorporate COPE into cluster based routing proto...
International Nuclear Information System (INIS)
Terra, Andre Miguel Barge Pontes Torres
2005-01-01
The Albedo method applied to criticality calculations to nuclear reactors is characterized by following the neutron currents, allowing to make detailed analyses of the physics phenomena about interactions of the neutrons with the core-reflector set, by the determination of the probabilities of reflection, absorption, and transmission. Then, allowing to make detailed appreciations of the variation of the effective neutron multiplication factor, keff. In the present work, motivated for excellent results presented in dissertations applied to thermal reactors and shieldings, was described the methodology to Albedo method for the analysis criticality of thermal reactors by using two energy groups admitting variable core coefficients to each re-entrant current. By using the Monte Carlo KENO IV code was analyzed relation between the total fraction of neutrons absorbed in the core reactor and the fraction of neutrons that never have stayed into the reflector but were absorbed into the core. As parameters of comparison and analysis of the results obtained by the Albedo method were used one dimensional deterministic code ANISN (ANIsotropic SN transport code) and Diffusion method. The keff results determined by the Albedo method, to the type of analyzed reactor, showed excellent agreement. Thus were obtained relative errors of keff values smaller than 0,78% between the Albedo method and code ANISN. In relation to the Diffusion method were obtained errors smaller than 0,35%, showing the effectiveness of the Albedo method applied to criticality analysis. The easiness of application, simplicity and clarity of the Albedo method constitute a valuable instrument to neutronic calculations applied to nonmultiplying and multiplying media. (author)
Energy Technology Data Exchange (ETDEWEB)
Fajeau, M; Nguyen, L T; Saunier, J [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)
1966-09-01
This code handles the following problems: -1) Analysis of thermal experiments on a water loop at high or low pressure; steady state or transient behavior; -2) Analysis of thermal and hydrodynamic behavior of water-cooled and moderated reactors, at either high or low pressure, with boiling permitted; fuel elements are assumed to be flat plates: - Flowrate in parallel channels coupled or not by conduction across plates, with conditions of pressure drops or flowrate, variable or not with respect to time is given; the power can be coupled to reactor kinetics calculation or supplied by the code user. The code, containing a schematic representation of safety rod behavior, is a one dimensional, multi-channel code, and has as its complement (FLID), a one-channel, two-dimensional code. (authors) [French] Ce code permet de traiter les problemes ci-dessous: 1. Depouillement d'essais thermiques sur boucle a eau, haute ou basse pression, en regime permanent ou transitoire; 2. Etudes thermiques et hydrauliques de reacteurs a eau, a plaques, a haute ou basse pression, ebullition permise: - repartition entre canaux paralleles, couples on non par conduction a travers plaques, pour des conditions de debit ou de pertes de charge imposees, variables ou non dans le temps; - la puissance peut etre couplee a la neutronique et une representation schematique des actions de securite est prevue. Ce code (Cactus) a une dimension d'espace et plusieurs canaux, a pour complement Flid qui traite l'etude d'un seul canal a deux dimensions. (auteurs)
Bridger, D. W.; Allen, D. M.
2014-01-01
A modeling study was carried out to evaluate the influence of aquifer heterogeneity, as represented by geologic layering, on heat transport and storage in an aquifer thermal energy storage (ATES) system in Agassiz, British Columbia, Canada. Two 3D heat transport models were developed and calibrated using the flow and heat transport code FEFLOW including: a "non-layered" model domain with homogeneous hydraulic and thermal properties; and, a "layered" model domain with variable hydraulic and thermal properties assigned to discrete geological units to represent aquifer heterogeneity. The base model (non-layered) shows limited sensitivity for the ranges of all thermal and hydraulic properties expected at the site; the model is most sensitive to vertical anisotropy and hydraulic gradient. Simulated and observed temperatures within the wells reflect a combination of screen placement and layering, with inconsistencies largely explained by the lateral continuity of high permeability layers represented in the model. Simulation of heat injection, storage and recovery show preferential transport along high permeability layers, resulting in longitudinal plume distortion, and overall higher short-term storage efficiencies.
Semi-flexible bimetal-based thermal energy harvesters
International Nuclear Information System (INIS)
Boisseau, S; Despesse, G; Monfray, S; Puscasu, O; Skotnicki, T
2013-01-01
This paper introduces a new semi-flexible device able to turn thermal gradients into electricity by using a curved bimetal coupled to an electret-based converter. In fact, a two-step conversion is carried out: (i) a curved bimetal turns the thermal gradient into a mechanical oscillation that is then (ii) converted into electricity thanks to an electrostatic converter using electrets in Teflon ® . The semi-flexible and low-cost design of these new energy converters pave the way to mass production over large areas of thermal energy harvesters. Raw output powers up to 13.46 μW per device were reached on a hot source at 60 °C with forced convection. Then, a DC-to-DC flyback converter has been sized to turn the energy harvesters’ raw output powers into a viable supply source for an electronic circuit (DC-3 V). At the end, 10 μW of directly usable output power were reached with 3 devices, which is compatible with wireless sensor network powering applications. (paper)
Version 4. 00 of the MINTEQ geochemical code
Energy Technology Data Exchange (ETDEWEB)
Eary, L.E.; Jenne, E.A.
1992-09-01
The MINTEQ code is a thermodynamic model that can be used to calculate solution equilibria for geochemical applications. Included in the MINTEQ code are formulations for ionic speciation, ion exchange, adsorption, solubility, redox, gas-phase equilibria, and the dissolution of finite amounts of specified solids. Since the initial development of the MINTEQ geochemical code, a number of undocumented versions of the source code and data files have come into use at the Pacific Northwest Laboratory (PNL). This report documents these changes, describes source code modifications made for the Aquifer Thermal Energy Storage (ATES) program, and provides comprehensive listings of the data files. A version number of 4.00 has been assigned to the MINTEQ source code and the individual data files described in this report.
Version 4.00 of the MINTEQ geochemical code
Energy Technology Data Exchange (ETDEWEB)
Eary, L.E.; Jenne, E.A.
1992-09-01
The MINTEQ code is a thermodynamic model that can be used to calculate solution equilibria for geochemical applications. Included in the MINTEQ code are formulations for ionic speciation, ion exchange, adsorption, solubility, redox, gas-phase equilibria, and the dissolution of finite amounts of specified solids. Since the initial development of the MINTEQ geochemical code, a number of undocumented versions of the source code and data files have come into use at the Pacific Northwest Laboratory (PNL). This report documents these changes, describes source code modifications made for the Aquifer Thermal Energy Storage (ATES) program, and provides comprehensive listings of the data files. A version number of 4.00 has been assigned to the MINTEQ source code and the individual data files described in this report.
High Density Thermal Energy Storage with Supercritical Fluids
Ganapathi, Gani B.; Wirz, Richard
2012-01-01
A novel approach to storing thermal energy with supercritical fluids is being investigated, which if successful, promises to transform the way thermal energy is captured and utilized. The use of supercritical fluids allows cost-affordable high-density storage with a combination of latent heat and sensible heat in the two-phase as well as the supercritical state. This technology will enhance penetration of several thermal power generation applications and high temperature water for commercial use if the overall cost of the technology can be demonstrated to be lower than the current state-of-the-art molten salt using sodium nitrate and potassium nitrate eutectic mixtures.
Demonstration of EnergyNest thermal energy storage (TES) technology
Hoivik, Nils; Greiner, Christopher; Tirado, Eva Bellido; Barragan, Juan; Bergan, Pâl; Skeie, Geir; Blanco, Pablo; Calvet, Nicolas
2017-06-01
This paper presents the experimental results from the EnergyNest 2 × 500 kWhth thermal energy storage (TES) pilot system installed at Masdar Institute of Science & Technology Solar Platform. Measured data are shown and compared to simulations using a specially developed computer program to verify the stability and performance of the TES. The TES is based on a solid-state concrete storage medium (HEATCRETE®) with integrated steel tube heat exchangers cast into the concrete. The unique concrete recipe used in the TES has been developed in collaboration with Heidelberg Cement; this material has significantly higher thermal conductivity compared to regular concrete implying very effective heat transfer, at the same time being chemically stable up to 450 °C. The demonstrated and measured performance of the TES matches the predictions based on simulations, and proves the operational feasibility of the EnergyNest concrete-based TES. A further case study is analyzed where a large-scale TES system presented in this article is compared to two-tank indirect molten salt technology.
Thermalization of positronium in helium: A numerical study
Energy Technology Data Exchange (ETDEWEB)
Marjanovic, S.; Suvakov, M. [Institute of Physics, University of Belgrade, Pregrevica 118, 11080 Belgrade (Serbia); Engbrecht, J.J. [Saint Olaf College, Northfield, MN 55057 (United States); Petrovic, Z.Lj., E-mail: zoran@ipb.ac.rs [Institute of Physics, University of Belgrade, Pregrevica 118, 11080 Belgrade (Serbia)
2012-05-15
In this paper we present a numerical study of positronium (Ps) thermalization in pure helium (He). Recent measurements of Ps thermalization yielded data that were analyzed to produce the scattering cross-sections in helium by using energy balance equations with an assumption of a Maxwell-Boltzmann distribution (MBD) function for Ps. We have applied a Monte Carlo code to test the cross-sections. As our code was developed without any approximations for the energy distribution function we have effectively also tested the assumptions and the validity of the simple theory based on Maxwellian distributions. We present the simulation results using the simulation technique that is limited only by the accuracy of the available cross-sections. We calculate thermalization profiles for several theoretical and measured cross-sections. Also, the temporal evolution of energy distributions has been shown along with diffusion coefficients and spatial ranges of penetration. Thermalization of the initial distribution is rapid and the data follow relatively closely, those calculated in recent experiment, which supports the choice of MBD and the obtained cross-section. However the distribution function most of the time deviates from the MBD due to strong scattering. Finally, we applied the same procedure to analyze Ps thermalization in water vapor.
Enthalpy estimation for thermal comfort and energy saving in air conditioning system
International Nuclear Information System (INIS)
Chu, C.-M.; Jong, T.-L.
2008-01-01
The thermal comfort control of a room must consider not only the thermal comfort level but also energy saving. This paper proposes an enthalpy estimation that is conducive for thermal comfort control and energy saving. The least enthalpy estimator (LEE) combines the concept of human thermal comfort with the theory of enthalpy to predict the load for a suitable setting pair in order to maintain more precisely the thermal comfort level and save energy in the air conditioning system
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)
Composite Materials for Thermal Energy Storage: Enhancing Performance through Microstructures
Ge, Zhiwei; Ye, Feng; Ding, Yulong
2014-01-01
Chemical incompatibility and low thermal conductivity issues of molten-salt-based thermal energy storage materials can be addressed by using microstructured composites. Using a eutectic mixture of lithium and sodium carbonates as molten salt, magnesium oxide as supporting material, and graphite as thermal conductivity enhancer, the microstructural development, chemical compatibility, thermal stability, thermal conductivity, and thermal energy storage performance of composite materials are investigated. The ceramic supporting material is essential for preventing salt leakage and hence provides a solution to the chemical incompatibility issue. The use of graphite gives a significant enhancement on the thermal conductivity of the composite. Analyses suggest that the experimentally observed microstructural development of the composite is associated with the wettability of the salt on the ceramic substrate and that on the thermal conduction enhancer. PMID:24591286
Energy Technology Data Exchange (ETDEWEB)
Wortman, D.; Echo-Hawk, L.
2005-02-01
This report is an attempt to describe the building code requirements and impediments to the application of EE and RE technologies in residential buildings. Several modern model building codes were reviewed. These are representative of the codes that will be adopted by most locations in the coming years. The codes reviewed for this report include: International Residential Code, First Draft, April 1998; International Energy Conservation Code, 1998; International Mechanical Code, 1998; International Plumbing Code, 1997; International Fuel Gas Code, 1997; National Electrical Code, 1996. These codes were reviewed as to their application to (1) PV systems in buildings and building-integrated PV systems and (2) active solar domestic hot water and space-heating systems. A discussion of general code issues that impact these technologies is also included. Examples of this are solar access and sustainability.
International Nuclear Information System (INIS)
Li, Tingxian; Wang, Ruzhu; Kiplagat, Jeremiah K.; Kang, YongTae
2013-01-01
An innovative dual-mode thermochemical sorption energy storage method is proposed for seasonal storage of solar thermal energy with little heat losses. During the charging phase in summer, solar thermal energy is stored in form of chemical bonds resulting from thermochemical decomposition process, which enables the stored energy to be kept several months at ambient temperature. During the discharging phase in winter, the stored thermal energy is released in the form of chemical reaction heat resulting from thermochemical synthesis process. Thermodynamic analysis showed that the advanced dual-mode thermochemical sorption energy storage is an effective method for the long-term seasonal storage of solar energy. A coefficient of performance (COP h ) of 0.6 and energy density higher than 1000 kJ/kg of salt can be attained from the proposed system. During the discharging phase at low ambient temperatures, the stored thermal energy can be upgraded by use of a solid–gas thermochemical sorption heat transformer cycle. The proposed thermochemical sorption energy storage has distinct advantages over the conventional sensible heat and latent heat storage, such as higher energy storage density, little heat losses, integrated energy storage and energy upgrade, and thus it can contribute to improve the seasonal utilization of solar thermal energy. - Highlights: ► A dual-mode solid thermochemical sorption is proposed for seasonal solar thermal energy storage. ► Energy upgrade techniques into the energy storage system are integrated. ► Performance of the proposed seasonal energy storage system is evaluated. ► Energy density and COP h from the proposed system are as high as 1043 kJ/kg of salt and 0.60, respectively
International Nuclear Information System (INIS)
Marais, D.; Greyvenstein, G. P.
2008-01-01
TINTE is a well established reactor analysis code which models the transient behaviour of pebble bed reactor cores but it does not include the capabilities to model a power conversion unit (PCU). This raises the issue that TINTE cannot model full system transients. One way to overcome this problem is to supply TINTE with time-dependant thermal-hydraulic boundary conditions which are obtained from PCU simulations. This study investigates a method to provide boundary conditions for the nuclear code TINTE during full system transients. This was accomplished by creating a high level interface between the systems CFD code Flownex and TINTE. An indirect coupling method is explored whereby characteristics of the PCU are matched to characteristics of the nuclear core. This method eliminates the need to iterate between the two codes. A number of transients are simulated using the coupled code and then compared against stand-alone Flownex simulations. The coupling method introduces relatively small errors when reproducing mass flow, temperature and pressure in steady state analysis, but become more pronounced when dealing with fast thermal-hydraulic transients. Decreasing the maximum time step length of TINTE reduces this problem, but increases the computational time. Copyright ASME 2008. (authors)
FEMAXI-III: a computer code for the analysis of thermal and mechanical behavior of fuel rods
International Nuclear Information System (INIS)
Nakajima, Tetsuo; Ichikawa, Michio; Iwano, Yoshihiko; Ito, Kenichi; Saito, Hiroaki; Kashima, Koichi; Kinoshita, Motoyasu; Okubo, Tadatsune.
1985-12-01
FEMAXI-III is a computer code to predict the thermal and mechanical behavior of a light water fuel rod during its irradiation life. It can analyze the integral behavior of a whole fuel rod throughout its life, as well as the localized behavior of a small part of fuel rod. The localized mechanical behavior such as the cladding ridge deformation is analyzed by the two-dimensional axisymmetric finite element method. FEMAXI-III calculates, in particular, the temperature distribution, the radial deformation, the fission gas release, and the inner gas pressure as a function of irradiation time and axial position, and the stresses and strains in the fuel and cladding at a small part of fuel rod as a function of irradiation time. For this purpose, Elasto-plasticity, creep, thermal expansion, fuel cracking and crack healing, relocation, densification, swelling, hot pressing, heat generation distribution, fission gas release, and fuel-cladding mechanical interaction are modelled and their interconnected effects are considered in the code. Efforts have been made to improve the accuracy and stability of finite element solution and to minimize the computer memory and running time. This report describes the outline of the code and the basic models involved, and also includes the application of the code and its input manual. (author)
Energy Technology Data Exchange (ETDEWEB)
Hartmann, Christoph Oliver
2016-06-13
Coupled Thermal-hydraulic/Neutron-kinetic (TH/NK) simulations of Boiling Water Reactor transients require well validated and accurate simulation tools. The generation of cross-section (XS) libraries, depending on the individual thermal-hydraulic state parameters, is of paramount importance for coupled simulations. Problem-dependent XS-sets for 3D core simulations are being generated mainly by well validated, fast running commercial and user-friendly lattice codes such as CASMO and HELIOS. In this dissertation a computational route, based on the lattice code SCALE6/TRITON, the cross-section interface GenPMAXS, the best-estimate thermal-hydraulic system code TRACE and the core simulator PARCS, for best-estimate simulations of Boiling Water (BWR) transients has been developed and validated. The computational route has been supplemented by a subsequent uncertainty and sensitivity study based on Monte Carlo sampling and propagation of the uncertainties of input parameters to the output (SUSA code). The analysis of a single BWR fuel assembly depletion problem with PARCS using SCALE/TRITON cross-sections has been shown a good agreement with the results obtained with CASMO cross-section sets. However, to compensate the deficiencies of the interface program GenPMAXS, PYTHON scripts had to be developed to incorporate missing data, as the yields of Iodine, Xenon and Promethium, into the cross-section-data sets (PMAXS-format) generated by GenPMAXS from the SCALE/TRITON output. The results of the depletion analysis of a full BWR core with PARCS have indicated the importance of considering history effects, adequate modeling of the reflector region and the control rods, as the PARCS simulations for depleted fuel and all control rods inserted (ARI) differs significantly at the fuel assembly top and bottom. Systematic investigations with the coupled codes TRACE/PARCS have been performed to analyse the core behaviour at different thermal conditions using nuclear data (XS
International Nuclear Information System (INIS)
Costin, L.S.; Bauer, S.J.
1991-10-01
Thermal and mechanical models for intact and jointed rock mass behavior are being developed, verified, and validated at Sandia National Laboratories for the Yucca Mountain Site Characterization Project. Benchmarking is an essential part of this effort and is one of the tools used to demonstrate verification of engineering software used to solve thermomechanical problems. This report presents the results of the third (and final) phase of the first thermomechanical benchmark exercise. In the first phase of this exercise, nonlinear heat conduction code were used to solve the thermal portion of the benchmark problem. The results from the thermal analysis were then used as input to the second and third phases of the exercise, which consisted of solving the structural portion of the benchmark problem. In the second phase of the exercise, a linear elastic rock mass model was used. In the third phase of the exercise, two different nonlinear jointed rock mass models were used to solve the thermostructural problem. Both models, the Sandia compliant joint model and the RE/SPEC joint empirical model, explicitly incorporate the effect of the joints on the response of the continuum. Three different structural codes, JAC, SANCHO, and SPECTROM-31, were used with the above models in the third phase of the study. Each model was implemented in two different codes so that direct comparisons of results from each model could be made. The results submitted by the participants showed that the finite element solutions using each model were in reasonable agreement. Some consistent differences between the solutions using the two different models were noted but are not considered important to verification of the codes. 9 refs., 18 figs., 8 tabs
Aquifer thermal energy stores in Germany
International Nuclear Information System (INIS)
Kabus, F.; Seibt, P.; Poppei, J.
2000-01-01
This paper describes the state of essential demonstration projects of heat and cold storage in aquifers in Germany. Into the energy supply system of the buildings of the German Parliament in Berlin, there are integrated both a deep brine-bearing aquifer for the seasonal storage of waste heat from power and heat cogeneration and a shallow-freshwater bearing aquifer for cold storage. In Neubrandenburg, a geothermal heating plant which uses a 1.200 m deep aquifer is being retrofitted into an aquifer heat storage system which can be charged with the waste heat from a gas and steam cogeneration plant. The first centralised solar heating plant including an aquifer thermal energy store in Germany was constructed in Rostock. Solar collectors with a total area of 1000m 2 serve for the heating of a complex of buildings with 108 flats. A shallow freshwater-bearing aquifer is used for thermal energy storage. (Authors)
Optimal initial fuel distribution in a thermal reactor for maximum energy production
International Nuclear Information System (INIS)
Moran-Lopez, J.M.
1983-01-01
Using the fuel burnup as objective function, it is desired to determine the initial distribution of the fuel in a reactor in order to obtain the maximum energy possible, for which, without changing a fixed initial fuel mass, the results for different initial fuel and control poison configurations are analyzed and the corresponding running times compared. One-dimensional, two energy-group theory is applied to a reflected cylindrical reactor using U-235 as fuel and light water as moderator and reflector. Fissions in both fast and thermal groups are considered. The reactor is divided into several annular regions, and the constant flux approximation in each depletion step is then used to solve the fuel and fission-product poisons differential equations in each region. The computer code OPTIME was developed to determine the time variation of core properties during the fuel cycle. At each depletion step, OPTIME calls ODMUG, [12] a criticality search program, from which the spatially-averaged neutron fluxes and control poison cross sections are obtained
International Nuclear Information System (INIS)
Biaty, Patricia Andrea Paladino; Sabundjian, Gaiane
2005-01-01
The thermal hydraulic study in accidents and transients analyses in nuclear power plants is realized with some special tools. These programs use the best estimate analyses and have been developed to simulate accidents and transients in Pressurized Water Reactors (PWR) and auxiliary systems. The RELAP5 code has been used as tool to licensing the nuclear facilities in our country, which is the objective of this study. The main problem when RELAP5 code is used is a lot of information necessary to simulate thermal hydraulic accidents. Moreover, there is the necessity of a reasonable amount of mathematical operations to calculation of the geometry of the components existents. Therefore, in order to facilitate the manipulation of this information, it is necessary the developing a friendly preprocessor for attainment of the mathematical calculations for RELAP5 code. One of the tools used for some of these calculations is the MS-EXCEL, which will be used in this work. (author)
Application of nanomaterials in solar thermal energy storage
Shamshirgaran, Seyed Reza; Khalaji Assadi, Morteza; Viswanatha Sharma, Korada
2018-06-01
Solar thermal conversion technology harvests the sun's energy, rather than fossil fuels, to generate low-cost, low/zero-emission energy in the form of heating, cooling or electrical form for residential, commercial, and industrial sectors. The advent of nanofluids and nanocomposites or phase change materials, is a new field of study which is adapted to enhance the efficiency of solar collectors. The concepts of thermal energy storage technologies are investigated and the role of nanomaterials in energy conversion is discussed. This review revealed that although the exploitation of nanomaterials will boost the performance of solar collectors almost in all cases, this would be accompanied by certain challenges such as production cost, instability, agglomeration and erosion. Earlier studies have dealt with the enhancement of thermal conductivity and heat capacity; however, less attention has been given to the facing challenges. Moreover, no exact criteria can be found for the selection of appropriate nanomaterials and their properties for a specific application. In most research studies, the nanoparticles' material and properties have not been selected based on estimated values so that all the aspects of desired application could be considered simultaneously. The wide spread use of nanomaterials can lead to cost effective solutions as well. Therefore, it seems there should be a sense of techno-economic optimization in exploiting nanomaterials for solar thermal energy storage applications. The optimization should cover the key parameters, particularly nanoparticle type, size, loading and shape which depends on the sort of application and also dispersion technology.
Application of nanomaterials in solar thermal energy storage
Shamshirgaran, Seyed Reza; Khalaji Assadi, Morteza; Viswanatha Sharma, Korada
2017-12-01
Solar thermal conversion technology harvests the sun's energy, rather than fossil fuels, to generate low-cost, low/zero-emission energy in the form of heating, cooling or electrical form for residential, commercial, and industrial sectors. The advent of nanofluids and nanocomposites or phase change materials, is a new field of study which is adapted to enhance the efficiency of solar collectors. The concepts of thermal energy storage technologies are investigated and the role of nanomaterials in energy conversion is discussed. This review revealed that although the exploitation of nanomaterials will boost the performance of solar collectors almost in all cases, this would be accompanied by certain challenges such as production cost, instability, agglomeration and erosion. Earlier studies have dealt with the enhancement of thermal conductivity and heat capacity; however, less attention has been given to the facing challenges. Moreover, no exact criteria can be found for the selection of appropriate nanomaterials and their properties for a specific application. In most research studies, the nanoparticles' material and properties have not been selected based on estimated values so that all the aspects of desired application could be considered simultaneously. The wide spread use of nanomaterials can lead to cost effective solutions as well. Therefore, it seems there should be a sense of techno-economic optimization in exploiting nanomaterials for solar thermal energy storage applications. The optimization should cover the key parameters, particularly nanoparticle type, size, loading and shape which depends on the sort of application and also dispersion technology.
Clean Energy in City Codes: A Baseline Analysis of Municipal Codification across the United States
Energy Technology Data Exchange (ETDEWEB)
Cook, Jeffrey J. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Aznar, Alexandra [National Renewable Energy Lab. (NREL), Golden, CO (United States); Dane, Alexander [National Renewable Energy Lab. (NREL), Golden, CO (United States); Day, Megan [National Renewable Energy Lab. (NREL), Golden, CO (United States); Mathur, Sivani [National Renewable Energy Lab. (NREL), Golden, CO (United States); Doris, Elizabeth [National Renewable Energy Lab. (NREL), Golden, CO (United States)
2016-12-01
Municipal governments in the United States are well positioned to influence clean energy (energy efficiency and alternative energy) and transportation technology and strategy implementation within their jurisdictions through planning, programs, and codification. Municipal governments are leveraging planning processes and programs to shape their energy futures. There is limited understanding in the literature related to codification, the primary way that municipal governments enact enforceable policies. The authors fill the gap in the literature by documenting the status of municipal codification of clean energy and transportation across the United States. More directly, we leverage online databases of municipal codes to develop national and state-specific representative samples of municipal governments by population size. Our analysis finds that municipal governments with the authority to set residential building energy codes within their jurisdictions frequently do so. In some cases, communities set codes higher than their respective state governments. Examination of codes across the nation indicates that municipal governments are employing their code as a policy mechanism to address clean energy and transportation.
Design and installation manual for thermal energy storage
Energy Technology Data Exchange (ETDEWEB)
Cole, R L; Nield, K J; Rohde, R R; Wolosewicz, R M
1980-01-01
The purpose of this manual is to provide information on the design and installation of thermal energy storage in active solar systems. It is intended for contractors, installers, solar system designers, engineers, architects, and manufacturers who intend to enter the solar energy business. The reader should have general knowledge of how solar heating and cooling systems operate and knowledge of construction methods and building codes. Knowledge of solar analysis methods such as f-Chart, SOLCOST, DOE-1, or TRNSYS would be helpful. The information contained in the manual includes sizing storage, choosing a location for the storage device, and insulation requirements. Both air-based and liquid-based systems are covered with topics on designing rock beds, tank types, pump and fan selection, installation, costs, and operation and maintenance. Topics relevant to latent heat storage include properties of phase-change materials, sizing the storage unit, insulating the storage unit, available systems, and cost. Topics relevant to heating domestic water include safety, single- and dual-tank systems, domestic water heating with air- and liquid-based space heating systems, and stand alone domestics hot water systems. Several appendices present common problems with storage systems and their solutions, heat transfer fluid properties, economic insulation thickness, heat exchanger sizing, and sample specifications for heat exchangers, wooden rock bins, steel tanks, concrete tanks, and fiberglass-reinforced plastic tanks.
International Nuclear Information System (INIS)
Ganguly, Sayantan; Mohan Kumar, M.S.; Date, Abhijit; Akbarzadeh, Aliakbar
2017-01-01
Highlights: • A 3D coupled thermo-hydrogeological numerical model of an ATES system is presented. • Importance of a few parameters involved in the study is determined. • Thermal energy discharge by the ATES system for two seasons is estimated. • A strategy and a safe well spacing are proposed to avoid thermal interference. • The proposed model is applied to simulate a real life ATES field study. - Abstract: A three-dimensional (3D) coupled thermo-hydrogeological numerical model for a confined aquifer thermal energy storage (ATES) system underlain and overlain by rock media has been presented in this paper. The ATES system operates in cyclic mode. The model takes into account heat transport processes of advection, conduction and heat loss to confining rock media. The model also includes regional groundwater flow in the aquifer in the longitudinal and lateral directions, geothermal gradient and anisotropy in the aquifer. Results show that thermal injection into the aquifer results in the generation of a thermal-front which grows in size with time. The thermal interference caused by the premature thermal-breakthrough when the thermal-front reaches the production well results in the fall of system performance and hence should be avoided. This study models the transient temperature distribution in the aquifer for different flow and geological conditions which may be effectively used in designing an efficient ATES project by ensuring safety from thermal-breakthrough while catering to the energy demand. Parameter studies are also performed which reveals that permeability of the confining rocks; well spacing and injection temperature are important parameters which influence transient heat transport in the subsurface porous media. Based on the simulations here a safe well spacing is proposed. The thermal energy produced by the system in two seasons is estimated for four different cases and strategy to avoid the premature thermal-breakthrough in critical cases is
Economic Evaluation of a Solar Charged Thermal Energy Store for Space Heating
Melo, Manuel
2013-01-01
A thermal energy store corrects the misalignment of heating demand in the winter relative to solar thermal energy gathered in the summer. This thesis reviews the viability of a solar charged hot water tank thermal energy store for a school at latitude 56.25N, longitude -120.85W
Transf ormation thermotics and the manipulation of thermal energy
Institute of Scientific and Technical Information of China (English)
Xiangfan Xu; Baowen Li
2017-01-01
Thermal energy has been proposed to have ever greater potential for human beings if the heat carriers, pho-nons can be controlled in micron-scale as easy as its counterpart, electrons in solid. However, it is a challenge to control phonons due to its relatively short wavelength, which is in the order of a few nanometers to a few tens of nanometers. Alternatively, in macroscopical scale, functional thermal materials are used to control thermal energy. The transfor-mation of macroscopical thermal diffusion equation is proposed to obtain the asymmetrical thermal conductivity in real space. This new type of thermal functional materials helps to control heat flow and to realize thermal cloak and thermal camouflage. In this review, we summarize the recent advances in constructing thermal functional materials (also called thermal metamaterials). In SecⅠ, we discussed the history of functional materials and the principles of constructing thermal functional materials , special focus was given to the thermal cloak, followed by the realization of thermal cloak in SecⅡ.Thermal camouflage, based on the realization of thermal cloak, was discussed in SecⅢ, which is proposed to have great potentials in military usage. We stressed both the principle and practical based challenges in thermal cloak and thermal camouflage in SecⅣ, in which outlooks were also given. It is worth noting that thermal transports consist of thermal conduction, thermal convection and thermal radiation. Recent progresses on thermal functional materials are based on the transformation of thermotics, i.e. spacial distortion of thermal conducting path, leaving thermal convection and thermal radiation untouched. We hope, though this review paper, to encourage more researchers in China to engage in this field, and to accelerate the practical usage of thermal cloak and thermal camouflage.
2011-03-10
... DEPARTMENT OF ENERGY 10 CFR Part 430 [Docket No. EERE-2011-BT-BC-0009] Building Energy Codes.... The IgCC is intended to provide a green model building code provisions for new and existing commercial... Code (IgCC) AGENCY: Office of Energy Efficiency and Renewable Energy, Department of Energy. ACTION...
Statistical core design methodology using the VIPRE thermal-hydraulics code
International Nuclear Information System (INIS)
Lloyd, M.W.; Feltus, M.A.
1995-01-01
An improved statistical core design methodology for developing a computational departure from nucleate boiling ratio (DNBR) correlation has been developed and applied in order to analyze the nominal 1.3 DNBR limit on Westinghouse Pressurized Water Reactor (PWR) cores. This analysis, although limited in scope, found that the DNBR limit can be reduced from 1.3 to some lower value and be accurate within an adequate confidence level of 95%, for three particular FSAR operational transients: turbine trip, complete loss of flow, and inadvertent opening of a pressurizer relief valve. The VIPRE-01 thermal-hydraulics code, the SAS/STAT statistical package, and the EPRI/Columbia University DNBR experimental data base were used in this research to develop the Pennsylvania State Statistical Core Design Methodology (PSSCDM). The VIPRE code was used to perform the necessary sensitivity studies and generate the EPRI correlation-calculated DNBR predictions. The SAS package used for these EPRI DNBR correlation predictions from VIPRE as a data set to determine the best fit for the empirical model and to perform the statistical analysis. (author)
Composite materials for thermal energy storage: enhancing performance through microstructures.
Ge, Zhiwei; Ye, Feng; Ding, Yulong
2014-05-01
Chemical incompatibility and low thermal conductivity issues of molten-salt-based thermal energy storage materials can be addressed by using microstructured composites. Using a eutectic mixture of lithium and sodium carbonates as molten salt, magnesium oxide as supporting material, and graphite as thermal conductivity enhancer, the microstructural development, chemical compatibility, thermal stability, thermal conductivity, and thermal energy storage performance of composite materials are investigated. The ceramic supporting material is essential for preventing salt leakage and hence provides a solution to the chemical incompatibility issue. The use of graphite gives a significant enhancement on the thermal conductivity of the composite. Analyses suggest that the experimentally observed microstructural development of the composite is associated with the wettability of the salt on the ceramic substrate and that on the thermal conduction enhancer. © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.
Establishment of a JSME code for the evaluation of high-cycle thermal fatigue in mixing tees
International Nuclear Information System (INIS)
Moriya, Shoichi; Fukuda, Toshihiko; Matsunaga, Tomoya; Hirayama, Hiroshi; Shiina, Kouji; Tanimoto, Koichi
2004-01-01
This paper describes a JSME code for high-cycle thermal fatigue evaluation by thermal striping in mixing tees with hot and cold water flows. The evaluation of thermal striping in a mixing tee has four steps to screen design parameters one-by-one according to the severity of the thermal load assessed from design conditions using several evaluation charts. In order to make these charts, visualization tests with acrylic pipes and temperature measurement tests with metal pipes were conducted. The influence of the configurations of mixing tees, flow velocity ratio, pipe diameter ratio and so on was examined from the results of the experiments. This paper makes a short mention of the process of providing these charts. (author)
International Nuclear Information System (INIS)
Ito, Masahiro; Imai, Yasutomo; Uwaba, Tomoyuki; Ohshima, Hiroyuki
2004-03-01
The bundle-duct interaction may occur in sodium cooled wire-wrapped FBR fuel subassemblies in high burn-up conditions. JNC has been developing a bundle deformation analysis code BAMBOO (Behavior Analysis code for Mechanical interaction of fuel Bundle under On-power Operation), a thermal hydraulics analysis code ASFRE-IV (Analysis of Sodium Flow in Reactor Elements - ver. IV) and their coupling method as a simulation system for the evaluation on the integrity of deformed FBR fuel pin bundles. In this study, the simulation system was applied to a coupling analysis of deformation and thermal-hydraulics in the fuel pin-bundle under a steady-state condition just after startup for the purpose of the verification of the simulation system. The iterative calculations of deformation and thermal-hydraulics employed in the coupling analysis provided numerically unstable solutions. From the result, it was found that improvement of the coupling algorithm of BAMBOO and ASFRE-IV is necessary to reduce numerical fluctuations and to obtain better convergence by introducing such computational technique as the optimized under-relaxation method. (author)
Three-dimensional thermal hydraulic best estimate code BAGIRA: new results of verification
International Nuclear Information System (INIS)
Peter Kohut; Sergey D Kalinichenko; Alexander E Kroshilin; Vladimir E Kroshilin; Alexander V Smirnov
2005-01-01
Full text of publication follows: BAGIRA is a three-dimensional inhomogeneous two-velocity two-temperature thermal hydraulic code of best estimate, elaborated in VNIIAES for modeling two-phase flows in the primary circuit and steam generators of VVER-type nuclear reactors under various accident, transient or normal operation conditions. In this talk we present verification results of the BAGIRA code, obtained on the basis of different experiments performed on special and integral thermohydraulic experimental facilities as well as on real NPPs. Special attention is paid to the verification of three-dimensional flow models. Besides that we expose new results of the code benchmark analysis made on the basis of two recent LOCA-type experiments - 'Leak 2 x 25% from the hot leg double-side rupture' and 'Leak 3% from the cold leg' - performed on the PSB-VVER integral test facility (Electrogorsk Research and Engineering Center, Electrogorsk, Russia) - the most up-to-date Russian large-scale four-loop unit which has been designed for modelling the primary circuit of VVER-1000 type reactors. (authors)
Reaction path of energetic materials using THOR code
Durães, L.; Campos, J.; Portugal, A.
1998-07-01
The method of predicting reaction path, using THOR code, allows for isobar and isochor adiabatic combustion and CJ detonation regimes, the calculation of the composition and thermodynamic properties of reaction products of energetic materials. THOR code assumes the thermodynamic equilibria of all possible products, for the minimum Gibbs free energy, using HL EoS. The code allows the possibility of estimating various sets of reaction products, obtained successively by the decomposition of the original reacting compound, as a function of the released energy. Two case studies of thermal decomposition procedure were selected, calculated and discussed—pure Ammonium Nitrate and its based explosive ANFO, and Nitromethane—because their equivalence ratio is respectively lower, near and greater than the stoicheiometry. Predictions of reaction path are in good correlation with experimental values, proving the validity of proposed method.
Thermal hydraulic analysis of Pb-Bi cooled HYPER fuel assemblies using SLTHEN code
International Nuclear Information System (INIS)
Tak, Nam Il; Song, Tae Y.; Park, Won S.; Kim, Chang Hyun
2002-12-01
In the present work, the existing SLTHEN code, which had been originally developed for subchannel analysis of sodium cooled fast reactors, was modified and applied to the Pb-Bi cooled HYPER core which consists of 237 fuel assemblies (TRU assemblies). In the analysis of single fuel assembly having chopped cosine power profile, the validation and the assessment of usefulness of the modified SLTHEN were focused. In the quantitative comparison, the results of the modified SLTHEN agreed well with those of analytical calculations and of MATRA. For the qualitative approaches, the sensitivity calculations for intra-assembly gap flow and turbulent mixing parameter were used. The sensitivity analysis results showed that the modified SLTHEN can provide reasonable simulations of subchannel thermal hydraulics. In particular, turbulent mixing parameter which is known as the most uncertain parameter in subchannel analyses did not affect largely the maximum cladding temperature. Therefore, it can be said that the results of single assembly show the usefulness of the modified SLTHEN code for thermal hydraulic analysis and design of HYPER under the conceptual design stage. In order to assess intra-assembly heat transfer, subchannel analyses were implemented for two types of 7 assemblies; 1) artificial 7 fuel assemblies to maximize intra-assembly heat transfer, 2) central 7 fuel assemblies in the HYPER reference core. The results showed that the modified SLTHEN can reasonably simulate intra-heat transfer and the amount of intra-assembly heat transfer is not so large in HYPER conditions. Particularly, intra-heat transfer did not affect the maximum coolant and the maximum cladding temperatures which are major parameters in conceptual core designs. The capability of full core thermal hydraulic analysis was confirmed by the analysis of 45 fuel assemblies in 1/6 HYPER core at the first cycle. The SLTHEN predicted that the reference design parameters are acceptable in terms of thermal
International Nuclear Information System (INIS)
Choi, Sun Rock; Cho, Chung Ho; Kim, Sang Ji
2011-01-01
In an SFR core analysis, a hot channel factors (HCF) method is most commonly used to evaluate uncertainty. It was employed to the early design such as the CRBRP and IFR. In other ways, the improved thermal design procedure (ITDP) is able to calculate the overall uncertainty based on the Root Sum Square technique and sensitivity analyses of each design parameters. The Monte Carlo method (MCM) is also employed to estimate the uncertainties. In this method, all the input uncertainties are randomly sampled according to their probability density functions and the resulting distribution for the output quantity is analyzed. Since an uncertainty analysis is basically calculated from the temperature distribution in a subassembly, the core thermal-hydraulic modeling greatly affects the resulting uncertainty. At KAERI, the SLTHEN and MATRA-LMR codes have been utilized to analyze the SFR core thermal-hydraulics. The SLTHEN (steady-state LMR core thermal hydraulics analysis code based on the ENERGY model) code is a modified version of the SUPERENERGY2 code, which conducts a multi-assembly, steady state calculation based on a simplified ENERGY model. The detailed subchannel analysis code MATRA-LMR (Multichannel Analyzer for Steady-State and Transients in Rod Arrays for Liquid Metal Reactors), an LMR version of MATRA, was also developed specifically for the SFR core thermal-hydraulic analysis. This paper describes comparative studies for core thermal-hydraulic models. The subchannel analysis and a hot channel factors based uncertainty evaluation system is established to estimate the core thermofluidic uncertainties using the MATRA-LMR code and the results are compared to those of the SLTHEN code
International Nuclear Information System (INIS)
Seitz, M.; Johnson, M.; Hübner, S.
2017-01-01
Highlights: • Integration of a latent heat thermal energy storage system into a solar direct steam generation power cycle. • Parametric study of solar field and storage size for determination of the optimal layout. • Evaluation of storage impact on the economic performance of the solar thermal power plant. • Economic comparison of new direct steam generation plant layout with state-of-the-art oil plant layout. - Abstract: One possible way to further reduce levelized costs of electricity of concentrated solar thermal energy is to directly use water/steam as the primary heat transfer fluid within a concentrated collector field. This so-called direct steam generation offers the opportunity of higher operating temperatures and better exergy efficiency. A technical challenge of the direct steam generation technology compared to oil-driven power cycles is a competitive storage technology for heat transfer fluids with a phase change. Latent heat thermal energy storages are suitable for storing heat at a constant temperature and can be used for direct steam generation power plants. The calculation of the economic impact of an economically optimized thermal energy storage system, based on a latent heat thermal energy storage system with phase change material, is the main focus of the presented work. To reach that goal, a thermal energy storage system for a direct steam generation power plant with parabolic troughs in the solar field was thermally designed to determine the boundary conditions. This paper discusses the economic impact of the designed thermal energy storage system based on the levelized costs of electricity results, provided via a wide parametric study. A state-of-the-art power cycle with a primary and a secondary heat transfer fluid and a two-tank thermal energy storage is used as a benchmark technology for electricity generation with solar thermal energy. The benchmark and direct steam generation systems are compared to each other, based respectively
Energy Technology Data Exchange (ETDEWEB)
Uwaba, Tomoyuki, E-mail: uwaba.tomoyuki@jaea.go.jp; Ohshima, Hiroyuki; Ito, Masahiro
2017-06-15
Highlights: • The coupled computational code system allowed for mechanical and thermal-hydraulic analyses in a fast reactor fuel subassembly. • In this system interactive calculations between flow area deformations and coolant temperature changes are repeated to their convergence state. • Effects on bundle-duct interaction on coolant temperature distributions were investigated by using the code system. - Abstract: The coupled numerical analysis of mechanical and thermal-hydraulic behaviors was performed for a wire-wrapped fuel pin bundle subassembly irradiated in a fast reactor. For the analysis, the fuel pin bundle deformation analysis code BAMBOO and the thermal-hydraulic analysis code ASFRE exchanged the deformation and temperature analysis results through the iterative calculations to attain convergence corresponding to the static balance between deformation and temperature. The analysis by the coupled code system showed that the radial distribution of coolant temperature in the subassembly tended to flatten as a result of the fuel pin bundle deformation governed by cladding void swelling and irradiation creep. Such flattening of temperature distribution was slightly observed as a result of fuel pin bowings due to the cladding-wire interaction even when no bundle-duct interaction occurred. The effect of the spacer wire-pitch on deformation and thermal-hydraulics was also investigated in this study.
Maximizing the energy storage performance of phase change thermal storage systems
Energy Technology Data Exchange (ETDEWEB)
Amin, N.A.M.; Bruno, F.; Belusko, M. [South Australia Univ., Mawson Lakes, South Australia (Australia). Inst. for Sustainable Systems and Technologies
2009-07-01
The demand for electricity in South Australia is highly influenced by the need for refrigeration and air-conditioning. An extensive literature review has been conducted on the use of phase change materials (PCMs) in thermal storage systems. PCMs use latent heat at the solid-liquid phase transition point to store thermal energy. They are considered to be useful as a thermal energy storage (TES) material because they can provide much higher energy storage densities compared to conventional sensible thermal storage materials. This paper reviewed the main disadvantages of using PCMs for energy storage, such as low heat transfer, super cooling and system design issues. Other issues with PCMs include incongruence and corrosion of heat exchanger surfaces. The authors suggested that in order to address these problems, future research should focus on maximizing heat transfer by optimizing the configuration of the encapsulation through a parametric analysis using a PCM numerical model. The effective conductivity in encapsulated PCMs in a latent heat thermal energy storage (LHTES) system can also be increased by using conductors in the encapsulation that have high thermal conductivity. 47 refs., 1 tab., 1 fig.