Computational Design of Advanced Nuclear Fuels

International Nuclear Information System (INIS)

Savrasov, Sergey; Kotliar, Gabriel; Haule, Kristjan

2014-01-01

The objective of the project was to develop a method for theoretical understanding of nuclear fuel materials whose physical and thermophysical properties can be predicted from first principles using a novel dynamical mean field method for electronic structure calculations. We concentrated our study on uranium, plutonium, their oxides, nitrides, carbides, as well as some rare earth materials whose 4f eletrons provide a simplified framework for understanding complex behavior of the f electrons. We addressed the issues connected to the electronic structure, lattice instabilities, phonon and magnon dynamics as well as thermal conductivity. This allowed us to evaluate characteristics of advanced nuclear fuel systems using computer based simulations and avoid costly experiments.

Enhancing instruction in Fuels and Combustion Laboratory via a developed computer-assisted program for establishing efficient coal-diesel oil mixture (CDOM) fuel proportions

Energy Technology Data Exchange (ETDEWEB)

Maglaya, A.B. [La Salle University, Manila (Philippines). Dept. of Mechanical Engineering

2004-07-01

This paper discusses the relevance of digital computation in Fuels and Combustion Laboratory experiments used by the senior students of the Department of Mechanical Engineering, De La Salle University-Manila, Philippines. One of the students' experiments involved the determination of the most efficient CDOM fuel proportion as alternative fuel to diesel oil for steam generators and other industrial applications. Theoretical calculations show that it requires tedious and repetitive computations. A computer-assisted program was developed to lessen the time-consuming activities. The formulation of algorithms were based on the system of equations of the heat interaction between the CDOM fuel, combustion air and products of combustion and by applying the principles of mass and energy equations (or the First Law of Thermodynamics) for reacting systems were utilized. The developed computer-assisted program output verified alternative fuel selected through actual experimentation.

Computational methods for more fuel-efficient ship

NARCIS (Netherlands)

Koren, B.

2008-01-01

The flow of water around a ship powered by a combustion engine is a key factor in the ship's fuel consumption. The simulation of flow patterns around ship hulls is therefore an important aspect of ship design. While lengthy computations are required for such simulations, research by Jeroen Wackers

35-We polymer electrolyte membrane fuel cell system for notebook computer using a compact fuel processor

Energy Technology Data Exchange (ETDEWEB)

Son, In-Hyuk; Shin, Woo-Cheol; Lee, Sung-Chul; Ahn, Jin-Gu; Han, Sang-Il; kweon, Ho-Jin; Kim, Ju-Yong; Park, Jun-Yong [Energy 1 Group, Energy Laboratory at Corporate R and D Center in Samsung SDI Co., Ltd., 575, Shin-dong, Yeongtong-gu, Suwon-si, Gyeonggi-do 443-731 (Korea); Lee, Yong-Kul [Department of Chemical Engineering, Dankook University, Youngin 448-701 (Korea); Kim, Moon-Chan [Department of Environmental Engineering, Chongju University, Chongju 360-764 (Korea)

2008-10-15

A polymer electrolyte membrane fuel cell (PEMFC) system is developed to power a notebook computer. The system consists of a compact methanol-reforming system with a CO preferential oxidation unit, a 16-cell PEMFC stack, and a control unit for the management of the system with a d.c.-d.c. converter. The compact fuel-processor system (260 cm{sup 3}) generates about 1.2 L min{sup -1} of reformate, which corresponds to 35 We, with a low CO concentration (<30 ppm, typically 0 ppm), and is thus proven to be capable of being targetted at notebook computers. (author)

Computer modelling of water reactor fuel element performance and life time

International Nuclear Information System (INIS)

Bibilashvili, Yu.K.; Golovnin, I.S.; Elesin, V.F.

1983-01-01

Well calibrated models and methods of calculation permit the confident prediction of fuel element behaviour under most different operational conditions; based on the prediction of this kind one can improve designs and fuel element behaviour. Therefore, in the Soviet Union in the development of reactor cores for NPP one of the leading parts is given to design problems associated with computer modelling of fuel element performance and reliability. Special attention is paid to methods of calculation that permit the prediction of fuel element behaviour under conditions which either make experimental studies very complicated (practically impossible) or require laborious and expensive in-pile tests. Primarily it concerns accidents of different types, off-normal conditions, transients, fuel element behaviour at high burn-up, when an accumulation of a great amount of fission fragments is accompanied by changes in physical and mechanical properties as induced by irradiation damage, mechanical fatigue, physical and chemical reactions with a coolant, fission products etc. Some major computer modelling programs for the prediction of water reactor fuel behaviour are briefly described below and tendencies in the further development of work in this area are summarized

Evolution of the Darlington NGS fuel handling computer systems

International Nuclear Information System (INIS)

Leung, V.; Crouse, B.

1996-01-01

The ability to improve the capabilities and reliability of digital control systems in nuclear power stations to meet changing plant and personnel requirements is a formidable challenge. Many of these systems have high quality assurance standards that must be met to ensure adequate nuclear safety. Also many of these systems contain obsolete hardware along with software that is not easily transported to newer technology computer equipment. Combining modern technology upgrades into a system of obsolete hardware components is not an easy task. Lastly, as users become more accustomed to using modern technology computer systems in other areas of the station (e.g. information systems), their expectations of the capabilities of the plant systems increase. This paper will present three areas of the Darlington NGS fuel handling computer system that have been or are in the process of being upgraded to current technology components within the framework of an existing fuel handling control system. (author). 3 figs

Evolution of the Darlington NGS fuel handling computer systems

Energy Technology Data Exchange (ETDEWEB)

Leung, V; Crouse, B [Ontario Hydro, Bowmanville (Canada). Darlington Nuclear Generating Station

1997-12-31

The ability to improve the capabilities and reliability of digital control systems in nuclear power stations to meet changing plant and personnel requirements is a formidable challenge. Many of these systems have high quality assurance standards that must be met to ensure adequate nuclear safety. Also many of these systems contain obsolete hardware along with software that is not easily transported to newer technology computer equipment. Combining modern technology upgrades into a system of obsolete hardware components is not an easy task. Lastly, as users become more accustomed to using modern technology computer systems in other areas of the station (e.g. information systems), their expectations of the capabilities of the plant systems increase. This paper will present three areas of the Darlington NGS fuel handling computer system that have been or are in the process of being upgraded to current technology components within the framework of an existing fuel handling control system. (author). 3 figs.

A computational methodology for formulating gasoline surrogate fuels with accurate physical and chemical kinetic properties

KAUST Repository

Ahmed, Ahfaz

2015-03-01

Gasoline is the most widely used fuel for light duty automobile transportation, but its molecular complexity makes it intractable to experimentally and computationally study the fundamental combustion properties. Therefore, surrogate fuels with a simpler molecular composition that represent real fuel behavior in one or more aspects are needed to enable repeatable experimental and computational combustion investigations. This study presents a novel computational methodology for formulating surrogates for FACE (fuels for advanced combustion engines) gasolines A and C by combining regression modeling with physical and chemical kinetics simulations. The computational methodology integrates simulation tools executed across different software platforms. Initially, the palette of surrogate species and carbon types for the target fuels were determined from a detailed hydrocarbon analysis (DHA). A regression algorithm implemented in MATLAB was linked to REFPROP for simulation of distillation curves and calculation of physical properties of surrogate compositions. The MATLAB code generates a surrogate composition at each iteration, which is then used to automatically generate CHEMKIN input files that are submitted to homogeneous batch reactor simulations for prediction of research octane number (RON). The regression algorithm determines the optimal surrogate composition to match the fuel properties of FACE A and C gasoline, specifically hydrogen/carbon (H/C) ratio, density, distillation characteristics, carbon types, and RON. The optimal surrogate fuel compositions obtained using the present computational approach was compared to the real fuel properties, as well as with surrogate compositions available in the literature. Experiments were conducted within a Cooperative Fuels Research (CFR) engine operating under controlled autoignition (CAI) mode to compare the formulated surrogates against the real fuels. Carbon monoxide measurements indicated that the proposed surrogates

Computer simulation of fuel behavior during loss-of-flow accidents in a gas-cooled fast reactor

International Nuclear Information System (INIS)

Wehner, T.R.

1980-01-01

The sequence of events in a loss-of-flow accident without reactor shutdown in a gas-cooled fast breeder reactor is strongly influenced by the manner in which the fuel deforms. In order to predict the mode of initial gross fuel deformation, welling, melting or cracking, a thermomechanical computer simulation program was developed. Methods and techniques used make the simulation an economical, efficient, and flexible engineering tool. An innovative application of the enthalpy model within a finite difference scheme is used to caculate temperatures in the fuel rod. The method of successive elastic solutions is used to calculate the thermoelastic-creep response. Calculated stresses are compared with a brittle-fracture stress criterion. An independent computer code is used to calculate fission-gas-induced fuel swelling. Results obtained with the computer simulation indicate that swelling is not a mode of initial fuel deformation. Faster transients result in fuel melting, while slower transients result in fuel cracking. For investigated faster coolant flow coastdowns with time constants of 1 second and 10 seconds, compressive stresses in the outer radial portion of the fuel limit fuel swelling and inhibit fuel cracking. For a slower coolant flow coastdown with a 300 second time constant, tensile stresses in the outer radial portion of the fuel induce early fuel cracking before any melting or significant fuel swelling has occurred. Suggestions for further research are discussed. A derived noniterative solution for mechanics calculations may offer an order of magnitude decrease in computational effort

Computer imaging of EBR-II fuel handling equipment

International Nuclear Information System (INIS)

Peters, G.G.; Hansen, L.H.

1995-01-01

This paper describes a three-dimensional graphics application used to visualize the positions of remotely operated fuel handling equipment in the EBR-II reactor. A three-dimensional (3D) visualization technique is necessary to simulate direct visual observation of the transfers of fuel and experiments into and out of the reactor because the fuel handling equipment is submerged in liquid sodium and therefore is not visible to the operator. The system described in this paper uses actual signals to drive a three-dimensional computer-generated model in real-time in response to movements of equipment in the plant This paper will present details on how the 3D model of the intank equipment was created and how real-time dynamic behavior was added to each of the moving components

Computer simulation of HTGR fuel microspheres using a Monte-Carlo statistical approach

International Nuclear Information System (INIS)

Hedrick, C.E.

1976-01-01

The concept and computational aspects of a Monte-Carlo statistical approach in relating structure of HTGR fuel microspheres to the uranium content of fuel samples have been verified. Results of the preliminary validation tests and the benefits to be derived from the program are summarized

Workstation computer systems for in-core fuel management

International Nuclear Information System (INIS)

Ciccone, L.; Casadei, A.L.

1992-01-01

The advancement of powerful engineering workstations has made it possible to have thermal-hydraulics and accident analysis computer programs operating efficiently with a significant performance/cost ratio compared to large mainframe computer. Today, nuclear utilities are acquiring independent engineering analysis capability for fuel management and safety analyses. Computer systems currently available to utility organizations vary widely thus requiring that this software be operational on a number of computer platforms. Recognizing these trends Westinghouse adopted a software development life cycle process for the software development activities which strictly controls the development, testing and qualification of design computer codes. In addition, software standards to ensure maximum portability were developed and implemented, including adherence to FORTRAN 77, and use of uniform system interface and auxiliary routines. A comprehensive test matrix was developed for each computer program to ensure that evolution of code versions preserves the licensing basis. In addition, the results of such test matrices establish the Quality Assurance basis and consistency for the same software operating on different computer platforms. (author). 4 figs

CRACKEL: a computer code for CFR fuel management calculations

International Nuclear Information System (INIS)

Burstall, R.F.; Ball, M.A.; Thornton, D.E.J.

1975-12-01

The CRACKLE computer code is designed to perform rapid fuel management surveys of CFR systems. The code calculates overall features such as reactivity, power distributions and breeding gain, and also calculates for each sub-assembly plutonium content and power output. A number of alternative options are built into the code, in order to permit different fuel management strategies to be calculated, and to perform more detailed calculations when necessary. A brief description is given of the methods of calculation, and the input facilities of CRACKLE, with examples. (author)

Neutron physics computation of CERCA fuel elements for Maria Reactor

International Nuclear Information System (INIS)

Andrzejewski, K.J.; Kulikowska, T.; Marcinkowska, Z.

2008-01-01

Neutron physics parameters of CERCA design fuel elements were calculated in the framework of the RERTR (Reduced Enrichment for Research and Test Reactors) program for Maria reactor. The analysis comprises burnup of experimental CERCA design fuel elements for 4 cycles in Maria Reactor To predict the behavior of the mixed core the differences between the CERCA fuel (485 g U-235 as U 3 Si 2 , 5 fuel tubes, low enrichment 19.75 % - LEU) and the presently used MR-6 fuel (430 g as UO 2 , 6 fuel tubes, high enrichment 36 % - HEU) had to be taken into account. The basic tool used in neutron-physics analysis of Maria reactor is program REBUS using in its dedicated libraries of effective microscopic cross sections. The cross sections were prepared using WIMS-ANL code, taking into account the actual structure, temperature and material composition of the fuel elements required preparation of new libraries.The problem is described in the first part of the present paper. In the second part the applicability of the new library is shown on the basis of the fuel core computational analysis. (author)

Nuclear fuel cycle risk assessment: survey and computer compilation of risk-related literature

International Nuclear Information System (INIS)

Yates, K.R.; Schreiber, A.M.; Rudolph, A.W.

1982-10-01

The US Nuclear Regulatory Commission has initiated the Fuel Cycle Risk Assessment Program to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. Both the once-through cycle and plutonium recycle are being considered. A previous report generated by this program defines and describes fuel cycle facilities, or elements, considered in the program. This report, the second from the program, describes the survey and computer compilation of fuel cycle risk-related literature. Sources of available information on the design, safety, and risk associated with the defined set of fuel cycle elements were searched and documents obtained were catalogued and characterized with respect to fuel cycle elements and specific risk/safety information. Both US and foreign surveys were conducted. Battelle's computer-based BASIS information management system was used to facilitate the establishment of the literature compilation. A complete listing of the literature compilation and several useful indexes are included. Future updates of the literature compilation will be published periodically. 760 annotated citations are included

Emission computer tomography on a Dodewaard mixed oxide fuel pin

International Nuclear Information System (INIS)

Buurveld, H.A.; Dassel, G.

1993-12-01

A nondestructive technique as well as a destructive PIE technique have been used to verify the results obtained with a newly 8-e computer tomography (GECT) system. Multi isotope Scanning (MIS), electron probe micro analysis (EPMA) and GECT were used on a mixed oxide (MOX) fuel rod from the Dodewaard reactor with an average burnup of 24 MWd/kg fuel. GECT shows migration of Cs to the periphery of fuel pellets and to radial cracks and pores in the fuel, whereas MIS shows Cs migration to pellet interfaces. The EPMA technique appeared not to be useful to show migration of Cs but, it shows the distribution of fission products from Pu. EPMA clearly shows the distribution of fission products from Pu, but did not reveal the Cs-migration. (orig./HP)

Estimate of fuel burnup spatial a multipurpose reactor in computer simulation

International Nuclear Information System (INIS)

Santos, Nadia Rodrigues dos; Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes

2015-01-01

In previous research, which aimed, through computer simulation, estimate the spatial fuel burnup for the research reactor benchmark, material test research - International Atomic Energy Agency (MTR/IAEA), it was found that the use of the code in FORTRAN language, based on the diffusion theory of neutrons and WIMSD-5B, which makes cell calculation, bespoke be valid to estimate the spatial burnup other nuclear research reactors. That said, this paper aims to present the results of computer simulation to estimate the space fuel burnup of a typical multipurpose reactor, plate type and dispersion. the results were considered satisfactory, being in line with those presented in the literature. for future work is suggested simulations with other core configurations. are also suggested comparisons of WIMSD-5B results with programs often employed in burnup calculations and also test different methods of interpolation values obtained by FORTRAN. Another proposal is to estimate the burning fuel, taking into account the thermohydraulics parameters and the appearance of xenon. (author)

TEMP: a computer code to calculate fuel pin temperatures during a transient

International Nuclear Information System (INIS)

Bard, F.E.; Christensen, B.Y.; Gneiting, B.C.

1980-04-01

The computer code TEMP calculates fuel pin temperatures during a transient. It was developed to accommodate temperature calculations in any system of axi-symmetric concentric cylinders. When used to calculate fuel pin temperatures, the code will handle a fuel pin as simple as a solid cylinder or as complex as a central void surrounded by fuel that is broken into three regions by two circumferential cracks. Any fuel situation between these two extremes can be analyzed along with additional cladding, heat sink, coolant or capsule regions surrounding the fuel. The one-region version of the code accurately calculates the solution to two problems having closed-form solutions. The code uses an implicit method, an explicit method and a Crank-Nicolson (implicit-explicit) method

Calculation of demands for nuclear fuels and fuel cycle services. Description of computer model and strategies developed by Working Group 1

International Nuclear Information System (INIS)

Working Group 1 examined a range of reactor deployment strategies and fuel cycle options, in oder to estimate the range of nuclear fuel requirements and fuel cycle service needs which would result. The computer model, its verification in comparison with other models, the strategies to be examined through use of the model, and the range of results obtained are described

Development of a graphical interface computer code for reactor fuel reloading optimization

International Nuclear Information System (INIS)

Do Quang Binh; Nguyen Phuoc Lan; Bui Xuan Huy

2007-01-01

This report represents the results of the project performed in 2007. The aim of this project is to develop a graphical interface computer code that allows refueling engineers to design fuel reloading patterns for research reactor using simulated graphical model of reactor core. Besides, this code can perform refueling optimization calculations based on genetic algorithms as well as simulated annealing. The computer code was verified based on a sample problem, which relies on operational and experimental data of Dalat research reactor. This code can play a significant role in in-core fuel management practice at nuclear research reactor centers and in training. (author)

STAT, GAPS, STRAIN, DRWDIM: a system of computer codes for analyzing HTGR fuel test element metrology data. User's manual

Energy Technology Data Exchange (ETDEWEB)

Saurwein, J.J.

1977-08-01

A system of computer codes has been developed to statistically reduce Peach Bottom fuel test element metrology data and to compare the material strains and fuel rod-fuel hole gaps computed from these data with HTGR design code predictions. The codes included in this system are STAT, STRAIN, GAPS, and DRWDIM. STAT statistically evaluates test element metrology data yielding fuel rod, fuel body, and sleeve irradiation-induced strains; fuel rod anisotropy; and additional data characterizing each analyzed fuel element. STRAIN compares test element fuel rod and fuel body irradiation-induced strains computed from metrology data with the corresponding design code predictions. GAPS compares test element fuel rod, fuel hole heat transfer gaps computed from metrology data with the corresponding design code predictions. DRWDIM plots the measured and predicted gaps and strains. Although specifically developed to expedite the analysis of Peach Bottom fuel test elements, this system can be applied, without extensive modification, to the analysis of Fort St. Vrain or other HTGR-type fuel test elements.

Development of a computer program for the cost analysis of spent fuel management

International Nuclear Information System (INIS)

Choi, Heui Joo; Lee, Jong Youl; Choi, Jong Won; Cha, Jeong Hun; Whang, Joo Ho

2009-01-01

So far, a substantial amount of spent fuels have been generated from the PWR and CANDU reactors. They are being temporarily stored at the nuclear power plant sites. It is expected that the temporary storage facility will be full of spent fuels by around 2016. The government plans to solve the problem by constructing an interim storage facility soon. The radioactive management act was enacted in 2008 to manage the spent fuels safety in Korea. According to the act, the radioactive waste management fund which will be used for the transportation, interim storage, and the final disposal of spent fuels has been established. The cost for the management of spent fuels is surprisingly high and could include a lot of uncertainty. KAERI and Kyunghee University have developed cost estimation tools to evaluate the cost for a spent fuel management based on an engineering design and calculation. It is not easy to develop a tool for a cost estimation under the situation that the national policy on a spent fuel management has not yet been fixed at all. Thus, the current version of the computer program is based on the current conceptual design of each management system. The main purpose of this paper is to introduce the computer program developed for the cost analysis of a spent fuel management. In order to show the application of the program, a spent fuel management scenario is prepared, and the cost for the scenario is estimated

Development of a Computer Program for the Analysis Logistics of PWR Spent Fuels

International Nuclear Information System (INIS)

Choi, Heui Joo; Choi, Jong Won; Cha, Jeong Hun

2008-01-01

It is expected that the temporary storage facilities at the nuclear power plants will be full of the spent fuels within 10 years. Provided that a centralized interim storage facility is constructed along the coast of the Korean peninsula to solve this problem, a substantial amount of spent fuels should be transported by sea or by land every year. In this paper we developed a computer program for the analysis of transportation logistics of the spent fuels from 4 different nuclear power plant sites to the hypothetical centralized interim storage facility and the final repository. Mass balance equations were used to analyze the logistics between the nuclear power plants and the interim storage facility. To this end a computer program, CASK, was developed by using the VISUAL BASIC language. The annual transportation rates of spent fuels from the four nuclear power plant sites were determined by using the CASK program. The parameter study with the program illustrated the easiness of logistics analysis. The program could be used for the cost analysis of the spent fuel transportation as well.

FRAPCON-2: A Computer Code for the Calculation of Steady State Thermal-Mechanical Behavior of Oxide Fuel Rods

Energy Technology Data Exchange (ETDEWEB)

Berna, G. A; Bohn, M. P.; Rausch, W. N.; Williford, R. E.; Lanning, D. D.

1981-01-01

FRAPCON-2 is a FORTRAN IV computer code that calculates the steady state response of light Mater reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, deformation, and tai lure histories of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (a) heat conduction through the fuel and cladding, (b) cladding elastic and plastic deformation, (c) fuel-cladding mechanical interaction, (d) fission gas release, (e} fuel rod internal gas pressure, (f) heat transfer between fuel and cladding, (g) cladding oxidation, and (h) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat transfer correlations. FRAPCON-2 is programmed for use on the CDC Cyber 175 and 176 computers. The FRAPCON-2 code Is designed to generate initial conditions for transient fuel rod analysis by either the FRAP-T6 computer code or the thermal-hydraulic code, RELAP4/MOD7 Version 2.

Neutron Computed Tomography of Freeze/thaw Phenomena in Polymer Electrolyte Fuel Cells

Energy Technology Data Exchange (ETDEWEB)

Matthew M. Mech; Jack Brenizer; Kenan Unlu; A.K. Heller

2008-12-12

This report summarizes the final year's progress of the three-year NEER program. The overall objectives of this program were to 1) design and construct a sophisticated hight-resolution neutron computed tomography (NCT) facility, 2) develop novel and sophisticated liquid water and ice quantification analysis software for computed tomography, and 3) apply the advanced software and NCT capability to study liquid and ice distribution in polymer electrolyte fuel cells (PEFCs) under cold-start conditions. These objectives have been accomplished by the research team, enabling a new capability for advanced 3D image quantification with neutron imaging for fuel cell and other applications. The NCT water quantification methodology and software will greatly add to the capabilities of the neutron imaging community, and the quantified liquid water and ice distribution provided by its application to PEFCs will enhance understanding and guide design in the fuel cell community.

Development, verification and validation of the fuel channel behaviour computer code FACTAR

Energy Technology Data Exchange (ETDEWEB)

Westbye, C J; Brito, A C; MacKinnon, J C; Sills, H E; Langman, V J [Ontario Hydro, Toronto, ON (Canada)

1996-12-31

FACTAR (Fuel And Channel Temperature And Response) is a computer code developed to simulate the transient thermal and mechanical behaviour of 37-element or 28-element fuel bundles within a single CANDU fuel channel for moderate loss of coolant accident conditions including transition and large break LOCA`s (loss of coolant accidents) with emergency coolant injection assumed available. FACTAR`s predictions of fuel temperature and sheath failure times are used to subsequent assessment of fission product releases and fuel string expansion. This paper discusses the origin and development history of FACTAR, presents the mathematical models and solution technique, the detailed quality assurance procedures that are followed during development, and reports the future development of the code. (author). 27 refs., 3 figs.

Simulation of nuclear fuel rods by using process computer-controlled power for indirect electrically heated rods

International Nuclear Information System (INIS)

Malang, S.

1975-11-01

An investigation was carried out to determine how the simulation of nuclear fuel rods with indirect electrically heated rods could be improved by use of a computer to control the electrical power during a loss-of-coolant accident (LOCA). To aid in the experiment, a new version of the HETRAP code was developed which simulates a LOCA with heater rod power controlled by a computer that adjusts rod power during a blowdown to minimize the difference in heat flux of the fuel and heater rods. Results show that without computer control of heater rod power, only the part of a blowdown up to the time when the heat transfer mode changes from nucleate boiling to transition or film boiling can be simulated well and then only for short times. With computer control, the surface heat flux and temperature of an electrically heated rod can be made nearly identical to that of a reactor fuel rod with the same cooling conditions during much of the LOCA. A small process control computer can be used to achieve close simulation of a nuclear fuel rod with an indirect electrically heated rod

Benchmarking Computational Fluid Dynamics for Application to PWR Fuel

International Nuclear Information System (INIS)

Smith, L.D. III; Conner, M.E.; Liu, B.; Dzodzo, B.; Paramonov, D.V.; Beasley, D.E.; Langford, H.M.; Holloway, M.V.

2002-01-01

The present study demonstrates a process used to develop confidence in Computational Fluid Dynamics (CFD) as a tool to investigate flow and temperature distributions in a PWR fuel bundle. The velocity and temperature fields produced by a mixing spacer grid of a PWR fuel assembly are quite complex. Before using CFD to evaluate these flow fields, a rigorous benchmarking effort should be performed to ensure that reasonable results are obtained. Westinghouse has developed a method to quantitatively benchmark CFD tools against data at conditions representative of the PWR. Several measurements in a 5 x 5 rod bundle were performed. Lateral flow-field testing employed visualization techniques and Particle Image Velocimetry (PIV). Heat transfer testing involved measurements of the single-phase heat transfer coefficient downstream of the spacer grid. These test results were used to compare with CFD predictions. Among the parameters optimized in the CFD models based on this comparison with data include computational mesh, turbulence model, and boundary conditions. As an outcome of this effort, a methodology was developed for CFD modeling that provides confidence in the numerical results. (authors)

Methods and computer programs for PWR's fuel management: Programs Sothis and Ciclon

International Nuclear Information System (INIS)

Aragones, J.M.; Corella, M.R.; Martinez-Val, J.M.

1976-01-01

Methos and computer programs developed at JEN for fuel management in PWR are discussed, including scope of model, procedures for sistematic selection of alternatives to be evaluated, basis of model for neutronic calculation, methods for fuel costs calculation, procedures for equilibrium and trans[tion cycles calculation with Soth[s and Ciclon codes and validation of methods by comparison of results with others of reference (author) ' [es

Development of computer models for fuel element behaviour in water reactors

International Nuclear Information System (INIS)

Gittus, J.H.

1987-03-01

Description of fuel behaviour during normal operation transients and accident conditions has always represented a most challenging and important problem. Reliable predictions constitute a basic demand for safety based calculations, for design purposes and for fuel performance. Therefore, computer codes based on deterministic and probabilistic models were developed. Possibility of comprehensive descriptions of the phenomena is precluded in view of the great number of individual processes, involving physical, chemical, thermohydraulical and mechanical parameters, to be considered in a wide range of situations. In case of fast thermal transients predictive capability is limited by the kinetics of evolution of the system and its eventual dynamic behaviour. Evidently, probabilistic approaches are also limited by the sparcity and limited breadth of the impirical data base. Code predictions have to be evaluated against power reactor data, results from simulation experiments and, if possible, include cross validation of different codes and validation of sub-models. Progress on this subject is reviewed in this report, which completes the co-ordinated research programme on 'Development of Computer Models for Fuel Element Behaviour in Water Reactors' (D-COM), initiated under the auspices of the IAEA in 1981

Techniques for computing reactivity changes caused by fuel axial expansion in LMR's

International Nuclear Information System (INIS)

Khalil, H.

1988-01-01

An evaluation is made of the accuracy of methods used to compute reactivity changes caused by axial fuel relocation in fast reactors. Results are presented to demonstrate the validity of assumptions commonly made such as linearity of reactivity with fuel elongation, additivity of local reactivity contributions, and the adequacy of standard perturbation techniques. Accurate prediction of the reactivity loss caused by axial swelling of metallic fuel is shown to require proper representation of the burnup dependence of the expansion reactivity. Some accuracy limitations in the methods used in transient analyses, which are based on the use of fuel worth tables, are identified, and efficient ways to improve accuracy are described. Implementation of these corrections produced expansion reactivity estimates within 5% of higher-order method for a metal-fueled FFTF core representation. 18 refs., 3 figs., 3 tabs

Integration of computational modeling and experimental techniques to design fuel surrogates

DEFF Research Database (Denmark)

Choudhury, H.A.; Intikhab, S.; Kalakul, Sawitree

2017-01-01

performance. A simplified alternative is to develop surrogate fuels that have fewer compounds and emulate certain important desired physical properties of the target fuels. Six gasoline blends were formulated through a computer aided model based technique “Mixed Integer Non-Linear Programming” (MINLP...... Virtual Process-Product Design Laboratory (VPPD-Lab) are applied onto the defined compositions of the surrogate gasoline. The aim is to primarily verify the defined composition of gasoline by means of VPPD-Lab. ρ, η and RVP are calculated with more accuracy and constraints such as distillation curve...... and flash point on the blend design are also considered. A post-design experiment-based verification step is proposed to further improve and fine-tune the “best” selected gasoline blends following the computation work. Here, advanced experimental techniques are used to measure the RVP, ρ, η, RON...

Recovery Act: Advanced Direct Methanol Fuel Cell for Mobile Computing

Energy Technology Data Exchange (ETDEWEB)

Fletcher, James H. [University of North Florida; Cox, Philip [University of North Florida; Harrington, William J [University of North Florida; Campbell, Joseph L [University of North Florida

2013-09-03

ABSTRACT Project Title: Recovery Act: Advanced Direct Methanol Fuel Cell for Mobile Computing PROJECT OBJECTIVE The objective of the project was to advance portable fuel cell system technology towards the commercial targets of power density, energy density and lifetime. These targets were laid out in the DOE’s R&D roadmap to develop an advanced direct methanol fuel cell power supply that meets commercial entry requirements. Such a power supply will enable mobile computers to operate non-stop, unplugged from the wall power outlet, by using the high energy density of methanol fuel contained in a replaceable fuel cartridge. Specifically this project focused on balance-of-plant component integration and miniaturization, as well as extensive component, subassembly and integrated system durability and validation testing. This design has resulted in a pre-production power supply design and a prototype that meet the rigorous demands of consumer electronic applications. PROJECT TASKS The proposed work plan was designed to meet the project objectives, which corresponded directly with the objectives outlined in the Funding Opportunity Announcement: To engineer the fuel cell balance-of-plant and packaging to meet the needs of consumer electronic systems, specifically at power levels required for mobile computing. UNF used existing balance-of-plant component technologies developed under its current US Army CERDEC project, as well as a previous DOE project completed by PolyFuel, to further refine them to both miniaturize and integrate their functionality to increase the system power density and energy density. Benefits of UNF’s novel passive water recycling MEA (membrane electrode assembly) and the simplified system architecture it enabled formed the foundation of the design approach. The package design was hardened to address orientation independence, shock, vibration, and environmental requirements. Fuel cartridge and fuel subsystems were improved to ensure effective fuel

SIEX3: A correlated computer code for prediction of fast reactor mixed oxide fuel and blanket pin performance

International Nuclear Information System (INIS)

Baker, R.B.; Wilson, D.R.

1986-04-01

The SIEX3 computer program was developed to calculate the fuel and cladding performance of oxide fuel and oxide blanket pins irradiated in the fast neutron environment of a liquid metal cooled reactor. The code is uniquely designed to be accurate yet quick running and use a minimum of computer core storage. This was accomplished through the correlation of physically based models to very large data bases of irradiation test results. Data from over 200 fuel pins and over 800 transverse fuel microscopy samples were used in the calibrations

Neutronics investigation of CANDU deuterium uranium 6 reactor fueled (transuranic-TH) O-2 using a computational method

Energy Technology Data Exchange (ETDEWEB)

Gholamzadeh, Zohreh; Mirvakili, Seyed Mohammad; Khalafi, Hossein [Reactor Research School, Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of)

2015-02-15

241Am, 243Am, and 237Np isotopes are among the most radiotoxic components of spent nuclear fuel. Recently, researchers have planned different incineration scenarios for the highly radiotoxic elements of nuclear waste in critical reactors. Computational methods are widely used to predict burnup rates of such nuclear wastes that are used under fuel matrixes in critical reactors. In this work, the Monte Carlo N-particle transport code was used to calculate the neutronic behavior of a transuranic (TRU)-bearing CANada Deuterium Uranium 6 reactor. The computational data showed that the 1.0% TRU-containing thorium-based fuel matrix presents higher proliferation resistance and TRU depletion rate than the other investigated fuel Matrixes. The fuel matrix includes higher negative temperature reactivity coefficients as well. The investigated thorium-based fuel matrix can be successfully used to decrease the production of highly radiotoxic isotopes.

CORCON: a computer program for modelling molten fuel/concrete interactions

International Nuclear Information System (INIS)

Muir, J.F.

1980-01-01

A computer program modelling the interaction between molten core materials and structural concrete is being developed to provide a capability for making quantitative estimates of reactor fuel-melt accidents. The principal phenomenological models, inter-component heat transfer, concrete erosion, and melt/gas chemical reactions, are described. A code test comparison calculation is discussed

A simplified computational scheme for thermal analysis of LWR spent fuel dry storage and transportation casks

International Nuclear Information System (INIS)

Kim, Chang Hyun

1997-02-01

A simplified computational scheme for thermal analysis of the LWR spent fuel dry storage and transportation casks has been developed using two-step thermal analysis method incorporating effective thermal conductivity model for the homogenized spent fuel assembly. Although a lot of computer codes and analytical models have been developed for application to the fields of thermal analysis of dry storage and/or transportation casks, some difficulties in its analysis arise from the complexity of the geometry including the rod bundles of spent fuel and the heat transfer phenomena in the cavity of cask. Particularly, if the disk-type structures such as fuel baskets and aluminium heat transfer fins are included, the thermal analysis problems in the cavity are very complex. To overcome these difficulties, cylindrical coordinate system is adopted to calculate the temperature profile of a cylindrical cask body using the multiple cylinder model as the step-1 analysis of the present study. In the step-2 analysis, Cartesian coordinate system is adopted to calculate the temperature distributions of the disk-type structures such as fuel basket and aluminium heat transfer fin using three- dimensional conduction analysis model. The effective thermal conductivity for homogenized spent fuel assembly based on Manteufel and Todreas model is incorporated in step-2 analysis to predict the maximum fuel temperature. The presented two-step computational scheme has been performed using an existing HEATING 7.2 code and the effective thermal conductivity for the homogenized spent fuel assembly has been calculated by additional numerical analyses. Sample analyses of five cases are performed for NAC-STC including normal transportation condition to examine the applicability of the presented simplified computational scheme for thermal analysis of the large LWR spent fuel dry storage and transportation casks and heat transfer characteristics in the cavity of the cask with the disk-type structures

Computational fluid dynamics modeling of two-phase flow in a BWR fuel assembly

International Nuclear Information System (INIS)

Andrey Ioilev; Maskhud Samigulin; Vasily Ustinenko; Simon Lo; Adrian Tentner

2005-01-01

Full text of publication follows: The goal of this project is to develop an advanced Computational Fluid Dynamics (CFD) computer code (CFD-BWR) that allows the detailed analysis of the two-phase flow and heat transfer phenomena in a Boiling Water Reactor (BWR) fuel bundle under various operating conditions. This code will include more fundamental physical models than the current generation of sub-channel codes and advanced numerical algorithms for improved computational accuracy, robustness, and speed. It is highly desirable to understand the detailed two-phase flow phenomena inside a BWR fuel bundle. These phenomena include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. Traditionally, the best analysis tools for the analysis of two-phase flow phenomena inside the BWR fuel assembly have been the sub-channel codes. However, the resolution of these codes is still too coarse for analyzing the detailed intra-assembly flow patterns, such as flow around a spacer element. Recent progress in Computational Fluid Dynamics (CFD), coupled with the rapidly increasing computational power of massively parallel computers, shows promising potential for the fine-mesh, detailed simulation of fuel assembly two-phase flow phenomena. However, the phenomenological models available in the commercial CFD programs are not as advanced as those currently being used in the sub-channel codes used in the nuclear industry. In particular, there are no models currently available which are able to reliably predict the nature of the flow regimes, and use the appropriate sub-models for those flow regimes. The CFD-BWR code is being developed as a customized module built on the foundation of the commercial CFD Code STAR-CD which provides general two-phase flow modeling capabilities. The paper describes the model development strategy which has been adopted by the development team for the

New weighted sum of gray gases model applicable to Computational Fluid Dynamics (CFD) modeling of oxy-fuel combustion

DEFF Research Database (Denmark)

Yin, Chungen; Johansen, Lars Christian Riis; Rosendahl, Lasse

2010-01-01

gases model (WSGGM) is derived, which is applicable to computational fluid dynamics (CFD) modeling of both air-fuel and oxy-fuel combustion. First, a computer code is developed to evaluate the emissivity of any gas mixture at any condition by using the exponential wide band model (EWBM...

Computer programs for the in-core fuel management of power reactors

International Nuclear Information System (INIS)

1981-08-01

This document gives a survey of the presently tested and used computer programs applicable to the in-core fuel management of light and heavy water moderated nuclear power reactors. Each computer program is described (provided that enough information was supplied) such that the nature of the physical problem solved and the basic mathematical or calculational approach are evident. In addition, further information regarding computer requirements, up-to-date applications and experiences and specific details concerning implementation, staff requirements, etc., are provided. Program procurement conditions, possible program implementation assistance and commercial conditions (where applicable) are given. (author)

A computational methodology for formulating gasoline surrogate fuels with accurate physical and chemical kinetic properties

KAUST Repository

Ahmed, Ahfaz; Goteng, Gokop; Shankar, Vijai; Al-Qurashi, Khalid; Roberts, William L.; Sarathy, Mani

2015-01-01

simpler molecular composition that represent real fuel behavior in one or more aspects are needed to enable repeatable experimental and computational combustion investigations. This study presents a novel computational methodology for formulating

FRAP-T1: a computer code for the transient analysis of oxide fuel rods

International Nuclear Information System (INIS)

Dearien, J.A.; Miller, R.L.; Hobbins, R.R.; Siefken, L.J.; Baston, V.F.; Coleman, D.R.

1977-02-01

FRAP-T is a FORTRAN IV computer code which can be used to solve for the transient response of a light water reactor (LWR) fuel rod during accident transients such as loss-of-coolant accident (LOCA) or a power-cooling-mismatch (PCM). The coupled effects of mechanical, thermal, internal gas, and material property response on the behavior of the fuel rod are considered. FRAP-T is a modular code with each major computational model isolated within the code and coupled to the main code by subroutine calls and data transfer through argument lists. FRAP-T is coupled to a materials properties subcode (MATPRO) which is used to provide gas, fuel, and cladding properties to the FRAP-T computational subcodes. No material properties need be supplied by the code user. The needed water properties are stored in tables built into the code. Critical heat flux (CHF) and heat transfer correlations for a wide range of coolant conditions are contained in modular subroutines. FRAP-T has been evaluated by making extensive comparisons between predictions of the code and experimental data. Comparison of predicted and experimental results are presented for a range of FRAP-T calculated parameters. The code is presently programmed and running on an IBM-360/75 and a CDC 7600 computer

Computational fluid dynamics application: slosh analysis of a fuel tank model

International Nuclear Information System (INIS)

Iu, H.S.; Cleghorn, W.L.; Mills, J.K.

2004-01-01

This paper presents the analysis of fluid slosh behaviour inside a fuel tank model. The fuel tank model was a simplified version of a stock fuel tank that has a sloshing noise problem. A commercial CFD software, FLOW-3D, was used to simulate the slosh behaviour. Slosh experiments were performed to verify the computer simulation results. High speed video equipment enhanced with a data acquisition system was used to record the slosh experiments and to obtain the instantaneous sound level of each video frame. Five baffle configurations including the no baffle configuration were considered in the computer simulations and the experiments. The simulation results showed that the best baffle configuration can reduce the mean kinetic energy by 80% from the no baffle configuration in a certain slosh situation. The experimental results showed that 15dB(A) noise reduction can be achieved by the best baffle configuration. The correlation analysis between the mean kinetic energy and the noise level showed that high mean kinetic energy of the fluid does not always correspond to high sloshing noise. High correlation between them only occurs for the slosh situations where the fluid hits the top of the tank and creates noise. (author)

COMETHE III J a computer code for predicting mechanical and thermal behaviour of a fuel pin

International Nuclear Information System (INIS)

Verbeek, P.; Hoppe, N.

1976-01-01

The design of fuel pins for power reactors requires a realistic evaluation of their thermal and mechanical performances throughout their irradiation life. This evaluation involves the knowledge of a number of parameters, very intricate and interconnected, for example, the temperature, the restructuring and the swelling rates of the fuel pellets, the dimensions, the stresses and the strains in the clad, the composition and the properties of gases, the inner gas pressure etc. This complex problem can only be properly handled by a computer programme which analyses the fuel pin thermal and mechanical behaviour at successive steps of its irradiation life. This report presents an overall description of the COMETHE III-J computer programme, designed to calculate the integral performance of oxide fuel pins with cylindrical metallic cladding irradiated in thermal or fast flux. (author)

A computational model of a PEM fuel cell with finite vapor absorption rate

Energy Technology Data Exchange (ETDEWEB)

Vorobev, A.; Zikanov, O.; Shamim, T. [Department of Mechanical Engineering, University of Michigan-Dearborn, 48128-1491 Dearborn, MI (United States)

2007-03-30

The paper presents a new computational model of non-steady operation of a PEM fuel cell. The model is based on the macroscopic hydrodynamic approach and assumptions of low humidity operation and one-dimensionality of transport processes. Its novelty and advantage in comparison with similar existing models is that it takes into account the finite-time equilibration between vapor and membrane-phase liquid water within the catalyst layers. The phenomenon is described using an additional parameter with the physical meaning of the typical reciprocal time of the equilibration. A computational parametric study is conducted to identify the effect of the finite-time equilibration on steady-state and transient operation of a PEM fuel cell. (author)

KWIKPLAN: a computer program for projecting the annual requirements of nuclear fuel cycle operations

International Nuclear Information System (INIS)

Salmon, R.; Kee, C.W.

1977-06-01

The computer code KWIKPLAN was written to facilitate the calculation of projected nuclear fuel cycle activities. Using given projections of power generation, the code calculates annual requirements for fuel fabrication, fuel reprocessing, uranium mining, and plutonium use and production. The code uses installed capacity projections and mass flow data for six types of reactors to calculate projected fuel cycle activities and inventories. It calculates fissile uranium and plutonium flows and inventories after allowing for an economy with limited reprocessing capacity and a backlog of unreprocessed fuel. All calculations are made on a quarterly basis; printed and punched output of the projected fuel cycle activities are made on an annual basis. Since the punched information is used in another code to determine waste inventories, the code punches a table from which the effective average burnup can be calculated for the fuel being reprocessed

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)

RAP-3A Computer code for thermal and hydraulic calculations in steady state conditions for fuel element clusters

International Nuclear Information System (INIS)

Popescu, C.; Biro, L.; Iftode, I.; Turcu, I.

1975-10-01

The RAP-3A computer code is designed for calculating the main steady state thermo-hydraulic parameters of multirod fuel clusters with liquid metal cooling. The programme provides a double accuracy computation of temperatures and axial enthalpy distributions of pressure losses and axial heat flux distributions in fuel clusters before boiling conditions occur. Physical and mathematical models as well as a sample problem are presented. The code is written in FORTRAN-4 language and is running on a IBM-370/135 computer

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.)

KUEBEL. A Fortran program for computation of cooling-agent-distribution within reactor fuel-elements

International Nuclear Information System (INIS)

Inhoven, H.

1984-12-01

KUEBEL is a Fortran-program for computation of cooling-agent-distribution within reactor fuel-elements or -zones of theirs. They may be assembled of max. 40 cooling-channels with laminar up to turbulent type of flow (respecting Reynolds' coefficients up to 2.0E+06) at equal pressure loss. Flow-velocity, dynamic flow-, contraction- and friction-losses will be calculated for each channel and for the total zone. Other computations will present mean heat-up of cooling-agent, mean outlet-temperature of the core, boiling-temperature and absolute pressure at flow-outlet. All characteristic coolant-values, including the factor of safety for flow-instability of the most-loaded cooling gap are computed by 'KUEBEL' too. Absolute pressure at flow-outlet or is-factor may be defined as dependent or independent variables of the program alternatively. In latter case 3 variations of solution will be available: Adapted flow of cooling-agent, inlet-temperature of the core and thermal power. All calculations can be done alternatively with variation of parameters: flow of cooling-agent, inlet-temperature of the core and thermal power, which are managed by the program itself. 'KUEBEL' is able to distinguish light- and heavy-water coolant, flow-direction of coolant and fuel elements with parallel, rectangular, respectively concentric, cylindrical shape of their gaps. Required material specifics are generated by the program. Segments of fuel elements or constructively unconnected gaps can also be computed by means of interposition of S.C. 'phantom channels'. (orig.) [de

Analysis of the Behavior of CAREM-25 Fuel Rods Using Computer Code BACO

International Nuclear Information System (INIS)

Estevez, Esteban; Markiewicz, Mario; Marino, Armando

2000-01-01

The thermo-mechanical behavior of a fuel rod subjected to irradiation is a complex process, on which a great quantity of interrelated physical-chemical phenomena are coupled.The code BACO simulates the thermo-mechanical behavior and the evolution of fission gases of a cylindrical rod in operation.The power history of fuel rods, arising from neutronic calculations, is the program input.The code calculates, among others, the temperature distribution and the principal stresses in the pellet and cladding, changes in the porosity and restructuring of pellet, the fission gases release, evolution of the internal gas pressure.In this work some of design limits of CAREM-25's fuel rods are analyzed by means of the computer code BACO.The main variables directly related with the integrity of the fuel rod are: Maximum temperature of pellet; Cladding hoop stresses; Gases pressure in the fuel rod; Cladding axial and radial strains, etc.The analysis of results indicates that, under normal operation conditions, the maximum fuel pellet temperature, cladding stresses, pressure of gases at end of life, etc, are below the design limits considered for the fuel rod of CAREM-25 reactor

FRAPCON-3: A computer code for the calculation of steady-state, thermal-mechanical behavior of oxide fuel rods for high burnup

International Nuclear Information System (INIS)

Berna, G.A.; Beyer, G.A.; Davis, K.L.; Lanning, D.D.

1997-12-01

FRAPCON-3 is a FORTRAN IV computer code that calculates the steady-state response of light water reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (1) heat conduction through the fuel and cladding, (2) cladding elastic and plastic deformation, (3) fuel-cladding mechanical interaction, (4) fission gas release, (5) fuel rod internal gas pressure, (6) heat transfer between fuel and cladding, (7) cladding oxidation, and (8) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat-transfer correlations. The codes' integral predictions of mechanical behavior have not been assessed against a data base, e.g., cladding strain or failure data. Therefore, it is recommended that the code not be used for analyses of cladding stress or strain. FRAPCON-3 is programmed for use on both mainframe computers and UNIX-based workstations such as DEC 5000 or SUN Sparcstation 10. It is also programmed for personal computers with FORTRAN compiler software and at least 8 to 10 megabytes of random access memory (RAM). The FRAPCON-3 code is designed to generate initial conditions for transient fuel rod analysis by the FRAPTRAN computer code (formerly named FRAP-T6)

Particle Image Velocimetry and Computational Fluid Dynamics Analysis of Fuel Cell Manifold

DEFF Research Database (Denmark)

Lebæk, Jesper; Blazniak Andreasen, Marcin; Andresen, Henrik Assenholm

2010-01-01

The inlet effect on the manifold flow in a fuel cell stack was investigated by means of numerical methods (computational fluid dynamics) and experimental methods (particle image velocimetry). At a simulated high current density situation the flow field was mapped on a 70 cell simulated cathode...

A computer model to predict temperatures and gas flows during AGR fuel handling

International Nuclear Information System (INIS)

Bishop, D.C.; Bowler, P.G.

1986-01-01

The paper describes the development of a comprehensive computer model (HOSTAGE) that has been developed for the Heysham II/Torness AGRs to predict temperature transients for all the important components during normal and fault conditions. It models not only the charge and discharge or fuel from an on-load reactor but also follows the fuel down the rest of the fuel route until it is dismantled. The main features of the physical model of gas and heat flow are described. Experimental results are used where appropriate and an indication will be given of how the predictions by HOSTAGE correlate with operating AGR reactors. The role of HOSTAGE in the Heysham II/Torness safety case is briefly discussed. (author)

FLATT - a computer programme for calculating flow and temperature transients in nuclear fuels

International Nuclear Information System (INIS)

Venkat Raj, V.; Koranne, S.M.

1976-01-01

FLATT is a computer code written in Fortran language for BESM-6 computer. The code calculates the flow transients in the coolant circuit of a nuclear reactor, caused by pump failure, and the consequent temperature transients in the fuel, clad, and the coolant. In addition any desired flow transient can be fed into the programme and the resulting temperature transients can be calculated. A case study is also presented. (author)

Application of Computational Intelligence Methods to In-Core Fuel Management

International Nuclear Information System (INIS)

Erdogan, A.

2001-01-01

In this study, a computer program package has been developed which supports the in-core fuel management activities for pressurized water reactors, generates and recommends an optimum loading pattern to ensure safe and efficient reactor operation. A search for an optimum fuel loading pattern must be conducted in the space of several core parameters such as power distribution, which is an excessively time consuming computational process. Global core calculation codes take a relatively long time to do the task. The time interval necessary for the iterative process was reduced by using an artificial neural network estimator for the calculations. In this way, it was possible to analyze more loading patterns in the same time interval and the probability of finding a desired optimum was increased. As a case study, the core of the Almaraz Nuclear Plant of Spain, a pressurized water reactor, was modeled for the core calculation code system. The 2-group cross sections for the fuel assembly types were calculated and stored for later usage with the diffusion code. 2000 loading patterns were generated by placing fuel assemblies to random positions in the core, and for each pattern the power distribution and effective multiplication factor (k e ff) were calculated with the diffusion code. At the next stage, 500 of the loading patterns were introduced to the neural network as input data for the training process. The remaining 1500 patterns were used to validate the neural network implementation. It was shown that the neural network estimates the power distribution and the K effective within acceptable error limits. To complete the system, a loading pattern generator was developed. This module consists of a set of rules and an algorithm that places the fuel assemblies to core positions. The neural network estimated the power distribution and k e ff for the loading patterns that were generated by this module. The patterns that have a maximum power fraction lower than, and a minimum

Computer simulation of fuel element performance

Energy Technology Data Exchange (ETDEWEB)

Sukhanov, G I

1979-01-01

The review presents reports made at the Conference on the Bahaviour and Production of Fuel for Water Reactors on March 13-17, 1979. Discussed at the Conference are the most developed and tested calculation models specially evolved to predict the behaviour of fuel elements of water reactors. The following five main aspects of the problem are discussed: general conceptions and programs; mechanical mock-ups and their applications; gas release, gap conductivity and fuel thermal conductivity; analysis of nonstationary processes; models of specific phenomena. The review briefly describes the physical principles of the following models and programs: the RESTR, providing calculation of the radii of zones of columnar and equiaxial grains as well as the radius of the internal cavity of the fuel core; programs for calculation of fuel-can interaction, based on the finite elements method; a model predicting the behaviour of the CANDU-PHW fuel elements in transient conditions. General results are presented of investigations of heat transfer through a can-fuel gap and thermal conductivity of UO/sub 2/ with regard for cracking and gas release of the fuel. Many programs already suit the accepted standards and are intensively tested at present.

The GC computer code for flow sheet simulation of pyrochemical processing of spent nuclear fuels

International Nuclear Information System (INIS)

Ahluwalia, R.K.; Geyer, H.K.

1996-01-01

The GC computer code has been developed for flow sheet simulation of pyrochemical processing of spent nuclear fuel. It utilizes a robust algorithm SLG for analyzing simultaneous chemical reactions between species distributed across many phases. Models have been developed for analysis of the oxide fuel reduction process, salt recovery by electrochemical decomposition of lithium oxide, uranium separation from the reduced fuel by electrorefining, and extraction of fission products into liquid cadmium. The versatility of GC is demonstrated by applying the code to a flow sheet of current interest

ROBOT3: a computer program to calculate the in-pile three-dimensional bowing of cylindrical fuel rods (AWBA Development Program)

International Nuclear Information System (INIS)

Kovscek, S.E.; Martin, S.E.

1982-10-01

ROBOT3 is a FORTRAN computer program which is used in conjunction with the CYGRO5 computer program to calculate the time-dependent inelastic bowing of a fuel rod using an incremental finite element method. The fuel rod is modeled as a viscoelastic beam whose material properties are derived as perturbations of the CYGRO5 axisymmetric model. Fuel rod supports are modeled as displacement, force, or spring-type nodal boundary conditions. The program input is described and a sample problem is given

The computer program ELCOM in the planning and structural analysis of PWR fuel elements: an example

International Nuclear Information System (INIS)

Silva Macedo, L.V. da

1990-01-01

Is's presented some results obtained with the ELCOM computer code, such as deflections, moments and natural frequencies, used in the design and structural analysis of PWR fuels assemblies. It's studied the behavior of these results varying the number of spacer grids, the rigidity of the joint between the fuel pin and the spacer grid, and the fuel assembly's boundary condition, considered in the analysis, in it's mounting into the core (if clamped-clamped, clamped-hinged or hinged-hinged). (author)

URANUS - a computer programme for the thermal and mechanical analysis of the fuel rods in a nuclear reactor

International Nuclear Information System (INIS)

Lassmann, K.

1978-01-01

The URANUS code, a digital computer programme for the thermal and mechanical analysis of integral fuel rods, is described. With this code the fuel rods found in the majority of power reactors can be analyzed. URANUS is built around a quasi two-dimensional analysis of fuel and cladding. The mechanical analysis can accommodate seven components of strain: elastic, time-independent plastic, creep and thermal strains, as well as strains due to swelling, cracking and densification. The heat generation and temperature distribution, cladding/fuel gap closure, pellet cracking and crack healing, fission-gas release, corrosion, O/M-distribution and plutonium redistribution are modelled. Geometric non-linearities (large displacements) are included; steady state or transient loading (pressure, temperature) is possible. In this paper special attention is paid to a theory for determining crack structures. The present status of the URANUS computer programme and a critical comparison with other fuel rod codes as well as sample analyses are given. (Auth.)

Citham a computer code for calculating fuel depletion-description, tests, modifications and evaluation

International Nuclear Information System (INIS)

Alvarenga, M.A.B.

1984-12-01

The CITHAN computer code was developed at IPEN (Instituto de Pesquisas Energeticas e Nucleares) to link the HAMMER computer code with a fuel depletion routine and to provide neutron cross sections to be read with the appropriate format of the CITATION code. The problem arised due to the efforts to addapt the new version denomined HAMMER-TECHION with the routine refered. The HAMMER-TECHION computer code was elaborated by Haifa Institute, Israel within a project with EPRI. This version is at CNEN to be used in multigroup constant generation for neutron diffusion calculation in the scope of the new methodology to be adopted by CNEN. The theoretical formulation of CITHAM computer code, tests and modificatins are described. (Author) [pt

Nuclear-fuel-cycle education: Module 5. In-core fuel management

International Nuclear Information System (INIS)

Levine, S.H.

1980-07-01

The purpose of this project was to develop a series of educational modules for use in nuclear-fuel-cycle education. These modules are designed for use in a traditional classroom setting by lectures or in a self-paced, personalized system of instruction. This module on in-core fuel management contains information on computational methods and theory; in-core fuel management using the Virginia Polytechnic Institute and State University computer modules; pressurized water reactor in-core fuel management; boiling water reactor in-core fuel management; and in-core fuel management for gas-cooled and fast reactors

Development of a computational model applied to a unitary 144 CM2 proton exchange membrane fuel cell

International Nuclear Information System (INIS)

Robalinho, Eric

2009-01-01

This work presents the development of a numerical computer model and methodology to study and design polymeric exchange membrane fuel cell - PEM. For the validation of experimental results, a sequence of routines, appropriate to fit the data obtained in the laboratory, was described. At the computational implementation it was created a new strategy of coupling two 3-dimensional models to satisfy the requirements of the comprehensive model of the fuel cell, including its various geometries and materials, as well as the various physical and chemical processes simulated. To effective assessment of the real cell analogy with numerical model, numerical studies were carried out. Comparisons with values obtained in the literature, characterization of variables through laboratory experiments and estimates from models already tested in the literature were also performed. Regarding the experimental part, a prototype of a fuel cell unit of 144 cm 2 of geometric area was designed, produced and operated at laboratory with the purpose of validating the numerical computer model proposed, with positive results. The results of simulations for the 2D and 3D geometries proposed are presented in the form of polarization curves, highlighting the catalytic layer model based on the geometry of agglomerates. Parametric and sensitivity studies are presented to illustrate the change in performance of the fuel cell studied. The final model is robust and useful as a tool for design and optimization of PEM type fuel cells in a wide range of operating conditions. (author)

Modification in the FUDA computer code to predict fuel performance at high burnup

Energy Technology Data Exchange (ETDEWEB)

Das, M; Arunakumar, B V; Prasad, P N [Nuclear Power Corp., Mumbai (India)

1997-08-01

The computer code FUDA (FUel Design Analysis) participated in the blind exercises organized by the IAEA CRP (Co-ordinated Research Programme) on FUMEX (Fuel Modelling at Extended Burnup). While the code prediction compared well with the experiments at Halden under various parametric and operating conditions, the fission gas release and fission gas pressure were found to be slightly over-predicted, particularly at high burnups. In view of the results of 6 FUMEX cases, the main models and submodels of the code were reviewed and necessary improvements were made. The new version of the code FUDA MOD 2 is now able to predict fuel performance parameter for burn-ups up to 50000 MWD/TeU. The validation field of the code has been extended to prediction of thorium oxide fuel performance. An analysis of local deformations at pellet interfaces and near the end caps is carried out considering the hourglassing of the pellet by finite element technique. (author). 15 refs, 1 fig.

Modification in the FUDA computer code to predict fuel performance at high burnup

International Nuclear Information System (INIS)

Das, M.; Arunakumar, B.V.; Prasad, P.N.

1997-01-01

The computer code FUDA (FUel Design Analysis) participated in the blind exercises organized by the IAEA CRP (Co-ordinated Research Programme) on FUMEX (Fuel Modelling at Extended Burnup). While the code prediction compared well with the experiments at Halden under various parametric and operating conditions, the fission gas release and fission gas pressure were found to be slightly over-predicted, particularly at high burnups. In view of the results of 6 FUMEX cases, the main models and submodels of the code were reviewed and necessary improvements were made. The new version of the code FUDA MOD 2 is now able to predict fuel performance parameter for burn-ups up to 50000 MWD/TeU. The validation field of the code has been extended to prediction of thorium oxide fuel performance. An analysis of local deformations at pellet interfaces and near the end caps is carried out considering the hourglassing of the pellet by finite element technique. (author). 15 refs, 1 fig

COXPRO-II: a computer program for calculating radiation and conduction heat transfer in irradiated fuel assemblies

International Nuclear Information System (INIS)

Rhodes, C.A.

1984-12-01

This report describes the computer program COXPRO-II, which was written for performing thermal analyses of irradiated fuel assemblies in a gaseous environment with no forced cooling. The heat transfer modes within the fuel pin bundle are radiation exchange among fuel pin surfaces and conduction by the stagnant gas. The array of parallel cylindrical fuel pins may be enclosed by a metal wrapper or shroud. Heat is dissipated from the outer surface of the fuel pin assembly by radiation and convection. Both equilateral triangle and square fuel pin arrays can be analyzed. Steady-state and unsteady-state conditions are included. Temperatures predicted by the COXPRO-II code have been validated by comparing them with experimental measurements. Temperature predictions compare favorably to temperature measurements in pressurized water reactor (PWR) and liquid-metal fast breeder reactor (LMFBR) simulated, electrically heated fuel assemblies. Also, temperature comparisons are made on an actual irradiated Fast-Flux Test Facility (FFTF) LMFBR fuel assembly

FUDA MOD-2: a computer program for simulation the performance of fuel element validation exercise

International Nuclear Information System (INIS)

Chouhan, S.K.; Tripathi, R.M.; Prasad, P.N.; Chauhan, Ashok

2014-01-01

The PHWR fuel element performance is evaluated using the fuel analysis computer code FUDA MOD2. It is specifically written for performance simulation of UO 2 fuel pellet, located in zirconium alloy sheath operating under coolant pressure. For specific element power histories, the code investigates the variables and their interactions that govern fuel element performance. The input data requires pellet dimensions, element dimensions, sheath properties, heat transfer data, thermal hydraulic parameters of coolant, the inner filler gas composition, flux gradient and linear heat ratings (LHR) at different burn up. The output data generated by the code are radial temperature profile of fuel and sheath, fuel sheath-gap heat transfer coefficient, fission gas generated and released, fission gas pressure, sheath stress and strain for different burn-up zones. The code has been verified against literature data and post irradiation examinations carried out. It is also bench marked against various international fuel element simulation programmes available with water cooled reactors operating countries. The present paper describes the FUDA MOD2 code verification studies carried out using the literature data and post irradiation examination data. (author)

Nuclear fuel cycle risk assessment: survey and computer compilation of risk-related literature. [Once-through Cycle and Plutonium Recycle

Energy Technology Data Exchange (ETDEWEB)

Yates, K.R.; Schreiber, A.M.; Rudolph, A.W.

1982-10-01

The US Nuclear Regulatory Commission has initiated the Fuel Cycle Risk Assessment Program to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. Both the once-through cycle and plutonium recycle are being considered. A previous report generated by this program defines and describes fuel cycle facilities, or elements, considered in the program. This report, the second from the program, describes the survey and computer compilation of fuel cycle risk-related literature. Sources of available information on the design, safety, and risk associated with the defined set of fuel cycle elements were searched and documents obtained were catalogued and characterized with respect to fuel cycle elements and specific risk/safety information. Both US and foreign surveys were conducted. Battelle's computer-based BASIS information management system was used to facilitate the establishment of the literature compilation. A complete listing of the literature compilation and several useful indexes are included. Future updates of the literature compilation will be published periodically. 760 annotated citations are included.

Analysis of the behavior under irradiation of high burnup nuclear fuels with the computer programs FRAPCON and FRAPTRAN

Energy Technology Data Exchange (ETDEWEB)

Reis, Regis; Silva, Antonio Teixeira e, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2017-11-01

The objective of this paper is to verify the validity and accuracy of the results provided by computer programs FRAPCON-3.4a and FRAPTRAN-1.4, used in the simulation process of the irradiation behavior of Pressurized Water Reactors (PWR) fuel rods, in steady-state and transient operational conditions at high burnup. To achieve this goal, the results provided by these computer simulations are compared with experimental data available in the database FUMEX III. Through the results, it was found that the computer programs used have a good ability to predict the operational behavior of PWR fuel rods in high burnup steady-state conditions and under the influence of Reactivity Initiated Accident (RIA). (author)

Manual of phosphoric acid fuel cell power plant cost model and computer program

Science.gov (United States)

Lu, C. Y.; Alkasab, K. A.

1984-01-01

Cost analysis of phosphoric acid fuel cell power plant includes two parts: a method for estimation of system capital costs, and an economic analysis which determines the levelized annual cost of operating the system used in the capital cost estimation. A FORTRAN computer has been developed for this cost analysis.

A computational technique to identify the optimal stiffness matrix for a discrete nuclear fuel assembly model

International Nuclear Information System (INIS)

Park, Nam-Gyu; Kim, Kyoung-Joo; Kim, Kyoung-Hong; Suh, Jung-Min

2013-01-01

Highlights: ► An identification method of the optimal stiffness matrix for a fuel assembly structure is discussed. ► The least squares optimization method is introduced, and a closed form solution of the problem is derived. ► The method can be expanded to the system with the limited number of modes. ► Identification error due to the perturbed mode shape matrix is analyzed. ► Verification examples show that the proposed procedure leads to a reliable solution. -- Abstract: A reactor core structural model which is used to evaluate the structural integrity of the core contains nuclear fuel assembly models. Since the reactor core consists of many nuclear fuel assemblies, the use of a refined fuel assembly model leads to a considerable amount of computing time for performing nonlinear analyses such as the prediction of seismic induced vibration behaviors. The computational time could be reduced by replacing the detailed fuel assembly model with a simplified model that has fewer degrees of freedom, but the dynamic characteristics of the detailed model must be maintained in the simplified model. Such a model based on an optimal design method is proposed in this paper. That is, when a mass matrix and a mode shape matrix are given, the optimal stiffness matrix of a discrete fuel assembly model can be estimated by applying the least squares minimization method. The verification of the method is completed by comparing test results and simulation results. This paper shows that the simplified model's dynamic behaviors are quite similar to experimental results and that the suggested method is suitable for identifying reliable mathematical model for fuel assemblies

IAMBUS, a computer code for the design and performance prediction of fast breeder fuel rods

International Nuclear Information System (INIS)

Toebbe, H.

1990-05-01

IAMBUS is a computer code for the thermal and mechanical design, in-pile performance prediction and post-irradiation analysis of fast breeder fuel rods. The code deals with steady, non-steady and transient operating conditions and enables to predict in-pile behavior of fuel rods in power reactors as well as in experimental rigs. Great effort went into the development of a realistic account of non-steady fuel rod operating conditions. The main emphasis is placed on characterizing the mechanical interaction taking place between the cladding tube and the fuel as a result of contact pressure and friction forces, with due consideration of axial and radial crack configuration within the fuel as well as the gradual transition at the elastic/plastic interface in respect to fuel behavior. IAMBUS can be readily adapted to various fuel and cladding materials. The specific models and material correlations of the reference version deal with the actual in-pile behavior and physical properties of the KNK II and SNR 300 related fuel rod design, confirmed by comparison of the fuel performance model with post-irradiation data. The comparison comprises steady, non-steady and transient irradiation experiments within the German/Belgian fuel rod irradiation program. The code is further validated by comparison of model predictions with post-irradiation data of standard fuel and breeder rods of Phenix and PFR as well as selected LWR fuel rods in non-steady operating conditions

Specialists' meeting on fuel element performance computer modelling, Preston, United Kingdom, 15-19 March 1982

International Nuclear Information System (INIS)

1983-03-01

The 46 papers of the meeting concerned with computer models of Water Reactor fuel elements cover practically all aspects of behavior of fuel elements in normal operation and in accident condition. Each session of the meeting produced a critical evaluation of one of the 5 topics into which the subject area had been divided. The sessions' report summarize the papers and make recommendations for further work. Separate abstracts were prepared for all the papers presented at this meeting

SIVAR - Computer code for simulation of fuel rod behavior in PWR during fast transients

International Nuclear Information System (INIS)

Dias, A.F.V.

1980-10-01

Fuel rod behavior during a stationary and a transitory operation, is studied. A computer code aiming at simulating PWR type rods, was developed; however, it can be adapted for simulating other type of rods. A finite difference method was used. (E.G.) [pt

Computer program for modelling the history of the in-service bending of fast power reactor fuel assemblies

International Nuclear Information System (INIS)

Dienstbier, J.

1979-04-01

The studies into stresses and deformations in the core are mainly focused on the fuel rod and the fuel assembly can. In high neutron doses austenitic steel swells and this is associated with a considerable increase in the volume of material. The SANDRA computer program is used for solving the problems of can deformations and stress during long-term reactor operation. The block for the mechanical interaction of cans is the key part of the program. The program input data include temperature distribution, fast neutron flux distribution and coolant overpressure inside the cans. Reactor operation is modelled using operating modes A, B, C which may arbitrarily be combined. Mode A computes bending deformations and the deformations of the can cross-section due to temperature dilatation in the change in temperature fields in the reactor; mode B computes deformations due to swelling and creep in long-term operation; mode C computes thermal deformations in reactor shut-down. A flowsheet is shown of program SANDRA as are examples of computed deformations. (M.S.)

Computational analysis of the behaviour of nuclear fuel under steady state, transient and accident conditions

International Nuclear Information System (INIS)

2007-12-01

Accident analysis is an important tool for ensuring the adequacy and efficiency of the provision in the defence in depth concept to cope with challenges to plant safety. Accident analysis is the milestone of the demonstration that the plant is capable of meeting any prescribed limits for radioactive releases and any other acceptable limits for the safe operation of the plant. It is used, by designers, utilities and regulators, in a number of applications such as: (a) licensing of new plants, (b) modification of existing plants, (c) analysis of operational events, (d) development, improvement or justification of the plant operational limits and conditions, and (e) safety cases. According to the defence in depth concept, the fuel rod cladding constitutes the first containment barrier of the fission products. Therefore, related safety objectives and associated criteria are defined, in order to ensure, at least for normal operation and anticipated transients, the integrity of the cladding, and for accident conditions, acceptable radiological consequences with regard to the postulated frequency of the accident, as usually identified in the safety analysis reports. Therefore, computational analysis of fuel behaviour under steady state, transient and accident conditions constitutes a major link of the safety case in order to justify the design and the safety of the fuel assemblies, as far as all relevant phenomena are correctly addressed and modelled. This publication complements the IAEA Safety Report on Accident Analysis for Nuclear Power Plants (Safety Report Series No. 23) that provides practical guidance for establishing a set of conceptual and formal methods and practices for performing accident analysis. Computational analysis of the behaviour of nuclear fuel under transient and accident conditions, including normal operation (e.g. power ramp rates) is developed in this publication. For design basis accidents, depending on the type of influence on a fuel element

Computational Investigation of On-Line Interrogation of Pebble Bed Reactor Fuel

Science.gov (United States)

Hawari, A. I.; Chen, Jianwei

2005-10-01

Pebble bed reactors are characterized by multipass fuel systems in which spherical fuel pebbles are circulated through the core until they reach a proposed burnup limit (80000-100000 MWD/MTU). For such reactors, the fuel is assayed on-line to ensure that the burnup limit is not breached. We considered assaying the fuel using an HPGe detector to perform passive gamma-ray spectrometry of fission products. Since neither fresh nor irradiated fuel is readily available, computer simulations were utilized to identify the radionuclides that can be used as burnup indicators, and to visualize the gamma-ray spectra at various levels of burnup. Specifically, we used the ORIGEN-MONTEBURNS-MCNP code system. This allowed the establishment of the burnup dependent one-group gas reactor cross-sections for the radionuclides of interest. Subsequently, ORIGEN was used to simulate in-core pebble depletion to establish the irradiated pebble isotopics. Finally, the codes MCNP and SYNTH were used to simulate the response of the HPGe gamma-ray spectrometer. The results show that absolute and relative indicators can be used on-line to determine unambiguously the enrichment and burnup on a pebble-by-pebble basis. The activity of Cs-137 or the activity ratio of Co-60/Cs-134 can be combined with the activity ratio of Np-239/I-132 to yield the enrichment and burnup information. To use the relative indicators, a relative efficiency calibration of the gamma-ray spectrometer can be performed using the La-140 gamma lines that are emitted by the irradiated pebble. I-132, Cs-134, Cs-137, La-140, and Np-239 are produced upon the irradiation of the fuel. Co-60 is produced by doping the fuel with a small amount (/spl sim/100 ppm) of Co-59. Using this approach, the uncertainty in burnup determination due to factors such as power history variation, detector efficiency calibration, and counting statistics is expected to remain in the range of /spl plusmn/5% to /spl plusmn/10%.

MEASUREMENTS AND COMPUTATIONS OF FUEL DROPLET TRANSPORT IN TURBULENT FLOWS

Energy Technology Data Exchange (ETDEWEB)

Joseph Katz and Omar Knio

2007-01-10

The objective of this project is to study the dynamics of fuel droplets in turbulent water flows. The results are essential for development of models capable of predicting the dispersion of slightly light/heavy droplets in isotropic turbulence. Since we presently do not have any experimental data on turbulent diffusion of droplets, existing mixing models have no physical foundations. Such fundamental knowledge is essential for understanding/modeling the environmental problems associated with water-fuel mixing, and/or industrial processes involving mixing of immiscible fluids. The project has had experimental and numerical components: 1. The experimental part of the project has had two components. The first involves measurements of the lift and drag forces acting on a droplet being entrained by a vortex. The experiments and data analysis associated with this phase are still in progress, and the facility, constructed specifically for this project is described in Section 3. In the second and main part, measurements of fuel droplet dispersion rates have been performed in a special facility with controlled isotropic turbulence. As discussed in detail in Section 2, quantifying and modeling the of droplet dispersion rate requires measurements of their three dimensional trajectories in turbulent flows. To obtain the required data, we have introduced a new technique - high-speed, digital Holographic Particle Image Velocimetry (HPIV). The technique, experimental setup and results are presented in Section 2. Further information is available in Gopalan et al. (2005, 2006). 2. The objectives of the numerical part are: (1) to develop a computational code that combines DNS of isotropic turbulence with Lagrangian tracking of particles based on integration of a dynamical equation of motion that accounts for pressure, added mass, lift and drag forces, (2) to perform extensive computations of both buoyant (bubbles) and slightly buoyant (droplets) particles in turbulence conditions

Personal computer based decision support system for routing nuclear spent fuel

International Nuclear Information System (INIS)

Chin, Shih-Miao; Joy, D.S.; Johnson, P.E.; Bobic, S.M.; Miaou, Shaw-Pin

1989-01-01

An approach has been formulated to route nuclear spent fuel over the US Interstate highway network. This approach involves the generation of alternative routes so that any potential adverse impacts will not only concentrate on regions along the shortest path between the nuclear power plant and repository. Extensive literature research on the shortest path finding algorithms has been carried out. Consequently, an extremely efficient shortest path algorithm has been implemented and significantly increases the overall system performance. State-of-the-art interactive computer graphics is used. In addition to easy-to-use pop-up menus, full color mapping and display capabilities are also incorporated. All of these features have been implemented on commonly available personal computers. 6 figs., 2 tabs

Application of a personal computer relational data base management system to fuel cycle economic scoping

International Nuclear Information System (INIS)

Malone, J.P.; Dooley, G.D.

1986-01-01

A personal computer (PC) relational data base management system (RDBMS) permits large quantities of data to be maintained in a data base composed of structured data sets or files and provides data access through a software environment, procedure, or program language. The features of an RDBMS-based system create an environment on a PC that can provide significant benefits to any fuel cycle economics analysis. The ability to maintain a separate data set for each fuel cycle parameter group and the ability to manipulate the data through a series of independent calculation modules combine to provide the fuel cycle analyst with more time to examine and use the data, because less time is required to manipulate it

Computer code for the analysis of destructive pressure generation process during a fuel failure accident, PULSE-2

International Nuclear Information System (INIS)

Fujishiro, Toshio

1978-03-01

The computer code PULSE-2 has been developed for the analysis of pressure pulse generation process when hot fuel particles come into contact with the coolant in a fuel rod failure accident. In the program, it is assumed that hot fuel fragments mix with the coolant instantly and homogeneously in the failure region. Then, the rapid vaporization of the coolant and transient pressure rise in failure region, and the movement of ejected coolant slugs are calculated. The effect of a fuel-particle size distribution is taken into consideration. Heat conduction in the fuel particles and heat transfer at fuel-coolant interface are calculated. Temperature, pressure and void fraction in the mixed region are calculated from the average enthalpy. With physical property subroutines for liquid sodium and water, the model is usable for both LMFBR and LWR conditions. (auth.)

RODSWELL: a computer code for the thermomechanical analysis of fuel rods under LOCA conditions

International Nuclear Information System (INIS)

Casadei, F.; Laval, H.; Donea, J.; Jones, P.M.; Colombo, A.

1984-01-01

The present report is the user's manual for the computer code RODSWELL developed at the JRC-Ispra for the thermomechanical analysis of LWR fuel rods under simulated loss-of-coolant accident (LOCA) conditions. The code calculates the variation in space and time of all significant fuel rod variables, including fuel, gap and cladding temperature, fuel and cladding deformation, cladding oxidation and rod internal pressure. The essential characteristics of the code are briefly outlined here. The model is particularly designed to perform a full thermal and mechanical analysis in both the azimuthal and radial directions. Thus, azimuthal temperature gradients arising from pellet eccentricity, flux tilt, arbitrary distribution of heat sources in the fuel and the cladding and azimuthal variation of coolant conditions can be treated. The code combines a transient 2-dimensional heat conduction code and a 1-dimentional mechanical model for the cladding deformation. The fuel rod is divided into a number of axial sections and a detailed thermomechanical analysis is performed within each section in radial and azimuthal directions. In the following sections, instructions are given for the definition of the data files and the semi-variable dimensions. Then follows a complete description of the input data. Finally, the restart option is described

TRUMP-BD: A computer code for the analysis of nuclear fuel assemblies under severe accident conditions

International Nuclear Information System (INIS)

Lombardo, N.J.; Marseille, T.J.; White, M.D.; Lowery, P.S.

1990-06-01

TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic in form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000 degree F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion (''bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled

TRUMP-BD: A computer code for the analysis of nuclear fuel assemblies under severe accident conditions

Energy Technology Data Exchange (ETDEWEB)

Lombardo, N.J.; Marseille, T.J.; White, M.D.; Lowery, P.S.

1990-06-01

TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic in form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000{degree}F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion ( bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled.

Computational Flame Diagnostics for Direct Numerical Simulations with Detailed Chemistry of Transportation Fuels

Energy Technology Data Exchange (ETDEWEB)

Lu, Tianfeng [Univ. of Connecticut, Storrs, CT (United States)

2017-02-16

The goal of the proposed research is to create computational flame diagnostics (CFLD) that are rigorous numerical algorithms for systematic detection of critical flame features, such as ignition, extinction, and premixed and non-premixed flamelets, and to understand the underlying physicochemical processes controlling limit flame phenomena, flame stabilization, turbulence-chemistry interactions and pollutant emissions etc. The goal has been accomplished through an integrated effort on mechanism reduction, direct numerical simulations (DNS) of flames at engine conditions and a variety of turbulent flames with transport fuels, computational diagnostics, turbulence modeling, and DNS data mining and data reduction. The computational diagnostics are primarily based on the chemical explosive mode analysis (CEMA) and a recently developed bifurcation analysis using datasets from first-principle simulations of 0-D reactors, 1-D laminar flames, and 2-D and 3-D DNS (collaboration with J.H. Chen and S. Som at Argonne, and C.S. Yoo at UNIST). Non-stiff reduced mechanisms for transportation fuels amenable for 3-D DNS are developed through graph-based methods and timescale analysis. The flame structures, stabilization mechanisms, local ignition and extinction etc., and the rate controlling chemical processes are unambiguously identified through CFLD. CEMA is further employed to segment complex turbulent flames based on the critical flame features, such as premixed reaction fronts, and to enable zone-adaptive turbulent combustion modeling.

A Computational and Experimental Study of Coflow Laminar Methane/Air Diffusion Flames: Effects of Fuel Dilution, Inlet Velocity, and Gravity

Science.gov (United States)

Cao, S.; Ma, B.; Bennett, B. A. V.; Giassi, D.; Stocker, D. P.; Takahashi, F.; Long, M. B.; Smooke, M. D.

2014-01-01

The influences of fuel dilution, inlet velocity, and gravity on the shape and structure of laminar coflow CH4-air diffusion flames were investigated computationally and experimentally. A series of nitrogen-diluted flames measured in the Structure and Liftoff in Combustion Experiment (SLICE) on board the International Space Station was assessed numerically under microgravity (mu g) and normal gravity (1g) conditions with CH4 mole fraction ranging from 0.4 to 1.0 and average inlet velocity ranging from 23 to 90 cm/s. Computationally, the MC-Smooth vorticity-velocity formulation was employed to describe the reactive gaseous mixture, and soot evolution was modeled by sectional aerosol equations. The governing equations and boundary conditions were discretized on a two-dimensional computational domain by finite differences, and the resulting set of fully coupled, strongly nonlinear equations was solved simultaneously at all points using a damped, modified Newton's method. Experimentally, flame shape and soot temperature were determined by flame emission images recorded by a digital color camera. Very good agreement between computation and measurement was obtained, and the conclusions were as follows. (1) Buoyant and nonbuoyant luminous flame lengths are proportional to the mass flow rate of the fuel mixture; computed and measured nonbuoyant flames are noticeably longer than their 1g counterparts; the effect of fuel dilution on flame shape (i.e., flame length and flame radius) is negligible when the flame shape is normalized by the methane flow rate. (2) Buoyancy-induced reduction of the flame radius through radially inward convection near the flame front is demonstrated. (3) Buoyant and nonbuoyant flame structure is mainly controlled by the fuel mass flow rate, and the effects from fuel dilution and inlet velocity are secondary.

RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Yuan, Y.C. [Square Y, Orchard Park, NY (United States); Chen, S.Y.; LePoire, D.J. [Argonne National Lab., IL (United States). Environmental Assessment and Information Sciences Div.; Rothman, R. [USDOE Idaho Field Office, Idaho Falls, ID (United States)

1993-02-01

This report presents the technical details of RISIUND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, semiinteractive program that can be run on an IBM or equivalent personal computer. The program language is FORTRAN-77. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incidentfree models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionudide inventory and dose conversion factors.

RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

International Nuclear Information System (INIS)

Yuan, Y.C.; Chen, S.Y.; LePoire, D.J.

1993-02-01

This report presents the technical details of RISIUND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, semiinteractive program that can be run on an IBM or equivalent personal computer. The program language is FORTRAN-77. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incidentfree models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionudide inventory and dose conversion factors

Computational Thermodynamics Analysis of Vaporizing Fuel Droplets in the Human Upper Airways

Science.gov (United States)

Zhang, Zhe; Kleinstreuer, Clement

The detailed knowledge of air flow structures as well as particle transport and deposition in the human lung for typical inhalation flow rates is an important precursor for dosimetry-and-health-effect studies of toxic particles as well as for targeted drug delivery of therapeutic aerosols. Focusing on highly toxic JP-8 fuel aerosols, 3-D airflow and fluid-particle thermodynamics in a human upper airway model starting from mouth to Generation G3 (G0 is the trachea) are simulated using a user-enhanced and experimentally validated finite-volume code. The temperature distributions and their effects on airflow structures, fuel vapor deposition and droplet motion/evaporation are discussed. The computational results show that the thermal effect on vapor deposition is minor, but it may greatly affect droplet deposition in human airways.

GAPCON-THERMAL-2: a computer program for calculating the thermal behavior of an oxide fuel rod

International Nuclear Information System (INIS)

Beyer, C.E.; Hann, C.R.; Lanning, D.D.; Panisko, F.E.; Parchen, L.J.

1975-11-01

A description is presented of the computer code GAPCON THERMAL-2, a light water reactor (LWR) fuel thermal performance prediction code. GAPCON-THERMAL-2, is intended to be used as a calculational tool for reactor fuel steady-state thermal performance and to provide input for accident analyses. Some models used in the code provide best estimate as well as conservative predictions. Each of the individual models in the code is based on the best available data

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)

Investigation on structural analysis computer program of spent nuclear fuel shipping cask

International Nuclear Information System (INIS)

Yagawa, Ganki; Ikushima, Takeshi.

1987-10-01

This report describes the results done by the Sub-Committee of Research Cooperation Committee (RC-62) of the Japan Society of Mechanical Engineers under the trust of the Japan Atomic Energy Research Institute. The principal fulfilments and accomplishments are summarized as follows: (1) Regarding the survey of structural analysis methods of spent fuel shipping cask, several documents, which explain the features and applications of the exclusive computer programs for impact analysis on the basis of 2 or 3 dimensional finite element or difference methods such as HONDO, STEALTH and DYNA-3D, were reviewed. (2) In comparative evaluation of the existing computer programs, the common benchmark test problems for 9 m vertical drop impact of the axisymmetric lead cylinder with and without stainless steel clads were adopted where the calculational evaluations for taking into account the strain rate effect were carried out. (3) Evaluation of impact analysis algorithm of computer programs were conducted and the requirements for computer programs to be developed in future and an index for further studies have been clarified. (author)

Dynamics of nuclear fuel assemblies in vertical flow channels: computer modelling and associated studies

International Nuclear Information System (INIS)

Mason, V.A.; Pettigrew, M.J.; Lelli, G.; Kates, L.; Reimer, E.

1978-10-01

A computer model, designed to predict the dynamic behaviour of nuclear fuel assemblies in axial flow, is described in this report. The numerical methods used to construct and solve the matrix equations of motion in the model are discussed together with an outline of the method used to interpret the fuel assembly stability data. The mathematics developed for forced response calculations are described in detail. Certain structural and hydrodynamic modelling parameters must be determined by experiment. These parameters are identified and the methods used for their evaluation are briefly described. Examples of typical applications of the dynamic model are presented towards the end of the report. (author)

PECITIS-II, a computer program to predict the performance of collapsible clad UO2 fuel elements

International Nuclear Information System (INIS)

Anand, A.K.; Anantharaman, K.; Sarda, V.

1978-01-01

The Indian power programme envisages the use of PHWRs, which use collapsible clad UO 2 fuel elements. A computer code, PECITIS-II, developed for the analysis of this type of fuel is described in detail. The sheath strain and fission gas pressure are evaluated by this method. The pellet clad gap conductance is calculated by Ross and Solute model. The pellet thermal expansion is calculated by assuming a two zone model, i.e. a plastic core surrounded by an elastic cracked annulus. (author)

Development of multimedia computer-based training for VXI integrated fuel monitors

International Nuclear Information System (INIS)

Keeffe, R.; Ellacott, T.; Truong, Q.S.

1999-01-01

The Canadian Safeguards Support Program has developed the VXI Integrated Fuel Monitor (VFIM) which is based on the international VXI instrument bus standard. This equipment is a generic radiation monitor which can be used in an integrated mode where several detection systems can be connected to a common system where information is collected, displayed, and analyzed via a virtual control panel with the aid of computers, trackball and computer monitor. The equipment can also be used in an autonomous mode as a portable radiation monitor with a very low power consumption. The equipment has been described at previous international symposia. Integration of several monitoring systems (bundle counter, core discharge monitor, and yes/no monitor) has been carried out at Wolsong 2. Performance results from one of the monitoring systems which was installed at CANDU nuclear stations are discussed in a companion paper at this symposium. This paper describes the development of an effective multimedia computer-based training package for the primary users of the equipment; namely IAEA inspectors and technicians. (author)

CASKETSS: a computer code system for thermal and structural analysis of nuclear fuel shipping casks

International Nuclear Information System (INIS)

Ikushima, Takeshi

1989-02-01

A computer program CASKETSS has been developed for the purpose of thermal and structural analysis of nuclear fuel shipping casks. CASKETSS measn a modular code system for CASK Evaluation code system Thermal and Structural Safety. Main features of CASKETSS are as follow; (1) Thermal and structural analysis computer programs for one-, two-, three-dimensional geometries are contained in the code system. (2) Some of the computer programs in the code system has been programmed to provide near optimal speed on vector processing computers. (3) Data libralies fro thermal and structural analysis are provided in the code system. (4) Input data generator is provided in the code system. (5) Graphic computer program is provided in the code system. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

Correction device for fuel positioning value

International Nuclear Information System (INIS)

Nakahara, Toshiro.

1993-01-01

In a computer for a fuel exchanger for controlling handling of fuels such as loading and unloading based on the data for settling values of fuels installed in a reactor core or fuel pool, data for burnup degree during reactor operation are inputted from a computer computing the reactor output into a correction device for fuel settling values in which fuel irradiation growth rate is calculated to determine a correction value. This makes it unnecessary for sampling measurement of fuel settling values in the reactor core practiced at the same time with the reactor opening and exchange operation for the stored date of settling values in the computer for the fuel exchanger conducted on every time. New data for settling values can be exchanged automatically based on the data for the burnup degree at the same time with the reactor shutdown, which can be conducted easily over the whole number. Accordingly, it is possible to improve reliability and safety of the fuel exchanging operation including the setting for the interlock of the fuel exchanging operation relevant to the fuel settling values, as well as moderate operator's burden. (N.H.)

Evaluation of pulmonary changes due to biomass fuels using high-resolution computed tomography

International Nuclear Information System (INIS)

Kara, Mustafa; Tas, Fikret; Bulut, Sema; Akkurt, Ibrahim; Seyfikli, Zehra

2003-01-01

Biomass fuels are frequently used in rural areas of the world for cooking and heating frequently. It has been reported that the use of these fuels causes hazardous effects on the lungs. In this study, we evaluated the pulmonary changes due to the use of biomass fuels in a female population that lives in our territory by high-resolution computed tomography (HRCT). The study analyzed three groups of women. The first group comprised those subjects who were exposed to biomass without respiratory symptoms (group 1; n=32). The second group comprised those individuals that were exposed to biomass and showed respiratory symptoms, such as cough, sputum production, and dyspnea (group 2; n=30). The third group was composed of women who were not exposed to biomass and also had no respiratory symptoms (group 3; n=30). Women with a history of concomitant pulmonary diseases were excluded from the study. All groups were examined with HRCT. Groups 1 and 2 (individuals exposed to biomass fuels) had more pathologic findings than group 3 (not exposed to biomass fuels). Ground-glass appearance was seen in 71.9% in group 1, 23.3% in group 2, and 3.3% in group 3. The difference between the groups was statistically significant (p<0.05). Fibrotic bands were seen 50% in group 1, 63.3% in group 2, and only 6.7% in group 3 (p<0.001). Exposure to biomass fuels was the cause or predisposing factor for many pulmonary diseases, ranging from chronic bronchitis to diffuse lung diseases. We believe that these pathological changes due to biomass fuels can be detected earlier by HRCT and the diseases might be prevented or treated earlier. (orig.)

Computational and experimental analysis of causes for local deformation of research reactor U-Mo fuel pin claddings in case of high burn-ups

International Nuclear Information System (INIS)

Popov, V.V.; Khmelevsky, M.Ya.; Lukichev, V.A.; Golosov, O.A.

2005-01-01

Post-reactor investigations of (U-Mo) fuel pins irradiated in the IVV-2M reactor have allowed to determine: the change in a fuel pin volume; the dimensions and the kind of the local deformation of fuel pin claddings; the amount of gases released under the cladding from the fuel composition, the thickness and appearance of the interaction layer of between the (U-Mo) particles and aluminium as a matrix material. The computational analysis of the stressed-strained state of fuel pins has shown that the major contribution to the increase of the fuel pin volume is made by the fuel swelling caused by the solid products of fission being formed in the process of operation. The emergence of the (U-Mo) fuel-aluminium matrix interaction layers around the (U-Mo) particles results in formation and evolution of lamination cavities inside the fuel composition under the joint action of the pressure of process gases and gaseous fission products. In case of high burn-up a local bulge of a fuel pin cladding is being formed in the fuel lamination area caused by the pressure of gases in the presence of creep in the fuel pin cladding material. The computational results relating to the local strain in a research reactor (U-Mo) fuel pin are in a good accordance with the results of the post-reactor investigations. (author)

Accuracy and computational time of a hierarchy of growth rate definitions for breeder reactor fuel

International Nuclear Information System (INIS)

Maudlin, P.J.; Borg, R.C.; Ott, K.O.

1979-01-01

For a hierarchy of four logically different definitions for calculating the asymptotic growth of fast breeder reactor fuel, an investigation is performed concerning the comparative accuracy and computational effort associated with each definition. The definition based on detailed calculation of the accumulating fuel in an expanding park of reactors asymptotically yields the most accurate value of the infinite time growth rate, γ/sup infinity/, which is used as a reference value. The computational effort involved with the park definition is very large. The definition based on the single reactor calculation of the equilibrium surplus production rate and fuel inventory gives a value for γ/sup infinity of comparable accuracy to the park definition and uses significantly less central processor unit (CPU) time. The third definition is based on a continuous treatment of the reactor fuel cycle for a single reactor and gives a value for γ/sup infinity/ that accurately approximates the second definition. The continuous definition requires very little CPU time. The fourth definition employs the isotopic breeding worths, w/sub i//sup */, for a projection of the asymptotic growth rate. The CPU time involved in this definition is practically nil if its calculation is based on the few-cycle depletion calculation normally performed for core design and critical enrichment evaluations. The small inaccuracy (approx. = 1%) of the breeding-worth-based definition is well within the inaccuracy range that results unavoidably from other sources such as nuclear cross sections, group constants, and flux calculations. This fully justifies the use of this approach in routine calculations

Fuel pattern recognition device

International Nuclear Information System (INIS)

Sato, Tomomi.

1995-01-01

The device of the present invention monitors normal fuel exchange upon fuel exchanging operation carried out in a reactor of a nuclear power plant. Namely, a fuel exchanger is movably disposed to the upper portion of the reactor and exchanges fuels. An exclusive computer receives operation signals of the fuel exchanger during operation as inputs, and outputs reactor core fuel pattern information signals to a fuel arrangement diagnosis device. An underwater television camera outputs image signals of a fuel pattern in the reactor core to an image processing device. If there is any change in the image signals for the fuel pattern as a result of the fuel exchange operation of the fuel exchanger, the image processing device outputs the change as image signals to the fuel pattern diagnosis device. The fuel pattern diagnosis device compares the pattern information signals from the exclusive computer with the image signals from the image processing device, to diagnose the result of the fuel exchange operation performed by the fuel exchanger and inform the diagnosis by means of an image display. (I.S.)

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

Thorium fuel cycle management

International Nuclear Information System (INIS)

Zajac, R.; Darilek, P.; Breza, J.; Necas, V.

2010-01-01

In this presentation author deals with the thorium fuel cycle management. Description of the thorium fuels and thorium fuel cycle benefits and challenges as well as thorium fuel calculations performed by the computer code HELIOS are presented.

Calculation of Plutonium content in RSG-GAS spent fuel using IAFUEL computer code

International Nuclear Information System (INIS)

Mochamad-Imron

2003-01-01

It has been calculated the contain of isotopes Pu-239, Pu-240, Pu-241, and isotope Pu-242 in MTR reactor fuel types which have U-235 contain about 250 gram. The calculation was performed in three steps. The first step is to determine the library of calculation output of BOC (Beginning of Cycle). The second step is to determine the core isotope density, the weight of plutonium for one core, and one fuel isotope density. The third step is to calculate weight of plutonium in gram. All calculation is performed by IAFUEL computer code. The calculation was produced content of each Pu isotopes were Pu-239 is 6.7666 gr, Pu-240 is 1.4628 gr, Pu-241 is 0.52951 gr, and Pu-242 is 0.068952 gr

User's guide to EPIC, a computer program to calculate the motion of fuel and coolant subsequent to pin failure in an LMFBR

Energy Technology Data Exchange (ETDEWEB)

Pizzica, P.A.; Garner, P.L.; Abramson, P.B.

1979-10-01

The computer code EPIC models fuel and coolant motion which results from internal fuel pin pressure (from fission gas or fuel vapor) and possibly from the generation of sodium vapor pressure in the coolant channel subsequent to pin failure in a liquid-metal fast breeder reactor. The EPIC model is restricted to conditions where fuel pin geometry is generally preserved and is not intended to treat the total disruption of the pin structure. The modeling includes the ejection of molten fuel from the pin into a coolant channel with any amount of voiding through a clad breach which may be of any length or which may extend with time. One-dimensional Eulerian hydrodynamics is used to treat the motion of fuel and fission gas inside a molten fuel cavity in the fuel pin as well as the mixture of two-phase sodium and fission gas in the coolant channel. Motion of fuel in the coolant channel is tracked with a type of particle-in-cell technique. EPIC is a Fortran-IV program requiring 400K bytes of storage on the IBM 370/195 computer. 21 refs., 2 figs.

Computational simulation of water transport in PEM fuel cells using an improved membrane model

International Nuclear Information System (INIS)

Cao, J.; Djilali, N.

2000-01-01

Computational models and simulation tools can provide valuable insight and guidance for design, performance optimization, and cost reduction of fuel cells. In proton-exchange membrane fuel cells it is particularly important to maintain appropriate water content and temperature in the electrolyte membrane. In this paper we describe a mathematical model for the membrane that takes into account the diffusion of water, the pressure variation, and the electro-osmotic drag in the membrane. Applying conservation laws for water and current and using an empirical relationship between electro-osmotic drag and water content, we obtain a transport equation for water molar concentration and derive a new equation for the electric potential that accounts for variable water content and is more accurate than the conventionally employed Laplace's equation does. The model is coupled with a computational fluid dynamics model for diffusive transport in the electrodes and convective transport in the reactant flow channels. Simulations for a two-dimensional cell are performed over nominal current densities ranging form i=0.1 A/cm≅ to 1.2 A/cm≅. The impact and importance of temperature and pressure non-uniformity, and of two-dimensionality are assessed and discussed. (author)

Investigation on structural analysis computer program of spent nuclear fuel shipping cask, (2)

International Nuclear Information System (INIS)

Yagawa, Ganki; Ikushima, Takeshi.

1987-10-01

This report describes the results (II) done by the Sub-Committee of Research Cooperation Committee (RC-62) of the Japan Society of Mechanical Engineers under the trust of the Japan Atomic Energy Research Institute. The principal fulfilments and accomplishments are summarized as follows: (1) Regarding the survey of structural analysis methods of spent fuel shipping cask, several documents, which explain the features and applications of the exclusive computer programs for impact analysis on the basis of 2 or 3 dimensional finite element or difference methods, were reviewed. (2) In comparative evaluation of the existing computer programs, the common benchmark test problems for drop impact of the axisymmetric cylinder and plate were adopted the calculational evaluations for taking into account the strain rate effect of material properties, effect of artificial viscosity and effect of time integration step size were carried out. (3) Evaluation of impact analysis algorithm of computer programs were conducted and the requirements for computer programs to be developed in future and an index for further studies have been clarified. (author)

Development of fuel prices and its impact on the future development of nuclear energy, the use of computer code DESAE

International Nuclear Information System (INIS)

Panik, M.; Necas, V.

2007-01-01

The thesis is an overview of fuel prices, its key components, such as the particular price and price of natural uranium fuel enrichment. The paper outlines the expected impact of higher fuel prices on the future development of nuclear energy. The last section is devoted to computer code DESAE, designed to calculate and compare advantages and disadvantages of different nuclear systems, but also to calculate the parameters of given nuclear system. They suggested the possibility of using code in practice. (author)

Radioactive decay properties of CANDU fuel. Volume 1: the natural uranium fuel cycle

International Nuclear Information System (INIS)

Clegg, L.J.; Coady, J.R.

1977-01-01

The computer code CANIGEN was used to obtain the mass, activity, decay heat and toxicity of CANDU fuel and its component isotopes. Data are also presented on gamma spectra and neutron emissions. Part 1 presents these data for unirradiated fuel, uranium ore and uranium mill tailings. In Part 2 they have been computed for fuel irradiated to levels of burnup ranging from 140 GJ/kg U to 1150 GJ/kg U. (author)

Neutronics investigation of CANada Deuterium Uranium 6 reactor fueled (transuranic–Th) O2 using a computational method

OpenAIRE

Zohreh Gholamzadeh; Seyed Mohammad Mirvakili; Hossein Khalafi

2015-01-01

Background: 241Am, 243Am, and 237Np isotopes are among the most radiotoxic components of spent nuclear fuel. Recently, researchers have planned different incineration scenarios for the highly radiotoxic elements of nuclear waste in critical reactors. Computational methods are widely used to predict burnup rates of such nuclear wastes that are used under fuel matrixes in critical reactors. Methods: In this work, the Monte Carlo N-particle transport code was used to calculate the neutronic b...

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

3-Dimensional Computational Fluid Dynamics Modeling of Solid Oxide Fuel Cell Using Different Fuels

Science.gov (United States)

2011-01-01

major types of fuel cells in practice are listed below: Polymer Electrolyte Membrane Fuel Cell ( PEMFC ) Alkaline Fuel cell (AFC) Phosphoric Acid...Material Operating Temperature (oC) Efficiency (%) PEMFC H2, Methanol, Formic Acid Hydrated Organic Polymer < 90 40-50 AFC Pure H2 Aqueous

Fuel behavior modeling using the MARS computer code

International Nuclear Information System (INIS)

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

1983-01-01

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

FARST: A computer code for the evaluation of FBR fuel rod behavior under steady-state/transient conditions

International Nuclear Information System (INIS)

Nakamura, M.; Sakagami, M.

1984-01-01

FARST, a computer code for the evaluation of fuel rod thermal and mechanical behavior under steady-state/transient conditions has been developed. The code characteristics are summarized as follows: (I) FARST evaluates the fuel rod behavior under the transient conditions. The code analyzes thermal and mechanical phenomena within a fuel rod, taking into account the temperature change in coolant surrounding the fuel rod. (II) Permanent strains such as plastic, creep and swelling strains as well as thermoelastic deformations can be analyzed by using the strain increment method. (III) Axial force and contact pressure which act on the fuel stack and cladding are analyzed based on the stick/slip conditions. (IV) FARST used a pellet swelling model which depends on the contact pressure between pellet and cladding, and an empirical pellet relocation model, designated as 'jump relocation model'. The code was successfully applied to analyses of the fuel rod irradiation data from pulse reactor for nuclear safety research in Cadarache (CABRI) and pulse reactor for nuclear safety research in Japan Atomic Energy Research Institute (NSRR). The code was further applied to stress analysis of a 1000 MW class large FBR plant fuel rod during transient conditions. The steady-state model which was used so far gave the conservative results for cladding stress during overpower transient, but underestimated the results for cladding stress during a rapid temperature decrease of coolant. (orig.)

Emission computer tomography on a Dodewaard mixed oxide fuel pin. Comparative PIE work with non-destructive and destructive techniques

Energy Technology Data Exchange (ETDEWEB)

Buurveld, H.A.; Dassel, G.

1993-12-01

A nondestructive technique as well as a destructive PIE technique have been used to verify the results obtained with a newly 8-e computer tomography (GECT) system. Multi isotope Scanning (MIS), electron probe micro analysis (EPMA) and GECT were used on a mixed oxide (MOX) fuel rod from the Dodewaard reactor with an average burnup of 24 MWd/kg fuel. GECT shows migration of Cs to the periphery of fuel pellets and to radial cracks and pores in the fuel, whereas MIS shows Cs migration to pellet interfaces. The EPMA technique appeared not to be useful to show migration of Cs but, it shows the distribution of fission products from Pu. EPMA clearly shows the distribution of fission products from Pu, but did not reveal the Cs-migration. (orig./HP)

Computational fluid dynamics (CFD) modelling of coal/biomass co-firing in pulverised fuel boilers

Energy Technology Data Exchange (ETDEWEB)

Moghtaderi, B.; Meesri, C. [University of Newcastle, Callaghan, NSW (Australia). CRC for Coal in Sustainable Development, Dept. of Chemical Engineering

2002-07-01

The present study is concerned with computational fluid dynamics (CFD) modelling of coal/biomass blends co-fired under conditions pertinent to pulverised fuel (PF) boilers. The attention is particularly focused on the near burner zone to examine the impact of biomass on the flame geometry and temperature. The predictions are obtained by numerical solution of the conservation equations for the gas and particle phases. The gas phase is solved in the Eulerian domain using steady-state time-averaged Navier-Stokes equations while the solution of the particle phase is obtained from a series of Lagrangian particle tracking equations. Turbulence is modelled using the {kappa}-{epsilon} and Reynolds Stress models. The comparison between the predictions and experimental measurement reported in the literature resulted in a good agreement. Other influences of biomass co-firing are observed for fuel devolatilisation and burnout. 19 refs., 6 figs.

10 CFR Appendix II to Part 504 - Fuel Price Computation

Science.gov (United States)

2010-01-01

... DEPARTMENT OF ENERGY (CONTINUED) ALTERNATE FUELS EXISTING POWERPLANTS Pt. 504, App. II Appendix II to Part... effects of future real price increases for each fuel. The delivered price of an alternate fuel used to calculate delivered fuel expenses must reflect the petitioner's delivered price of the alternate fuel and...

Effects of pressure and fuel dilution on coflow laminar methane-air diffusion flames: A computational and experimental study

Science.gov (United States)

Cao, Su; Ma, Bin; Giassi, Davide; Bennett, Beth Anne V.; Long, Marshall B.; Smooke, Mitchell D.

2018-03-01

In this study, the influence of pressure and fuel dilution on the structure and geometry of coflow laminar methane-air diffusion flames is examined. A series of methane-fuelled, nitrogen-diluted flames has been investigated both computationally and experimentally, with pressure ranging from 1.0 to 2.7 atm and CH4 mole fraction ranging from 0.50 to 0.65. Computationally, the MC-Smooth vorticity-velocity formulation was employed to describe the reactive gaseous mixture, and soot evolution was modelled by sectional aerosol equations. The governing equations and boundary conditions were discretised on a two-dimensional computational domain by finite differences, and the resulting set of fully coupled, strongly nonlinear equations was solved simultaneously at all points using a damped, modified Newton's method. Experimentally, chemiluminescence measurements of CH* were taken to determine its relative concentration profile and the structure of the flame front. A thin-filament ratio pyrometry method using a colour digital camera was employed to determine the temperature profiles of the non-sooty, atmospheric pressure flames, while soot volume fraction was quantified, after evaluation of soot temperature, through an absolute light calibration using a thermocouple. For a broad spectrum of flames in atmospheric and elevated pressures, the computed and measured flame quantities were examined to characterise the influence of pressure and fuel dilution, and the major conclusions were as follows: (1) maximum temperature increases with increasing pressure or CH4 concentration; (2) lift-off height decreases significantly with increasing pressure, modified flame length is roughly independent of pressure, and flame radius decreases with pressure approximately as P-1/2; and (3) pressure and fuel stream dilution significantly affect the spatial distribution and the peak value of the soot volume fraction.

Computer code SICHTA-85/MOD 1 for thermohydraulic and mechanical modelling of WWER fuel channel behaviour during LOCA and comparison with original version of the SICHTA code

International Nuclear Information System (INIS)

Bujan, A.; Adamik, V.; Misak, J.

1986-01-01

A brief description is presented of the expansion of the SICHTA-83 computer code for the analysis of the thermal history of the fuel channel for large LOCAs by modelling the mechanical behaviour of fuel element cladding. The new version of the code has a more detailed treatment of heat transfer in the fuel-cladding gap because it also respects the mechanical (plastic) deformations of the cladding and the fuel-cladding interaction (magnitude of contact pressure). Also respected is the change in pressure of the gas filling of the fuel element, the mechanical criterion is considered of a failure of the cladding and the degree is considered of the blockage of the through-flow cross section for coolant flow in the fuel channel. The LOCA WWER-440 model computation provides a comparison of the new SICHTA-85/MOD 1 code with the results of the original 83 version of SICHTA. (author)

Experimental and computation method for determination of burnup and isotopic composition of the WWER-440 fuel using the 134Cs and 137Cs concentrations

International Nuclear Information System (INIS)

Babichev, B.A.; Kozharin, V.V.

1990-01-01

An experimental and computational method for determination of burnup and actinoid concentrations in WWER fuel elements using 134 Cs and 137 Cs concentrations in fuel is considered. It is shown that the error in calculation of fuel burnup and U and Pu isotope concentrations in WWER-440 fuel elements is 1.3-4.9% provided that the error in 134 Cs and 137 Cs concentration measurements does not exceed 1.7 and 1.2%. 9 refs.; 10 figs.; 4 tabs

RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Yuan, Y.C. [Square Y Consultants, Orchard Park, NY (US); Chen, S.Y.; Biwer, B.M.; LePoire, D.J. [Argonne National Lab., IL (US)

1995-11-01

This report presents the technical details of RISKIND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, interactive program that can be run on an IBM or equivalent personal computer under the Windows{trademark} environment. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incident-free models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionuclide inventory and dose conversion factors. In addition, the flexibility of the models allows them to be used for assessing any accidental release involving radioactive materials. The RISKIND code allows for user-specified accident scenarios as well as receptor locations under various exposure conditions, thereby facilitating the estimation of radiological consequences and health risks for individuals. Median (50% probability) and typical worst-case (less than 5% probability of being exceeded) doses and health consequences from potential accidental releases can be calculated by constructing a cumulative dose/probability distribution curve for a complete matrix of site joint-wind-frequency data. These consequence results, together with the estimated probability of the entire spectrum of potential accidents, form a comprehensive, probabilistic risk assessment of a spent nuclear fuel transportation accident.

RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

International Nuclear Information System (INIS)

Yuan, Y.C.; Chen, S.Y.; Biwer, B.M.; LePoire, D.J.

1995-11-01

This report presents the technical details of RISKIND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, interactive program that can be run on an IBM or equivalent personal computer under the Windows trademark environment. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incident-free models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionuclide inventory and dose conversion factors. In addition, the flexibility of the models allows them to be used for assessing any accidental release involving radioactive materials. The RISKIND code allows for user-specified accident scenarios as well as receptor locations under various exposure conditions, thereby facilitating the estimation of radiological consequences and health risks for individuals. Median (50% probability) and typical worst-case (less than 5% probability of being exceeded) doses and health consequences from potential accidental releases can be calculated by constructing a cumulative dose/probability distribution curve for a complete matrix of site joint-wind-frequency data. These consequence results, together with the estimated probability of the entire spectrum of potential accidents, form a comprehensive, probabilistic risk assessment of a spent nuclear fuel transportation accident

Performance Evaluation of Solid Oxide Fuel Cell by Computer ...

African Journals Online (AJOL)

The search for sustainable energy source that can compete with the existing one led to the discovery and acceptance of fuel cell technologies as a perfect replacement for fossil fuel. The ability of Solid Oxide Fuel Cells (SOFC) to capture the heat generation during the process of energy generation from electrochemical ...

Prediction of air-fuel and oxy-fuel combustion through a generic gas radiation property model

International Nuclear Information System (INIS)

Yin, Chungen

2017-01-01

Highlights: • A gas radiation model for general combustion CFD presented, programmed & verified. • Its general applicability/practical accuracy demonstrated in air-fuel and oxy-fuel. • Useful guidelines for air-fuel and oxy-fuel combustion CFD suggested. • Important to include the impact of CO in gas radiation for oxy-fuel combustion CFD. - Abstract: Thermal radiation plays an important role in heat transfer in combustion furnaces. The weighted-sum-of-gray-gases model (WSGGM), representing a good compromise between computational efficiency and accuracy, is commonly used in computational fluid dynamics (CFD) modeling of combustion processes for evaluating gaseous radiative properties. However, the WSGGMs still have some limitations in practical use, e.g., unable to naturally accommodate different combustion environments, difficult to accurately address the variations in species concentrations in a flame, and inconvenient to account for the impacts of participating species other than H_2O and CO_2. As a result, WSGGMs with different coefficients have been published for specific applications. In this paper, a reliable generic model for gaseous radiation property calculation, which is a computationally efficient exponential wide band model (E-EWBM) applicable to combustion CFD and able to naturally solve all the practical limitations of the WSGGMs, is presented, programmed and verified. The model is then implemented to CFD simulation of a 300 kW air-fuel and a 0.8 MW oxy-fuel combustion furnace, respectively, to demonstrate its computational applicability to general combustion CFD and its capability in producing reliable CFD results for different combustion environments. It is found that the usefulness of the WSGGMs in oxy-fuel combustion CFD is compromised if the important impacts of high levels of CO under oxy-fuel combustion cannot be accounted for. The E-EWBM that appropriately takes the impacts of H_2O, CO_2, CO and CH_4 into account is a good replacement

EXTENDCHAIN: a package of computer programs for calculating the buildup of heavy metals, fission products, and activation products in reactor fuel elements

International Nuclear Information System (INIS)

Robertson, M.W.

1977-01-01

Design of HTGR recycle and refabrication facilities requires a detailed knowledge of the concentrations of around 400 nuclides which are segregated into four different fuel particle types. The EXTENDCHAIN package of computer programs and the supporting input data files were created to provide an efficient method for calculating the 1600 different concentrations required. The EXTENDCHAIN code performs zero-dimensional nuclide burnup, decay, and activation calculations in nine energy groups for up to 108 nuclides per run. Preparation and handling of the input and output for the sixteen EXTENDCHAIN runs required to produce the desired data are the most time consuming tasks in the computation of the spent fuel element composition. The EXTENDCHAIN package of computer programs contains four codes to aid in the preparation and handling of these data. Most of the input data such as cross sections, decay constants, and the nuclide interconnection scheme will not change when calculating new cases. These data were developed for the life cycle of a typical HTGR and stored on archive tapes for future use. The fuel element composition for this typical HTGR life has been calculated and the results for an equilibrium recycle reload are presented. 12 figures, 7 tables

EXPEL - a computing module for molten fuel/coolant interactions in fast reactor sub-assemblies

International Nuclear Information System (INIS)

Fishlock, T.P.

1975-10-01

This report describes a module for computing the effects of a molten fuel/coolant interaction in a fast reactor subassembly. The module is to be incorporated into the FRAX code which calculates the consequences of hypothetical whole core accidents. Details of the interaction are unknown and in consequence the model contains a large number of parameters which must be set by assumption. By variation of these parameters the interaction may be made mild or explosive. Results of a parametric survey are included. (author)

Reactor fueling system

International Nuclear Information System (INIS)

Hattori, Noriaki; Hirano, Haruyoshi.

1983-01-01

Purpose: To optimally position a fuel catcher by mounting a television camera to a fuel catching portion and judging video images by the use of a computer or the like. Constitution: A television camera is mounted to the lower end of a fuel catching mechanism for handling nuclear fuels and a fuel assembly disposed within a reactor core or a fuel storage pool is observed directly from above to judge the position for the fuel assembly by means of video signals. Then, the relative deviation between the actual position of the fuel catcher and that set in a memory device is determined and the positional correction is carried out automatically so as to reduce the determined deviation to zero. This enables to catch the fuel assembly without failure and improves the efficiency for the fuel exchange operation. (Moriyama, K.)

Computer program for obtaining thermodynamic and transport properties of air and products of combustion of ASTM-A-1 fuel and air

Science.gov (United States)

Hippensteele, S. A.; Colladay, R. S.

1978-01-01

A computer program for determining desired thermodynamic and transport property values by means of a three-dimensional (pressure, fuel-air ratio, and either enthalpy or temperature) interpolation routine was developed. The program calculates temperature (or enthalpy), molecular weight, viscosity, specific heat at constant pressure, thermal conductivity, isentropic exponent (equal to the specific heat ratio at conditions where gases do not react), Prandtl number, and entropy for air and a combustion gas mixture of ASTM-A-1 fuel and air over fuel-air ratios from zero to stoichiometric, pressures from 1 to 40 atm, and temperatures from 250 to 2800 K.

CASKETSS-2: a computer code system for thermal and structural analysis of nuclear fuel shipping casks (version 2)

International Nuclear Information System (INIS)

Ikushima, Takeshi

1991-08-01

A computer program CASKETSS-2 has been developed for the purpose of thermal and structural analysis of nuclear fuel shipping casks. CASKETSS-2 means a modular code system for CASK Evaluation code system Thermal and Structural Safety (Version 2). Main features of CASKETSS-2 are as follow; (1) Thermal and structural analysis computer programs for one-, two-, three-dimensional geometries are contained in the code system. (2) There are simplified computer programs and a detailed one in the structural analysis part in the code system. (3) Input data generator is provided in the code system. (4) Graphic computer program is provided in the code system. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

Thermal Hydraulic Computational Fluid Dynamics Simulations and Experimental Investigation of Deformed Fuel Assemblies

Energy Technology Data Exchange (ETDEWEB)

Mays, Brian [AREVA Federal Services, Lynchburg, VA (United States); Jackson, R. Brian [TerraPower, Bellevue, WA (United States)

2017-03-08

The project, Toward a Longer Life Core: Thermal Hydraulic CFD Simulations and Experimental Investigation of Deformed Fuel Assemblies, DOE Project code DE-NE0008321, was a verification and validation project for flow and heat transfer through wire wrapped simulated liquid metal fuel assemblies that included both experiments and computational fluid dynamics simulations of those experiments. This project was a two year collaboration between AREVA, TerraPower, Argonne National Laboratory and Texas A&M University. Experiments were performed by AREVA and Texas A&M University. Numerical simulations of these experiments were performed by TerraPower and Argonne National Lab. Project management was performed by AREVA Federal Services. The first of a kind project resulted in the production of both local point temperature measurements and local flow mixing experiment data paired with numerical simulation benchmarking of the experiments. The project experiments included the largest wire-wrapped pin assembly Mass Index of Refraction (MIR) experiment in the world, the first known wire-wrapped assembly experiment with deformed duct geometries and the largest numerical simulations ever produced for wire-wrapped bundles.

Computational fluid dynamics modeling of two-phase flow in a BWR fuel assembly. Final CRADA Report

International Nuclear Information System (INIS)

Tentner, A.

2009-01-01

A direct numerical simulation capability for two-phase flows with heat transfer in complex geometries can considerably reduce the hardware development cycle, facilitate the optimization and reduce the costs of testing of various industrial facilities, such as nuclear power plants, steam generators, steam condensers, liquid cooling systems, heat exchangers, distillers, and boilers. Specifically, the phenomena occurring in a two-phase coolant flow in a BWR (Boiling Water Reactor) fuel assembly include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. Traditionally, the best analysis tools for this purpose of two-phase flow phenomena inside the BWR fuel assembly have been the sub-channel codes. However, the resolution of these codes is too coarse for analyzing the detailed intra-assembly flow patterns, such as flow around a spacer element. Advanced CFD (Computational Fluid Dynamics) codes provide a potential for detailed 3D simulations of coolant flow inside a fuel assembly, including flow around a spacer element using more fundamental physical models of flow regimes and phase interactions than sub-channel codes. Such models can extend the code applicability to a wider range of situations, which is highly important for increasing the efficiency and to prevent accidents.

Computer analysis of elongation of the WWER fuel rod claddings

International Nuclear Information System (INIS)

Scheglov, A.; Proselkov, V.

2008-01-01

In this paper description of mechanisms influencing changes of the WWER fuel cladding length and axial forces influencing fuel and cladding are presented. It is shown that shortening of the fuel claddings in case of high burnup can be explained by the change of the fuel and cladding reference state caused by reduction of the fuel rod power level - during reactor outages. It is noted that the presented calculated data are to be reviewed and interpreted as the preliminary results; further work is needed for their confirmation. (authors)

Prediction of CRUD deposition on PWR fuel using a state-of-the-art CFD-based multi-physics computational tool

Energy Technology Data Exchange (ETDEWEB)

Petrov, Victor [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, 2355 Bonisteel Boulv, Ann Arbor, MI (United States); Kendrick, Brian K. [Theoretical Division (T-1, MS B221), Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Walter, Daniel [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, 2355 Bonisteel Boulv, Ann Arbor, MI (United States); Manera, Annalisa, E-mail: manera@umich.edu [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, 2355 Bonisteel Boulv, Ann Arbor, MI (United States); Secker, Jeffrey [Westinghouse Electric Company Nuclear Fuel Division, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2016-04-01

In the present paper we report about the first attempt to demonstrate and assess the ability of state-of-the-art high-fidelity computational tools to reproduce the complex patterns of CRUD deposits found on the surface of operating Pressurized Water Reactors (PWRs) fuel rods. A fuel assembly of the Seabrook Unit 1 PWR was selected as the test problem. During Seabrook Cycle 5, CRUD induced power shift (CIPS) and CRUD induced localized corrosion (CILC) failures were observed. Measurements of the clad oxide thickness on both failed and non-failed rods are available, together with visual observations and the results from CRUD scrapes of peripheral rods. Blind simulations were performed using the Computational Fluid Dynamics (CFD) code STAR-CCM+ coupled to an advanced chemistry code, MAMBA, developed at Los Alamos National Laboratory. The blind simulations were then compared to plant data, which were released after completion of the simulations.

A COMPUTATIONAL FLUID DYNAMICS ANALYSIS OF AIR FLOW THROUGH A TELECOM BACK-UP UNIT POWERED BY AN AIR-COOLED PROTON EXCHANGE MEMBRANE FUEL CELL

DEFF Research Database (Denmark)

Gao, Xin; Berning, Torsten; Kær, Søren Knudsen

2016-01-01

Proton exchange membrane fuel cells (PEMFC’s) are currently being commercialized for various applications ranging from automotive to stationary such as powering telecom back-up units. In PEMFC’s, oxygen from air is internally combined with hydrogen to form water and produce electricity and heat....... This product heat has to be effectively removed from the fuel cell, and while automotive fuel cells are usually liquid-cooled using a secondary coolant loop similar to the internal combustion engines, stationary fuel cell systems as they are used for telecom back-up applications often rely on excessive air fed...... to the fuel cell cathode to remove the heat. Thereby, the fuel cell system is much simpler and cheaper while the fuel cell performance is substantially lower compared to automotive fuel cells. This work presents a computational fluid dynamics analysis on the heat management of an air-cooled fuel cell powered...

Reduced-order computational model in nonlinear structural dynamics for structures having numerous local elastic modes in the low-frequency range. Application to fuel assemblies

International Nuclear Information System (INIS)

Batou, A.; Soize, C.; Brie, N.

2013-01-01

Highlights: • A ROM of a nonlinear dynamical structure is built with a global displacements basis. • The reduced order model of fuel assemblies is accurate and of very small size. • The shocks between grids of a row of seven fuel assemblies are computed. -- Abstract: We are interested in the construction of a reduced-order computational model for nonlinear complex dynamical structures which are characterized by the presence of numerous local elastic modes in the low-frequency band. This high modal density makes the use of the classical modal analysis method not suitable. Therefore the reduced-order computational model is constructed using a basis of a space of global displacements, which is constructed a priori and which allows the nonlinear dynamical response of the structure observed on the stiff part to be predicted with a good accuracy. The methodology is applied to a complex industrial structure which is made up of a row of seven fuel assemblies with possibility of collisions between grids and which is submitted to a seismic loading

Reduced-order computational model in nonlinear structural dynamics for structures having numerous local elastic modes in the low-frequency range. Application to fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Batou, A., E-mail: anas.batou@univ-paris-est.fr [Université Paris-Est, Laboratoire Modélisation et Simulation Multi Echelle, MSME UMR 8208 CNRS, 5 bd Descartes, 77454 Marne-la-Vallee (France); Soize, C., E-mail: christian.soize@univ-paris-est.fr [Université Paris-Est, Laboratoire Modélisation et Simulation Multi Echelle, MSME UMR 8208 CNRS, 5 bd Descartes, 77454 Marne-la-Vallee (France); Brie, N., E-mail: nicolas.brie@edf.fr [EDF R and D, Département AMA, 1 avenue du général De Gaulle, 92140 Clamart (France)

2013-09-15

Highlights: • A ROM of a nonlinear dynamical structure is built with a global displacements basis. • The reduced order model of fuel assemblies is accurate and of very small size. • The shocks between grids of a row of seven fuel assemblies are computed. -- Abstract: We are interested in the construction of a reduced-order computational model for nonlinear complex dynamical structures which are characterized by the presence of numerous local elastic modes in the low-frequency band. This high modal density makes the use of the classical modal analysis method not suitable. Therefore the reduced-order computational model is constructed using a basis of a space of global displacements, which is constructed a priori and which allows the nonlinear dynamical response of the structure observed on the stiff part to be predicted with a good accuracy. The methodology is applied to a complex industrial structure which is made up of a row of seven fuel assemblies with possibility of collisions between grids and which is submitted to a seismic loading.

Fuel Burn Estimation Using Real Track Data

Science.gov (United States)

Chatterji, Gano B.

2011-01-01

A procedure for estimating fuel burned based on actual flight track data, and drag and fuel-flow models is described. The procedure consists of estimating aircraft and wind states, lift, drag and thrust. Fuel-flow for jet aircraft is determined in terms of thrust, true airspeed and altitude as prescribed by the Base of Aircraft Data fuel-flow model. This paper provides a theoretical foundation for computing fuel-flow with most of the information derived from actual flight data. The procedure does not require an explicit model of thrust and calibrated airspeed/Mach profile which are typically needed for trajectory synthesis. To validate the fuel computation method, flight test data provided by the Federal Aviation Administration were processed. Results from this method show that fuel consumed can be estimated within 1% of the actual fuel consumed in the flight test. Next, fuel consumption was estimated with simplified lift and thrust models. Results show negligible difference with respect to the full model without simplifications. An iterative takeoff weight estimation procedure is described for estimating fuel consumption, when takeoff weight is unavailable, and for establishing fuel consumption uncertainty bounds. Finally, the suitability of using radar-based position information for fuel estimation is examined. It is shown that fuel usage could be estimated within 5.4% of the actual value using positions reported in the Airline Situation Display to Industry data with simplified models and iterative takeoff weight computation.

Borrowing brainpower - sharing insecurities

DEFF Research Database (Denmark)

Wegener, Charlotte; Meier, Ninna; Ingerslev, Karen

2016-01-01

Academic writing is a vital, yet complex skill that must be developed within a doctoral training process. In addition, becoming an academic researcher is a journey of changing sense of self and identity. Through analysis of a group session, we show how the feedback of peers addresses questions...... of structure and writing style along with wider issues of researcher identity. Thus, peer learning is demonstrated as a process of simultaneously building a text and an identity as scholarly researcher. The paper advocates ‘borrowing brainpower’ from peers in order to write better texts and, at the same time......, ‘share insecurities’ during the development of the researcher identity. Based on a distributed notion of peer learning and identity, we point to the need for further research into the everyday activities of doctoral writing groups in order to understand the dynamic relationship between production of text...

Transient heat transfer analysis up to dryout in 3D fuel rods under unideal conditions through the development of a computer code

International Nuclear Information System (INIS)

Martins, Rodolfo I.; Affonso, Renato R.W.; Moreira, Maria de Lourdes; Sampaio, Paulo A. B. de

2017-01-01

In this paper we analyze a conjugated transient heat transfer problem consisting of a nuclear reactor's fuel rod and its intrinsic coolant channel. Our analysis is made possible through a computer code being developed at the Instituto de Engenharia Nuclear (IEN/CNEN). This code is meant to study the temperature behavior in fuel rods which exhibit deviation from their ideal conditions, that is, rods in which the cladding is deformed or the fuel is dislocated. It is also designed to avoid the use of the computationally expensive Navier-Stokes equations. For these reasons, its physical model has as basis a three-dimensional fuel rod coupled to a one-dimensional coolant channel, which are discretized using the finite element method. Intending to study accidental conditions in which the coolant (light water) transcends its saturation temperature, turning into vapor, a homogeneous mixture is used to represent the two-phase flow, and so the coolant channel's energy equation is described using enthalpy. Owing to the fact that temperature and enthalpy are used in the physical model, it became impractical to generate a fully coupled method for solving the pertinent equations. Thus, the conjugated heat transfer problem is solved in a segregated manner through the implementation of an iterative method. Finally, as study cases for this paper we present analyses concerning the behavior of the hottest fuel rod in a Pressurized Water Reactor during a shutdown wherein the residual heat removal system is lost (loss of the reactor's coolant pumps). These studies contemplate cases in which the fuel rod's geometry is ideal or curved. Analyses are also performed for two circumstances of positioning of the fuel inside the rod: concentric and eccentric. (author)

Probabilistic assessment of light water reactor fuel performance

International Nuclear Information System (INIS)

Misfeldt, I.

1978-10-01

A computer system for the statistical evaluation of LWR fuel performance has been developed. The computer code FRP, Fuel Reliability Predictor, calculates the distributions for parameters characterizing the fuel performance and failure probability. The statistical methods employed are either Monte Carlo simulations or low order Taylor approximation. Included in the computer system is a deterministic fuel performance code, which has been verified by comparison with data from irradiation experiments. The distributions for all material data utilized in the fuel simulations are estimations from the best available information in the literature. For the failure prediction, a stress corrosion failure criterion has been derived. The failure criterion is based on data from out-of-reactor stress corrosion experiments performed on unirradiated and irradiated zircaloy with iodine present. By means of an example the typical distributions of the variables characterizing the fuel performance and the accuracy of the methods themselves have been investigated. The application of the computer system is illustrated by a number of examples, these include the evaluation of irradiation experiments, design comparisons, and analyses of minor accidents. (author)

V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

Energy Technology Data Exchange (ETDEWEB)

Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

2010-07-15

V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

International Nuclear Information System (INIS)

Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

2010-07-01

V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

Examination of fuel reinsertion strategies for out-of core fuel management

International Nuclear Information System (INIS)

Comes, S.A.; Turinsky, P.J.

1986-01-01

A computer code for determining out-of-core fuel loading strategies in order to minimize levelized fuel cycle cost within constraints has been developed and previously reported by the authors. While past work in this area has dealt with optimizations during equilibrium operating conditions, this work has considered the more realistic conditions of nonequilibrium cycles. The code, called OCEON, seeks to determine a family of economically attractive fuel reload strategies through the optimum selection of feed batch sizes, enrichments, and partially burned fuel reinsertion strategies within operating constraints. This paper presents recent work on expanding the code to allow for different fuel reinsertion options when determining the family of near-optimum fuel reload strategies

Guide to improving the performance of a manipulator system for nuclear fuel handling through computer controls. Final report

International Nuclear Information System (INIS)

Evans, J.M. Jr.; Albus, J.S.; Barbera, A.J.; Rosenthal, R.; Truitt, W.B.

1975-11-01

The Office of Developmental Automation and Control Technology of the Institute for Computer Sciences and Technology of the National Bureau of Standards provides advising services, standards and guidelines on interface and computer control systems, and performance specifications for the procurement and use of computer controlled manipulators and other computer based automation systems. These outputs help other agencies and industry apply this technology to increase productivity and improve work quality by removing men from hazardous environments. In FY 74 personnel from the Oak Ridge National Laboratory visited NBS to discuss the feasibility of using computer control techniques to improve the operation of remote control manipulators in nuclear fuel reprocessing. Subsequent discussions led to an agreement for NBS to develop a conceptual design for such a computer control system for the PaR Model 3000 manipulator in the Thorium Uranium Recycle Facility (TURF) at ORNL. This report provides the required analysis and conceptual design. Complete computer programs are included for testing of computer interfaces and for actual robot control in both point-to-point and continuous path modes

Comparison of structural computer programs used for the analysis of spent fuel shipping casks

International Nuclear Information System (INIS)

Friley, J.R.

1984-09-01

Several structural analysis computer programs were selected and used in analyses relevant to the hypothetical impact requirements for spent fuel shipping cask designs. The objective of the study was to evaluate the computer codes by performing a series of analyses and comparing results. The code evaluation efforts treated end and side impact situations only. As a result, the models were either one or two dimensional. Both clad lead and solid wall construction types were considered. For clad lead models, frictionless sliding between the lead and cladding was assumed. General agreement was achieved between the codes for problems involving non-clad models. For clad models, agreement between the codes was poor. This work was sponsored by the Department of Energy through the Transportation Technology Center at Sandia National Laboratories, Albuquerque, New Mexico. 16 references, 18 figures, 9 tables

FRAP-T6: a computer code for the transient analysis of oxide fuel rods

International Nuclear Information System (INIS)

Siefken, L.J.; Shah, V.N.; Berna, G.A.; Hohorst, J.K.

1983-06-01

FRAP-T6 is a computer code which is being developed to calculate the transient behavior of a light water reactor fuel rod. This report is an addendum to the FRAP-T6/MODO user's manual which provides the additional user information needed to use FRAP-T6/MOD1. This includes model changes, improvements, and additions, coding changes and improvements, change in input and control language, and example problem solutions to aid the user. This information is designed to supplement the FRAP-T6/MODO user's manual

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

International Nuclear Information System (INIS)

Andreeva, Marina; Totev, Totju; Stoyanov, Stoyan

2008-01-01

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

Comprehensive Analysis of Trends and Emerging Technologies in All Types of Fuel Cells Based on a Computational Method

Directory of Open Access Journals (Sweden)

Takaya Ogawa

2018-02-01

Full Text Available Fuel cells have been attracting significant attention recently as highly efficient and eco-friendly energy generators. Here, we have comprehensively reviewed all types of fuel cells using computational analysis based on a citation network that detects emerging technologies objectively and provides interdisciplinary data to compare trends. This comparison shows that the technologies of solid oxide fuel cells (SOFCs and electrolytes in polymer electrolyte fuel cells (PEFCs are at the mature stage, whereas those of biofuel cells (BFCs and catalysts in PEFCs are currently garnering attention. It does not mean, however, that the challenges of SOFCs and PEFC electrolytes have been overcome. SOFCs need to be operated at lower temperatures, approximately 500 °C. Electrolytes in PEFCs still suffer from a severe decrease in proton conductivity at low relative humidity and from their high cost. Catalysts in PEFCs are becoming attractive as means to reduce the platinum catalyst cost. The emerging technologies in PEFC catalysts are mainly heteroatom-doped graphene/carbon nanotubes for metal-free catalysts and supports for iron- or cobalt-based catalysts. BFCs have also received attention for wastewater treatment and as miniaturized energy sources. Of particular interest in BFCs are membrane reactors in microbial fuel cells and membrane-less enzymatic biofuel cells.

Improvement of the computing speed of the FBR fuel pin bundle deformation analysis code 'BAMBOO'

International Nuclear Information System (INIS)

Ito, Masahiro; Uwaba, Tomoyuki

2005-04-01

JNC has developed a coupled analysis system of a fuel pin bundle deformation analysis code 'BAMBOO' and a thermal hydraulics analysis code ASFRE-IV' for the purpose of evaluating the integrity of a subassembly under the BDI condition. This coupled analysis took much computation time because it needs convergent calculations to obtain numerically stationary solutions for thermal and mechanical behaviors. We improved the computation time of the BAMBOO code analysis to make the coupled analysis practicable. 'BAMBOO' is a FEM code and as such its matrix calculations consume large memory area to temporarily stores intermediate results in the solution of simultaneous linear equations. The code used the Hard Disk Drive (HDD) for the virtual memory area to save Random Access Memory (RAM) of the computer. However, the use of the HDD increased the computation time because Input/Output (I/O) processing with the HDD took much time in data accesses. We improved the code in order that it could conduct I/O processing only with the RAM in matrix calculations and run with in high-performance computers. This improvement considerably increased the CPU occupation rate during the simulation and reduced the total simulation time of the BAMBOO code to about one-seventh of that before the improvement. (author)

COBRA-SFS [Spent Fuel Storage]: A thermal-hydraulic analysis computer code: Volume 1, Mathematical models and solution method

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

Transient heat transfer analysis up to dryout in 3D fuel rods under unideal conditions through the development of a computer code

Energy Technology Data Exchange (ETDEWEB)

Martins, Rodolfo I.; Affonso, Renato R.W.; Moreira, Maria de Lourdes; Sampaio, Paulo A. B. de, E-mail: rodolfoienny@gmail.com [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2017-11-01

In this paper we analyze a conjugated transient heat transfer problem consisting of a nuclear reactor's fuel rod and its intrinsic coolant channel. Our analysis is made possible through a computer code being developed at the Instituto de Engenharia Nuclear (IEN/CNEN). This code is meant to study the temperature behavior in fuel rods which exhibit deviation from their ideal conditions, that is, rods in which the cladding is deformed or the fuel is dislocated. It is also designed to avoid the use of the computationally expensive Navier-Stokes equations. For these reasons, its physical model has as basis a three-dimensional fuel rod coupled to a one-dimensional coolant channel, which are discretized using the finite element method. Intending to study accidental conditions in which the coolant (light water) transcends its saturation temperature, turning into vapor, a homogeneous mixture is used to represent the two-phase flow, and so the coolant channel's energy equation is described using enthalpy. Owing to the fact that temperature and enthalpy are used in the physical model, it became impractical to generate a fully coupled method for solving the pertinent equations. Thus, the conjugated heat transfer problem is solved in a segregated manner through the implementation of an iterative method. Finally, as study cases for this paper we present analyses concerning the behavior of the hottest fuel rod in a Pressurized Water Reactor during a shutdown wherein the residual heat removal system is lost (loss of the reactor's coolant pumps). These studies contemplate cases in which the fuel rod's geometry is ideal or curved. Analyses are also performed for two circumstances of positioning of the fuel inside the rod: concentric and eccentric. (author)

Computational Methods for Nanoscale X-ray Computed Tomography Image Analysis of Fuel Cell and Battery Materials

Science.gov (United States)

Kumar, Arjun S.

Over the last fifteen years, there has been a rapid growth in the use of high resolution X-ray computed tomography (HRXCT) imaging in material science applications. We use it at nanoscale resolutions up to 50 nm (nano-CT) for key research problems in large scale operation of polymer electrolyte membrane fuel cells (PEMFC) and lithium-ion (Li-ion) batteries in automotive applications. PEMFC are clean energy sources that electrochemically react with hydrogen gas to produce water and electricity. To reduce their costs, capturing their electrode nanostructure has become significant in modeling and optimizing their performance. For Li-ion batteries, a key challenge in increasing their scope for the automotive industry is Li metal dendrite growth. Li dendrites are structures of lithium with 100 nm features of interest that can grow chaotically within a battery and eventually lead to a short-circuit. HRXCT imaging is an effective diagnostics tool for such applications as it is a non-destructive method of capturing the 3D internal X-ray absorption coefficient of materials from a large series of 2D X-ray projections. Despite a recent push to use HRXCT for quantitative information on material samples, there is a relative dearth of computational tools in nano-CT image processing and analysis. Hence, we focus on developing computational methods for nano-CT image analysis of fuel cell and battery materials as required by the limitations in material samples and the imaging environment. The first problem we address is the segmentation of nano-CT Zernike phase contrast images. Nano-CT instruments are equipped with Zernike phase contrast optics to distinguish materials with a low difference in X-ray absorption coefficient by phase shifting the X-ray wave that is not diffracted by the sample. However, it creates image artifacts that hinder the use of traditional image segmentation techniques. To restore such images, we setup an inverse problem by modeling the X-ray phase contrast

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.

Fuel element performance computer modelling

International Nuclear Information System (INIS)

Locke, D.H.

1978-01-01

The meeting was attended by 88 participants from 17 countries. Altogether 47 papers were presented. The majority of the presentations contained a description of the equations and solutions used to describe and evaluate some of the physical processes taking place in water reactor fuel pins under irradiation. At the same time, particular attention was paid to the ''bench marking'' of the codes wherein solutions arrived at for particular experiments are compared with the results at the experiments

CFD thermal-hydraulic analysis of a CANDU fuel channel with SEU43 type fuel bundle

International Nuclear Information System (INIS)

Catana, A.; Prisecaru, Ilie; Dupleac, D.; Danila, Nicolae

2009-01-01

This paper presents the numerical investigation of a CANDU fuel channel using CFD (Computational Fluid Dynamics) methodology approach, when SEU43 fuel bundles are used. Comparisons with STD37 fuel bundles are done in order to evaluate the influence of geometrical differences of the fuel bundle types on fluid flow properties. We adopted a strategy to analyze only the significant segments of fuel channel, namely : - the fuel bundle junctions with adjacent segments; - the fuel bundle spacer planes with adjacent segments; - the fuel bundle segments with turbulence enhancement buttons; - and the regular segments of fuel bundles. The computer code used is an academic version of FLUENT code, available from UPB. The complex flow domain of fuel bundles contained in pressure tube and operating conditions determine a high turbulence flow and in some parts of fuel channel also a multi-phase flow. Numerical simulation of the flow in the fuel channel has been achieved by solving the equations for conservation of mass, momentum and energy. For turbulence model the standard k-model is employed although other turbulence models can be used. In this paper we do not consider heat generation and heat transfer capabilities of CFD methods. Boundary conditions for CFD analysis are provided by system and sub-channel analysis. In this paper the discussion is focused on some flow parameters behaviour at the bundle junction, spacer's plane configuration, etc. of a SEU43 fuel bundle in conditions of a typical CANDU 6 fuel channel starting from some experience gained in a previous work. (authors)

No one can play snooker that fast. SCK•CEN performs computer simulations of nuclear fuel behaviour

International Nuclear Information System (INIS)

2014-01-01

The article discusses the project F-Bridge, which is funded under the seventh European framework programme. The objective of F-Bridge apply knowledge gained from fundamental science to to the development of Gen-IV fuels. The ultimate goal is to set up an international exchange platform for the integration and the coordination of nuclear fuel research. As part of the F-Bridge project, the Belgian Nuclear Research Center SCK-CEN collaborates with other institutes in Europe in a work package on multi-scale modelling. The main contribution of SCK-CEN focuses on atomic-scale computer simulations. The employed techniques are based on empirical force fields and describe atomic interactions using simple parameters and analytical formulas. Based on Newton’s theory, the evolution of a series of atoms and their interaction over a given time period can be predicted.

Computational fluid dynamics analysis of an innovative start-up method of high temperature fuel cells using dynamic 3d model

Directory of Open Access Journals (Sweden)

Kupecki Jakub

2017-03-01

Full Text Available The article presents a numerical analysis of an innovative method for starting systems based on high temperature fuel cells. The possibility of preheating the fuel cell stacks from the cold state to the nominal working conditions encounters several limitations related to heat transfer and stability of materials. The lack of rapid and safe start-up methods limits the proliferation of MCFCs and SOFCs. For that reason, an innovative method was developed and verified using the numerical analysis presented in the paper. A dynamic 3D model was developed that enables thermo-fluidic investigations and determination of measures for shortening the preheating time of the high temperature fuel cell stacks. The model was implemented in ANSYS Fluent computational fluid dynamic (CFD software and was used for verification of the proposed start-up method. The SOFC was chosen as a reference fuel cell technology for the study. Results obtained from the study are presented and discussed.

Computational analysis of modern HTGR fuel performance and fission product release during the HFR-EU1 irradiation experiment

Energy Technology Data Exchange (ETDEWEB)

Verfondern, Karl, E-mail: k.verfondern@fz-juelich.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Xhonneux, André, E-mail: xhonneux@lrst.rwth-aachen.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Nabielek, Heinz, E-mail: heinznabielek@me.com [Research Center Jülich, Monschauerstrasse 61, 52355 Düren (Germany); Allelein, Hans-Josef, E-mail: h.j.allelein@fz-juelich.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); RWTH Aachen, Chair for Reactor Safety and Reactor Technology, 52072 Aachen (Germany)

2014-07-01

Highlights: • HFR-EU1 irradiation test demonstrates high quality of HTGR spherical fuel elements. • Irradiation performance is in good agreement with German fuel performance modeling. • International benchmark exercise expected first particle to fail at ∼13–17% FIMA. • EOL silver release is predicted to be in the percentage range. • EOL cesium and strontium are expected to remain at a low level. - Abstract: Various countries engaged in the development and fabrication of modern HTGR fuel have initiated activities of modeling the fuel and fission product release behavior with the aim of predicting the fuel performance under HTGR operating and accident conditions. Verification and validation studies are conducted by code-to-code benchmarking and code-to-experiment comparisons as part of international exercises. The methodology developed in Germany since the 1980s represents valuable and efficient tools to describe fission product release from spherical fuel elements and TRISO fuel performance, respectively, under given conditions. Continued application to new results of irradiation and accident simulation testing demonstrates the appropriateness of the models in terms of a conservative estimation of the source term as part of interactions with HTGR licensing authorities. Within the European irradiation testing program for HTGR fuel and as part of the former EU RAPHAEL project, the HFR-EU1 irradiation experiment explores the potential for high performance of the presently existing German and newly produced Chinese fuel spheres under defined conditions up to high burnups. The fuel irradiation was completed in 2010. Test samples are prepared for further postirradiation examinations (PIE) including heatup simulation testing in the KÜFA-II furnace at the JRC-ITU, Karlsruhe, to be conducted within the on-going ARCHER Project of the European Commission. The paper will describe the application of the German computer models to the HFR-EU1 irradiation test and

Calculation of DND-signals in case of fuel pin failures in KNK II with the computer code FICTION III

International Nuclear Information System (INIS)

Schmuck, I.

1990-11-01

In KNK II two delayed neutron detectors are installed for quick detection of fuel subassembly cladding failures. They record the release of the precursors of the emitters of delayed neutrons into the sodium. The computer code FICTION III calculates the expected delayed neutron signals for certain fuel pin failures, where the user has to set the boundary conditions interactively. In view of FICTION II the advancement of FICTION III consists of the following items: application of the data sets of 105 isotopes, distinction of thermal and fast neutron induced fission, partitioning of the sodium flow into two circuits, consideration of the specific fission rates in 10 fuel pin sections, elaboration of the user's interaction possibilities for input/ output. The capability of FICTION III is shown by means of two applications (UNi-test pin on position 100 and the third KNK fuel subassembly cladding failure). Object of further evaluations will be among other things the analysis of increased delayed neutron signals in regard to the fault location and dimension

Design and construction of wireless personal computer based fuel ...

African Journals Online (AJOL)

Results of the different fuel levels obtained through wireless medium are graphical represented on the PC display for monitoring and prompt actions. The PC interface was made possible using C++ programming language. Keywords: Monitoring system, fuel level, floater, transceiver and Radio frequency. Nigerian Journal of ...

Estimate of fuel burnup spatial a multipurpose reactor in computer simulation; Estimativa da queima espacial do combustivel de um reator multiproposito por simulacao computacional

Energy Technology Data Exchange (ETDEWEB)

Santos, Nadia Rodrigues dos, E-mail: nadia.santos@ifrj.edu.br [Instituto Federal de Educacao, Ciencia e Tecnologia do Rio de Janeiro (IFRJ), Paracambi, RJ (Brazil); Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: malu@ien.gov.br, E-mail: zrlima@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2015-07-01

In previous research, which aimed, through computer simulation, estimate the spatial fuel burnup for the research reactor benchmark, material test research - International Atomic Energy Agency (MTR/IAEA), it was found that the use of the code in FORTRAN language, based on the diffusion theory of neutrons and WIMSD-5B, which makes cell calculation, bespoke be valid to estimate the spatial burnup other nuclear research reactors. That said, this paper aims to present the results of computer simulation to estimate the space fuel burnup of a typical multipurpose reactor, plate type and dispersion. the results were considered satisfactory, being in line with those presented in the literature. for future work is suggested simulations with other core configurations. are also suggested comparisons of WIMSD-5B results with programs often employed in burnup calculations and also test different methods of interpolation values obtained by FORTRAN. Another proposal is to estimate the burning fuel, taking into account the thermohydraulics parameters and the appearance of xenon. (author)

Progress of the DUPIC Fuel Compatibility Analysis (IV) - Fuel Performance

International Nuclear Information System (INIS)

Choi, Hang Bok; Ryu, Ho Jin; Roh, Gyu Hong; Jeong, Chang Joon; Park, Chang Je; Song, Kee Chan; Lee, Jung Won

2005-10-01

This study describes the mechanical compatibility of the direct use of spent pressurized water reactor (PWR) fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel, when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and the fuel handling system in the reactor core by both the experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown that the integrity of the DUPIC fuel is mostly maintained under the high power and high burnup conditions even though some material properties like the thermal conductivity is a little lower compared to the uranium fuel. However it is required to slightly change the current DUPIC fuel design to accommodate the high internal pressure of the fuel element. It is also strongly recommended to perform more irradiation tests of the DUPIC fuel to accumulate a database for the demonstration of the DUPIC fuel performance in the CANDU reactor

Progress of the DUPIC Fuel Compatibility Analysis (IV) - Fuel Performance

Energy Technology Data Exchange (ETDEWEB)

Choi, Hang Bok; Ryu, Ho Jin; Roh, Gyu Hong; Jeong, Chang Joon; Park, Chang Je; Song, Kee Chan; Lee, Jung Won

2005-10-15

This study describes the mechanical compatibility of the direct use of spent pressurized water reactor (PWR) fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel, when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and the fuel handling system in the reactor core by both the experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown that the integrity of the DUPIC fuel is mostly maintained under the high power and high burnup conditions even though some material properties like the thermal conductivity is a little lower compared to the uranium fuel. However it is required to slightly change the current DUPIC fuel design to accommodate the high internal pressure of the fuel element. It is also strongly recommended to perform more irradiation tests of the DUPIC fuel to accumulate a database for the demonstration of the DUPIC fuel performance in the CANDU reactor.

Nuclear fuel cycle facility accident analysis handbook

International Nuclear Information System (INIS)

Ayer, J.E.; Clark, A.T.; Loysen, P.; Ballinger, M.Y.; Mishima, J.; Owczarski, P.C.; Gregory, W.S.; Nichols, B.D.

1988-05-01

The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the radiological risk from accidents in nuclear fuel cycle facility operations. The AAH will enable the user to calculate source term releases from accident scenarios manually or by computer. A major feature of the AAH is development of accident sample problems to provide input to source term analysis methods and transport computer codes. Sample problems and illustrative examples for different accident types are included in the AAH

Uranium accountability for ATR fuel fabrication: Part II. A computer simulation

International Nuclear Information System (INIS)

Dolan, C.A.; Nieschmidt, E.B.; Vegors, S.H. Jr.; Wagner, E.P. Jr.

1977-08-01

A stochastic computer model has been designed to simulate the material control system used during the production of fuel plates for the Advanced Test Reactor. Great care has been taken to see that this model follows the manufacturing and measuring processes used. The model is designed so that manufacturing process and measurement parameters are fed in as input; hence, changes in the manufacturing process and measurement procedures are easily simulated. Individual operations in the plant are described by program subroutines. By varying the calling sequence of these subroutines, variations in the manufacturing process may be simulated. By using this model values for MUF and LEMUF may be calculated for predetermined plant operating conditions. Furthermore the effect on MUF and LEMUF produced by changing plant operating procedures and measurement techniques may also be examined. A sample calculation simulating one inventory period of the plant's operation is included

ADVANCING THE FUNDAMENTAL UNDERSTANDING AND SCALE-UP OF TRISO FUEL COATERS VIA ADVANCED MEASUREMENT AND COMPUTATIONAL TECHNIQUES

Energy Technology Data Exchange (ETDEWEB)

Biswas, Pratim; Al-Dahhan, Muthanna

2012-11-01

to advance the fundamental understanding of the hydrodynamics by systematically investigating the effect of design and operating variables, to evaluate the reported dimensionless groups as scaling factors, and to establish a reliable scale-up methodology for the TRISO fuel particle spouted bed coaters based on hydrodynamic similarity via advanced measurement and computational techniques. An additional objective is to develop an on-line non-invasive measurement technique based on gamma ray densitometry (i.e. Nuclear Gauge Densitometry) that can be installed and used for coater process monitoring to ensure proper performance and operation and to facilitate the developed scale-up methodology. To achieve the objectives set for the project, the work will use optical probes and gamma ray computed tomography (CT) (for the measurements of solids/voidage holdup cross-sectional distribution and radial profiles along the bed height, spouted diameter, and fountain height) and radioactive particle tracking (RPT) (for the measurements of the 3D solids flow field, velocity, turbulent parameters, circulation time, solids lagrangian trajectories, and many other of spouted bed related hydrodynamic parameters). In addition, gas dynamic measurement techniques and pressure transducers will be utilized to complement the obtained information. The measurements obtained by these techniques will be used as benchmark data to evaluate and validate the computational fluid dynamic (CFD) models (two fluid model or discrete particle model) and their closures. The validated CFD models and closures will be used to facilitate the developed methodology for scale-up, design and hydrodynamic similarity. Successful execution of this work and the proposed tasks will advance the fundamental understanding of the coater flow field and quantify it for proper and safe design, scale-up, and performance. Such achievements will overcome the barriers to AGR applications and will help assure that the US maintains

Modeling of PHWR fuel elements using FUDA code

International Nuclear Information System (INIS)

Tripathi, Rahul Mani; Soni, Rakesh; Prasad, P.N.; Pandarinathan, P.R.

2008-01-01

The computer code FUDA (Fuel Design Analysis) is used for modeling PHWR fuel bundle operation history and carry out fuel element thermo-mechanical analysis. The radial temperature profile across fuel and sheath, fission gas release, internal gas pressure, sheath stress and strains during the life of fuel bundle are estimated

Software in support of fuel operation in WWERS

International Nuclear Information System (INIS)

Evdokimov, I.A; Novikov, V.V; Ugrumov, A.V; Shishkin, A.A

2013-01-01

A software package comprising computer codes and fuel monitoring tools is under development in Russia in support of WWER fuel operation. The software package includes an expert computer system designed for failure diagnosis in course of reactor operation, prediction of activity evolution in primary coolant and express analysis of pellet-to-cladding mechanical interaction (PCMI) on rod-by-rod basis under normal and transient modes of operation. Coupled with the expert system, the first version of a graphical interface computer program is developed for NPP operating bodies. One of the features of this program is to launch automatically a fuel performance code for a series of detailed calculations for fuel rods with severe PCMI. The particular rods for calculations are determined by the expert system during the express core analysis. A greater attention is paid to recent results in prediction of fuel behavior after a primary failure has occurred. One of the major risks to further operation of leaking fuel comes from secondary fuel degradation due to massive cladding hydriding. Threshold conditions for initiation of secondary hydriding have been found on the basis of physical modeling. Final criteria of secondary failure occurrence were deduced by applying the model to analysis of post-irradiation examinations of leaking WWER fuel. (authors)

Radioactive decay properties of CANDU fuel. Volume 1: the natural uranium fuel cycle

International Nuclear Information System (INIS)

Clegg, L.J.; Coady, J.R.

1977-01-01

The two books of Volume 1 comprise the first in a three-volume series of compilations on the radioactive decay propertis of CANDU fuel and deal with the natural uranium fuel cycle. Succeeding volumes will deal with fuel cycles based on plutonium recycle and thorium. In Volume 1 which is divided into three parts, the computer code CANIGEN was used to obtain the mass, activity, decay heat and toxicity of CANDU fuel and its component isotopes. Data are also presented on gamma spectra and neutron emissions. Part 3 contains the data relating to the plutonium product and the high level wastes produced during fuel reprocessing. (author)

Increasing TRIGA fuel lifetime with 12 wt.% U TRIGA fuel

Energy Technology Data Exchange (ETDEWEB)

Naughton, W F; Cenko, M J; Levine, S H; Witzig, W F [Pennsylvania State University (United States)

1974-07-01

In-core fuel management studies have been performed for the Penn State Breazeale Reactor (PSBR) wherein 12 wt % U fuel elements are used to replace the standard 8.5 wt % U TRIGA fuel. The core configuration used to develop a calculational model was a 90-element hexagonal array, which is representative of the PSBR core, and consists of five hexagonal rings surrounding a central thimble containing water. The technique employed for refueling the core fully loaded with 8.5 wt % U fuel involves replacing 8.5 wt % U fuel with 12 wt % U fuel using an in-out reloading scheme. A batch reload consists of 6 new 12 wt % U fuel elements. Placing the 12 wt % U fuel in the B ring produces fuel temperatures ({approx}450 {sup o}C) that are well below the 800{sup o}C maximum limitation when the PSBR is operating at its maximum allowed power of 1 Megawatt. The advantages of using new 12 wt % U fuel to replace the burned up 8.5 wt % U fuel in the B ring over refueling strictly with 8.5 wt % U-Zr TRIGA fuel are clearly delineated in Table 1 where cost calculations used the General Atomic pre-1972 prices for TRIGA fuel, i.e., $1500 and $1650 for an 8.5 and 12 wt % U fuel element, respectively. Experimental results obtained to date utilizing the 12 wt % U fuel elements agree with the computed results. (author)

Preliminary analysis of aircraft fuel systems for use with broadened specification jet fuels

Science.gov (United States)

Pasion, A. J.; Thomas, I.

1977-01-01

An analytical study was conducted on the use of broadened specification hydrocarbon fuels in present day aircraft. A short range Boeing 727 mission and three long range Boeing 747 missions were used as basis of calculation for one-day-per-year extreme values of fuel loading, airport ambient and altitude ambient temperatures with various seasonal and climatic conditions. Four hypothetical fuels were selected; two high-vapor-pressure fuels with 35 kPa and 70 kPa RVP and two high-freezing-point fuels with -29 C and -18 C freezing points. In-flight fuel temperatures were predicted by Boeing's aircraft fuel tank thermal analyzer computer program. Boil-off rates were calculated for the high vapor pressure fuels and heating/insulation requirements for the high freezing point fuels were established. Possible minor and major heating system modifications were investigated with respect to heat output, performance and economic penalties for the high freezing point fuels.

The sphere-PAC fuel code 'SPHERE-3'

International Nuclear Information System (INIS)

Wallin, H.

2000-01-01

Sphere-PAC fuel is an advanced nuclear fuel, in which the cladding tube is filled with small fuel spheres instead of the more usual fuel pellets. At PSI, the irradiation behaviour of sphere-PAC fuel is calculated using the computer code SPHERE-3. The paper describes the present status of the SPHERE-3 code, and some results of the qualification process against experimental data. (author)

The sphere-pac fuel code 'SPHERE-3'

International Nuclear Information System (INIS)

Wallin, H.; Nordstroem, L.A.; Hellwig, C.

2001-01-01

Sphere-pac fuel is an advanced nuclear fuel, in which the cladding tube is filled with small fuel spheres instead of the more usual fuel pellets. At PSI, the irradiation behaviour of sphere-pac fuel is calculated using the computer code SPHERE-3. The paper describes the present status of the SPHERE-3 code, and some results of the qualification process against experimental data. (author)

Development of a Computer Program for an Analysis of the Logistics and Transportation Costs of the PWR Spent Fuels in Korea

International Nuclear Information System (INIS)

Cha, Jeong Hun; Choi, Heui Joo; Lee, Jong Youl; Choi, Jong Won

2009-01-01

It is expected that a substantial amount of spent fuels will be transported from the four nuclear power plant (NPP) sites in Korea to a hypothetical centralized interim storage facility or a final repository in the near future. The cost for the transportation is proportional to the amount of spent fuels. In this paper, a cost estimation program is developed based on the conceptual design of a transportation system and a logistics analysis. Using the developed computer program, named as CASK, the minimum capacity of a centralized interim storage facility (CISF) and the transportation cost for PWR spent fuels are calculated. The PWR spent fuels are transported from 4 NPP sites to a final repository (FR) via the CISF. Since NPP sites and the CISF are located along the coast, a sea-transportation is considered and a road-transportation is considered between the CISF and the FR. The result shows that the minimum capacity of the interim storage facility is 15,000 MTU

Core status computing system

International Nuclear Information System (INIS)

Yoshida, Hiroyuki.

1982-01-01

Purpose: To calculate power distribution, flow rate and the like in the reactor core with high accuracy in a BWR type reactor. Constitution: Total flow rate signals, traverse incore probe (TIP) signals as the neutron detector signals, thermal power signals and pressure signals are inputted into a process computer, where the power distribution and the flow rate distribution in the reactor core are calculated. A function generator connected to the process computer calculates the absolute flow rate passing through optional fuel assemblies using, as variables, flow rate signals from the introduction part for fuel assembly flow rate signals, data signals from the introduction part for the geometrical configuration data at the flow rate measuring site of fuel assemblies, total flow rate signals for the reactor core and the signals from the process computer. Numerical values thus obtained are given to the process computer as correction signals to perform correction for the experimental data. (Moriyama, K.)

Development of COMPAS, computer aided process flowsheet design and analysis system of nuclear fuel reprocessing

International Nuclear Information System (INIS)

Homma, Shunji; Sakamoto, Susumu; Takanashi, Mitsuhiro; Nammo, Akihiko; Satoh, Yoshihiro; Soejima, Takayuki; Koga, Jiro; Matsumoto, Shiro

1995-01-01

A computer aided process flowsheet design and analysis system, COMPAS has been developed in order to carry out the flowsheet calculation on the process flow diagram of nuclear fuel reprocessing. All of equipments, such as dissolver, mixer-settler, and so on, in the process flowsheet diagram are graphically visualized as icon on a bitmap display of UNIX workstation. Drawing of a flowsheet can be carried out easily by the mouse operation. Not only a published numerical simulation code but also a user's original one can be used on the COMPAS. Specifications of the equipment and the concentration of components in the stream displayed as tables can be edited by a computer user. Results of calculation can be also displayed graphically. Two examples show that the COMPAS is applicable to decide operating conditions of Purex process and to analyze extraction behavior in a mixer-settler extractor. (author)

Improved fuel design economics - a new evaluation tool

International Nuclear Information System (INIS)

Aboudara, J.L.

1987-01-01

Advanced fuel design technology is now beginning to be implemented with new reload regions for large pressurized water reactors. Until recently there has not been an integrated computer modeling product that would allow easy assessment of the economics of various advanced fuel design alternatives now available to utilities. The Fuel Cycle Scoping System (FCSS) was designed to fill this need. The FCSS is a personnel computer (PC) software package that is used to evaluate alternative strategies for supplying and using nuclear fuel in nuclear power reactors. The FCSS is an extremely flexible package that permits evaluation of in-core and out-of-core fuel management strategy options. For each strategy option, unit and reactor operating assumptions and assumptions for uranium supply, conversion, enrichment, fabrication, and spent fuel disposal can be made

Three-dimensional multi-phase flow computational fluid dynamics model for analysis of transport phenomena and thermal stresses in PEM fuel cells

Energy Technology Data Exchange (ETDEWEB)

Maher, A.R.; Al-Baghdadi, S. [International Technological Univ., London (United Kingdom). Dept. of Mechanical Engineering; Haroun, A.K.; Al-Janabi, S. [Babylon Univ., Babylon (Iraq). Dept. of Mechanical Engineering

2007-07-01

Fuel cell technology is expected to play an important role in meeting the growing demand for distributed generation because it can convert the chemical energy of a clean fuel directly into electrical energy. An operating fuel cell has varying local conditions of temperature, humidity, and power generation across the active area of the fuel cell in 3D. This paper presented a model that was developed to improve the basic understanding of the transport phenomena and thermal stresses in PEM fuel cells, and to investigate the behaviour of polymer membrane under hygro and thermal stresses during the cell operation. This comprehensive 3D, multiphase, non-isothermal model accounts for the major transport phenomena in a PEM fuel cell, notably convective and diffusive heat and mass transfer; electrode kinetics; transport and phase change mechanism of water; and potential fields. The model accounts for the liquid water flux inside the gas diffusion layers by viscous and capillary forces and can therefore predict the amount of liquid water inside the gas diffusion layers. This study also investigated the key parameters affecting fuel cell performance including geometry, materials and operating conditions. The model considers the many interacting, complex electrochemical, transport phenomena, thermal stresses and deformation that cannot be studied experimentally. It was concluded that the model can provide a computer-aided tool for the design and optimization of future fuel cells with much higher power density and lower cost. 21 refs., 2 tabs., 14 figs.

CFD thermal-hydraulic analysis of a CANDU fuel channel

International Nuclear Information System (INIS)

Catana, A.; Prisecaru, I.; Dupleac, D.; Danila, N.

2009-01-01

This paper presents the numerical investigation of a CANDU fuel channel using CFD (Computational fluid dynamics) methodology approach. Limited computer power available at Bucharest University POLITEHNICA forced the authors to analyse only segments of fuel channel namely the significant ones: fuel bundle junctions with adjacent segments, fuel bundle spacer planes with adjacent segments, regular segments of fuel bundles. The computer code used is FLUENT. Fuel bundles contained in pressure tubes forms a complex flow domain. The flow is characterized by high turbulence and in some parts of fuel channel also by multi-phase flow. The flow in the fuel channel has been simulated by solving the equations for conservation of mass and momentum. For turbulence modelling the standard k-e model is employed although other turbulence models can be used as well. In this paper we do not consider heat generation and heat transfer capabilities of CFD methods. Since we consider only some relatively short segments of a CANDU fuel channel we can assume, for this starting stage, that heat transfer is not very important for these short segments of fuel channel. The boundary conditions for CFD analysis are provided by system and sub-channel analysis. In this paper the discussion is focused on some flow parameters behaviour at the bundle junction, spacer's plane configuration, etc. In this paper we present results for Standard CANDU 6 Fuel Bundles as a basis for CFD thermal-hydraulic analysis of INR proposed SEU43 and other new nuclear fuels. (authors)

Model investigation of fuel rod behaviour

International Nuclear Information System (INIS)

Girgis, M.M.; Wiesenack, W.; Stegemann, D.

1985-06-01

Thermal and mechanical behaviour of fuel rods can be explained but unsatisfactorily by models based of an axial symmetry concept. Recently developed models include, with respect to their thermal components, a simple method for the computation of the temperature distribution within the fuel, and they also take into account the influence of excentrically placed pellets for the computation of heat transfer in the cold gap. Additionally, a finite-element model is used to evaluate the effects of cracking and fragmentation on the thermal behaviour of pellets. The reaction of fuel and fuel cladding to external and internal loadings and the axial interaction between fuel and cladding are described in the mechanical portion of the model. A special case of axial coupling is the so-called random stacking interaction caused by fuel pellets placed excentrically at the cladding and sliding radially and axially. In the comparison of measurement results, both thermal and mechanical behaviour of different rods from the OECD Halden Reactor Project are subject to investigations. (RF) [de

CFD analysis of bubble hydrodynamics in a fuel reactor for a hydrogen-fueled chemical looping combustion system

International Nuclear Information System (INIS)

Harichandan, Atal Bihari; Shamim, Tariq

2014-01-01

Highlights: • Computational study of the fuel reactor of chemical looping combustion technology. • The results yield better understanding of the bubble hydrodynamics in fuel reactor. • Increasing the reactor bed length increases the conversion rate. • Small oxygen carrier particles improves the conversion rate. - Abstract: This study investigates the temporal development of bubble hydrodynamics in the fuel reactor of a hydrogen-fueled chemical looping combustion (CLC) system by using a computational model. The model also investigates the molar fraction of products in gas and solid phases. The study assists in developing a better understanding of the CLC process, which has many advantages such as being a potentially promising candidate for an efficient carbon dioxide capture technology. The study employs the kinetic theory of granular flow. The reactive fluid dynamic system of the fuel reactor is customized by incorporating the kinetics of an oxygen carrier reduction into a commercial computational fluid dynamics (CFD) code. An Eulerian multiphase treatment is used to describe the continuum two-fluid model for both gas and solid phases. CaSO 4 and H 2 are used as an oxygen carrier and a fuel, respectively. The computational results are validated with the experimental and numerical results available in the open literature. The CFD simulations are found to capture the features of the bubble formation, rise and burst in unsteady and quasi-steady states very well. The results show a significant increase in the conversion rate with higher dense bed height, lower bed width, higher free board height and smaller oxygen carrier particles which upsurge an overall performance of the CLC plant

Computer Simulations of Composite Electrodes in Solid-Oxide Fuel-Cells

Energy Technology Data Exchange (ETDEWEB)

Sunde, Svein

1999-07-01

Fuel cells are devices for converting the combined chemical (free) energy of fuels and oxygen (air) directly to electrical energy without relying on the dynamic action of steam heated by reacting fuel-oxygen mixtures, like in steam turbines, or of the reacting gas mixtures themselves, like in gas turbines. The basic rationale for fuel cells is their high efficiencies as compared to indirect-conversion methods. Fuel cells are currently being considered for a number of applications, among them de-centralised power supply. Fuel cells come in five basic types and are usually classified according to the type of electrolyte used, which in turn to a significant degree limits the options for anode and cathode materials. The solid-oxide fuel-cell (SOFC) , with which this thesis is concerned, is thus named after its oxide electrolyte, typically the oxide-ion conducting material yttria-stabilised zirconia (YSZ). While the cathode of an SOFC is often uniform in chemical composition (or at least intended to be), various problems of delamination, cracking etc. associated with the use of metallic anode electrocatalysts led to the development of composite SOFC anodes. Porous anodes consisting of Ni and YSZ particles in roughly 50/50 wt-% mixtures are now almost standard with any SOFC-development programme. The designer of composite SOFC electrodes is faced with at least three, interrelated questions: (1) What will be the optimum microstructure and composition of the composite electrode? (2) If the structure changes during operation, as is often observed, what will be the consequences for the internal losses in the cell? (3) How do we interpret electrochemical and conductivity measurements with regard to structure and composition? It is the primary purpose of this thesis to provide a framework for modelling the electrochemical and transport properties of composite electrodes for SOFC, and to arrive at some new insights that cannot be offered by experiment alone. Emphasis is put on

Monitoring fuel consumption for reciprocating engines

International Nuclear Information System (INIS)

Zebelean, D.C.

1991-01-01

This paper reports that Northwest Pipeline Corp. has experienced substantial fuel cost savings through a program that provides continuous monitoring of predicted consumption against actual usage for reciprocating engines. With the continuous monitoring program, the company can gauge specific fuel consumption, flag inefficient engines and determine possible problems with computer software or hardware used to measure engine operations. The plan was initiated as part of an overall effort to reduce pipe line operating cost to remain market competitive by reducing cost of services to the customer. One of the factors in determining Northwest Pipeline's cost of service is the cost of fueling engines to transport natural gas. In 1990, Northwest consumed approximately 9,600,000 MMBtu in engine fuel. Fuel gas always has been accounted for and measured. However, the thermal efficiency of the fuel gas consumed was never quantified. Engineering management set a goal in 1989 to audit fuel consumption monthly on the basis of actual fuel consumed compared to the manufacturer's predicted consumption curves. The fuel consumption comparison between actual consumption with manufacturer's predicted consumption decreased 4% in six months after the first report was published. Unbalanced engines, faulty spark plugs, fuel valves, engines requiring overhauls, and even computer software and hardware problems were found, based on this report. Total decrease in the comparison of 4% was not all realized. Approximately 1.5% of the decrease was due to revising the manufacturer's fuel consumption curves to correctly predict Northwest's vintage of engine

PAPIRUS - a computer code for FBR fuel performance analysis

International Nuclear Information System (INIS)

Kobayashi, Y.; Tsuboi, Y.; Sogame, M.

1991-01-01

The FBR fuel performance analysis code PAPIRUS has been developed to design fuels for demonstration and future commercial reactors. A pellet structural model was developed to describe the generation, depletion and transport of vacancies and atomic elements in unified fashion. PAPIRUS results in comparison with the power - to - melt test data from HEDL showed validity of the code at the initial reactor startup. (author)

Computer modelling of the WWER fuel elements under high burnup conditions by the computer codes PIN-W and RODQ2D

Energy Technology Data Exchange (ETDEWEB)

Valach, M; Zymak, J; Svoboda, R [Nuclear Research Inst. Rez plc, Rez (Czech Republic)

1997-08-01

This paper presents the development status of the computer codes for the WWER fuel elements thermomechanical behavior modelling under high burnup conditions at the Nuclear Research Institute Rez. The accent is given on the analysis of the results from the parametric calculations, performed by the programmes PIN-W and RODQ2D, rather than on their detailed theoretical description. Several new optional correlations for the UO2 thermal conductivity with degradation effect caused by burnup were implemented into the both codes. Examples of performed calculations document differences between previous and new versions of both programmes. Some recommendations for further development of the codes are given in conclusion. (author). 6 refs, 9 figs.

Computer modelling of the WWER fuel elements under high burnup conditions by the computer codes PIN-W and RODQ2D

International Nuclear Information System (INIS)

Valach, M.; Zymak, J.; Svoboda, R.

1997-01-01

This paper presents the development status of the computer codes for the WWER fuel elements thermomechanical behavior modelling under high burnup conditions at the Nuclear Research Institute Rez. The accent is given on the analysis of the results from the parametric calculations, performed by the programmes PIN-W and RODQ2D, rather than on their detailed theoretical description. Several new optional correlations for the UO2 thermal conductivity with degradation effect caused by burnup were implemented into the both codes. Examples of performed calculations document differences between previous and new versions of both programmes. Some recommendations for further development of the codes are given in conclusion. (author). 6 refs, 9 figs

Impact of physical properties of mixture of diesel and biodiesel fuels on hydrodynamic characteristics of fuel injection system

Directory of Open Access Journals (Sweden)

Filipović Ivan M.

2014-01-01

Full Text Available One of the alternative fuels, originating from renewable sources, is biodiesel fuel, which is introduced in diesel engines without major construction modifications on the engine. Biodiesel fuel, by its physical and chemical properties, is different from diesel fuel. Therefore, it is expected that by the application of a biodiesel fuel, the characteristic parameters of the injection system will change. These parameters have a direct impact on the process of fuel dispersion into the engine cylinder, and mixing with the air, which results in an impact on the quality of the combustion process. Method of preparation of the air-fuel mixture and the quality of the combustion process directly affect the efficiency of the engine and the level of pollutant emissions in the exhaust gas, which today is the most important criterion for assessing the quality of the engine. The paper presents a detailed analysis of the influence of physical properties of a mixture of diesel and biodiesel fuels on the output characteristics of the fuel injection system. The following parameters are shown: injection pressure, injection rate, the beginning and duration of injection, transformation of potential into kinetic energy of fuel and increase of energy losses in fuel injection system of various mixtures of diesel and biodiesel fuels. For the analysis of the results a self-developed computer program was used to simulate the injection process in the system. Computational results are verified using the experiment, for a few mixtures of diesel and biodiesel fuels. This paper presents the verification results for diesel fuel and biodiesel fuel in particular.

Perturbation theory in nuclear fuel management optimization

International Nuclear Information System (INIS)

Ho, L.W.

1981-01-01

Nuclear in-core fuel management involves all the physical aspects which allow optimal operation of the nuclear fuel within the reactor core. In most nuclear power reactors, fuel loading patterns which have a minimum power peak are economically desirable to allow the reactors to operate at the highest power density and to minimize the possibility of fuel failure. In this study, perturbation theory along with a binary fuel shuffling technique is applied to predict the effects of various core configurations, and hence, the optimization of in-core fuel management. The computer code FULMNT has been developed to shuffle the fuel assemblies in search of the lowest possible power peaking factor. An iteration approach is used in the search routine. A two-group diffusion theory method is used to obtain the power distribution for the iterations. A comparison of the results of this method with other methods shows that this approach can save computer time. The code also has a burnup capability which can be used to check power peaking throughout the core life

Improvement of Computer Codes Used for Fuel Behaviour Simulation (FUMEX-III). Report of a Coordinated Research Project 2008-2012

International Nuclear Information System (INIS)

2013-03-01

It is fundamental to the future of nuclear power that reactors can be run safely and economically to compete with other forms of power generation. As a consequence, it is essential to develop the understanding of fuel performance and to embody that knowledge in codes to provide best estimate predictions of fuel behaviour. This in turn leads to a better understanding of fuel performance, a reduction in operating margins, flexibility in fuel management and improved operating economics. The IAEA has therefore embarked on a series of programmes addressing different aspects of fuel behaviour modelling with the following objectives: - To assess the maturity and prediction capabilities of fuel performance codes, and to support interaction and information exchange between countries with code development and application needs (FUMEX series); - To build a database of well defined experiments suitable for code validation in association with the OECD Nuclear Energy Agency (OECD/NEA); - To transfer a mature fuel modelling code to developing countries, to support teams in these countries in their efforts to adapt the code to the requirements of particular reactors, and to provide guidance on applying the code to reactor operation and safety assessments; - To provide guidelines for code quality assurance, code licensing and code application to fuel licensing. This report describes the results of the coordinated research project on the ''Improvement of computer codes used for fuel behaviour simulation (FUMEX-III)''. This programme was initiated in 2008 and completed in 2012. It followed previous programmes on fuel modelling: D-COM 1982-1984, FUMEX 1993-1996 and FUMEX-II 2002-2006. The participants used a mixture of data derived from commercial and experimental irradiation histories, in particular data designed to investigate the mechanical interactions occurring in fuel during normal, transient and severe transient operation. All participants carried out calculations on priority

Analysis of Double-encapsulated Fuel Rods

Energy Technology Data Exchange (ETDEWEB)

Hales, Jason Dean [Idaho National Laboratory; Medvedev, Pavel G [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Perez, Danielle Marie [Idaho National Laboratory; Williamson, Richard L [Idaho National Laboratory

2014-09-01

In an LWR fuel rod, the cladding encapsulates the fuel, contains fission products, and transfers heat directly to the water coolant. In some situations, it may be advantageous to separate the cladding from the coolant through use of a secondary cladding or capsule. This may be done to increase confidence that the fuel or fission products will not mix with the coolant, to provide a mechanism for controlling the rod temperature, or to place multiple experimental rodlets within a single housing. With an axisymmetric assumption, it is possible to derive closed-form expressions for the temperature profile in a fuel rod using radially-constant thermal conductivity in the fuel. This is true for both a traditional fuel-cladding rod and a double-encapsulated fuel (fuel, cladding, capsule) configuration. Likewise, it is possible to employ a fuel performance code to analyse both a traditional and a double-encapsulated fuel. In the case of the latter, two sets of gap heat transfer conditions must be imposed. In this work, we review the equations associated with radial heat transfer in a cylindrical system, present analytic and computational results for a postulated power and gas mixture history for IFA-744, and describe the analysis of the AFC-2A, 2B metallic fuel alloy experiments at the Advanced Test Reactor, including the effect of a release of fission products into the cladding-capsule gap. The computational results for these two cases were obtained using BISON, a fuel performance code under development at Idaho National Laboratory.

Potential health hazard of nuclear fuel waste and uranium ore

International Nuclear Information System (INIS)

Mehta, K.; Sherman, G.R.; King, S.G.

1991-06-01

The variation of the radioactivity of nuclear fuel waste (used fuel and fuel reprocessing waste) with time, and the potential health hazard (or inherent radiotoxicity) resulting from its ingestion are estimated for CANDU (Canada Deuterium Uranium) natural-uranium reactors. Four groups of radionuclides in the nuclear fuel waste are considered: actinides, fission products, activation products of zircaloy, and activation products of fuel impurities. Contributions from each of these groups to the radioactivity and to the potential health hazard are compared and discussed. The potential health hazard resulting from used fuel is then compared with that of uranium ore, mine tailings and refined uranium (fresh fuel) on the basis of equivalent amounts of uranium. The computer code HAZARD, specifically developed for these computations, is described

Fuel behavior in advanced water reactors

International Nuclear Information System (INIS)

Bolme, A.B.

1996-01-01

Fuel rod behavior of advanced pressurized water reactors under steady state conditions has been investigated in this study. System-80+ and Westinghouse Vantage-5 fuels have been considered as advanced pressurized water reactor fuels to be analyzed. The purpose of this study is to analyze the sensitivity of ditferent models and the effect of selected design parameters on the overall fuel behavior. FRAPCON-II computer code has been used for the analyses. Different modelling options of FRAPCON-II have also been considered in these analyses. Analyses have been performed in two main parts. In the first part, effects of operating conditions on fuel behavior have been investigated. First, fuel rod response under normal operating conditions has been analyzed. Then, fuel rod response to different fuel ratings has been calculated. In the second part, in order to estimate the effect of design parameters on fuel behavior, parametric analyses have been performed. In this part, the effects of initial gap thickness, as fabricated fuel density, and initial fill gas pressure on fuel behavior have been analyzed. The computations showed that both of the fuel rods used in this study operate within the safety limits. However, FRAPCON-II modelling options have been resulted in different behavior due to their modelling characteristics. Hence, with the absence of experimental data, it is difficult to make assesment for the best fuel parameters. It is also difficult to estimate error associated with the results. To improve the performance of the code, it is necessary to develop better experimental correlations for material properties in order to analyze the eftect ot considerably different design parameters rather than nominal rod parameters

Molten fuel-coolant interaction behaviours of various fast reactor fuels (Paper No. HMT-45-87)

International Nuclear Information System (INIS)

Doshi, J.B.

1987-01-01

A parametric computational model of molten fuel-coolant interaction (MFCI) including a particle size distribution is developed and employed to analyse behaviours of various possible reactor fuels, such as oxide, carbide and metal in MFCI scenario. It is observed that while higher thermal conductivity and lower specific heat of carbide compared to oxide is responsible for higher peak pressure and work done per unit mass, the trend is not observed in the metal fuel. The reason for this is the lower operation temperature and latent heat of metallic fuel. (author). 9 refs., 1 fig

Modeling of constituent redistribution in U-Pu-Zr metallic fuel

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeon Soo [Argonne National Laboratory, Nuclear Engineering, RERTR, 9700 South Cass Avenue, Argonne, IL 60439 (United States)]. E-mail: yskim@anl.gov; Hayes, S.L. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Hofman, G.L. [Argonne National Laboratory, Nuclear Engineering, RERTR, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Yacout, A.M. [Argonne National Laboratory, Nuclear Engineering, RERTR, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

2006-12-01

A computer model was developed to analyze constituent redistribution in U-Pu-Zr metallic nuclear fuels. Diffusion and thermochemical properties were parametrically determined to fit the postirradiation data from a fuel test performed in the Experimental Breeder Reactor II (EBR-II). The computer model was used to estimate redistribution profiles of fuels proposed for the conceptual designs of small modular fast reactors. The model results showed that the level of redistribution of the fuel constituents of the designs was similar to the measured data from EBR-II.

A feasibility study on direct methanol fuel cells for laptop computers based on a cost comparison with lithium-ion batteries

International Nuclear Information System (INIS)

Wee, Jung-Ho

2007-01-01

This paper compares the total cost of direct methanol fuel cell (DMFC) and lithium (Li)-ion battery systems when applied as the power supply for laptop computers in the Korean environment. The average power output and operational time of the laptop computers were assumed to be 20 W and 3000 h, respectively. Considering the status of their technologies and with certain conditions assumed, the total costs were calculated to be US$140 for the Li-ion battery and US$362 for DMFC. The manufacturing costs of the DMFC and Li-ion battery systems were calculated to be $16.65 W -1 and $0.77 W h -1 , and the energy consumption costs to be $0.00051 W h -1 and $0.00032 W h -1 , respectively. The higher fuel consumption cost of the DMFC system was due to the methanol (MeOH) crossover loss. Therefore, the requirements for DMFCs to be able to compete with Li-ion batteries in terms of energy cost include reducing the crossover level to at an order magnitude of -9 and the MeOH price to under $0.5 kg -1 . Under these conditions, if the DMFC manufacturing cost could be reduced to $6.30 W -1 , then the DMFC system would become at least as competitive as the Li-ion battery system for powering laptop computers in Korea. (author)

The spent fuel safety experiment

International Nuclear Information System (INIS)

Harmms, G.A.; Davis, F.J.; Ford, J.T.

1995-01-01

The Department of Energy is conducting an ongoing investigation of the consequences of taking fuel burnup into account in the design of spent fuel transportation packages. A series of experiments, collectively called the Spent Fuel Safety Experiment (SFSX), has been devised to provide integral benchmarks for testing computer-generated predictions of spent fuel behavior. A set of experiments is planned in which sections of unirradiated fuel rods are interchanged with similar sections of spent PWR fuel rods in a critical assembly. By determining the critical size of the arrays, one can obtain benchmark data for comparison with criticality safety calculations. The integral reactivity worth of the spent fuel can be assessed by comparing the measured delayed critical fuel loading with and without spent fuel. An analytical effort to model the experiments and anticipate the core loadings required to yield the delayed critical conditions runs in parallel with the experimental effort

A computer program for structural analysis of fuel elements

International Nuclear Information System (INIS)

Hayashi, I.M.V.; Perrotta, J.A.

1988-01-01

It's presented the code ELCOM for the matrix analysis of tubular structures coupled by rigid spacers, typical of PWR's fuel elements. The code ELCOM makes a static structural analysis, where the displacements and internal forces are obtained for each structure at the joints with the spacers, and also, the natural frequencies and vibrational modes of an equivalent integrated structure are obtained. The ELCOM result is compared to a PWR fuel element structural analysis obtained in published paper. (author) [pt

Nuclear fuel licensing requirements: present status and future trends

International Nuclear Information System (INIS)

Gantchev, T.; Vitkova, M.; Gorinov, I.; Datcheva, D.; Rashkova, N.

2001-01-01

The nuclear fuel licensing process must be directed to establishing of criteria for licensing (fuel safety criteria) and relationship between safety limits, technical specifications and operational conditions. This paper discusses the fuel safety criteria as used by NRC and Russian vendor. A survey on the available fuel behavior, modeling and related computer codes is given with respect to help the licensing process including new safety features of general changes in fuel design and operational conditions. Several types of computer codes that are used in safety analysis are sensitive to fuel-related parameters. The need for further code development and verification has been stated on many occasions: new design elements, such as different cladding materials, higher burnup, different fuel microstructure and use of MOX fuel can affect the performance of these codes. Regulatory inspection practices during operation and refueling in different countries are also shown. Future trends are discussed in particular with regard to the coming high burnup and to new core management schemes

Design of fuel loading for Bohunice V-1 Unit 2 reaktor for fuel cycle No.19

International Nuclear Information System (INIS)

Majercik, J.

1998-01-01

The report contains description of the design of fuel loading for the fuel cycle No. 19 in the V-1 Bohunice Unit 2 reactor. Input data and computer codes used for the development of the design are shown. The fuel loading is characterized by the assortment of the fuel loaded and by the scheme of re shuffling of assemblies in the core. An evaluation of basic neutronic core parameters as relates to the compliance with safety criteria is a part of the report as well

Parameterization of Fuel-Optimal Synchronous Approach Trajectories to Tumbling Targets

Directory of Open Access Journals (Sweden)

David Charles Sternberg

2018-04-01

Full Text Available Docking with potentially tumbling Targets is a common element of many mission architectures, including on-orbit servicing and active debris removal. This paper studies synchronized docking trajectories as a way to ensure the Chaser satellite remains on the docking axis of the tumbling Target, thereby reducing collision risks and enabling persistent onboard sensing of the docking location. Chaser satellites have limited computational power available to them and the time allowed for the determination of a fuel optimal trajectory may be limited. Consequently, parameterized trajectories that approximate the fuel optimal trajectory while following synchronous approaches may be used to provide a computationally efficient means of determining near optimal trajectories to a tumbling Target. This paper presents a method of balancing the computation cost with the added fuel expenditure required for parameterization, including the selection of a parameterization scheme, the number of parameters in the parameterization, and a means of incorporating the dynamics of a tumbling satellite into the parameterization process. Comparisons of the parameterized trajectories are made with the fuel optimal trajectory, which is computed through the numerical propagation of Euler’s equations. Additionally, various tumble types are considered to demonstrate the efficacy of the presented computation scheme. With this parameterized trajectory determination method, Chaser satellites may perform terminal approach and docking maneuvers with both fuel and computational efficiency.

Recent metal fuel safety tests in TREAT

International Nuclear Information System (INIS)

Wright, A.E.; Bauer, T.H.; Lo, R.K.; Robinson, W.R.; Palm, R.G.

1986-01-01

In-reactor safety tests have been performed on metal-alloy reactor fuel to study its response to transient-overpower conditions, in particular, the margin to cladding breach and the axial self-extrusion of fuel within intact cladding. Uranium-fissium EBR-II driver fuel elements of several burnups were tested, some to cladding breach and others to incipient breach. Transient fuel motions were monitored, and time and location of breach were measured. The test results and computations of fuel extrusion and cladding failure in metal-alloy fuel are described

A computer code for calculation of solvent-extraction separation in a multicomponent system with reference to nuclear fuel reprocessing

International Nuclear Information System (INIS)

Carassiti, F.; Liuzzo, G.; Morelli, A.

1982-01-01

Nuclear technology development pointed out the need for a new assessment of the fuel cycle back-end. Treatment and disposal of radioactive wastes arising from nuclear fuel reprocessing is known as one of the problems not yet satisfactorily solved, together with separation process of uranium and plutonium from fission products in highly irradiated fuels. Aim of this work is to present an improvement of the computer code for solvent extraction process calculation previously designed by the authors. The modeling of the extraction system has been modified by introducing a new method for calculating the distribution coefficients. The new correlations were based on deriving empirical functions for not only the apparent equilibrium constants, but also the solvation number. The mathematical model derived for calculating separation performance has been then tested for up to ten components and twelve theoretical stages with minor modifications to the convergence criteria. Suitable correlations for the calculation of the distribution coefficients of Uranium, Plutonium, Nitric Acid and fission products were constructed and used to successfully simulate several experimental conditions. (Author)

Investigation of the Feasibility of Utilizing Gamma Emission Computed Tomography in Evaluating Fission Product Migration in Irradiated TRISO Fuel Experiments

International Nuclear Information System (INIS)

Harp, Jason M.; Demkowicz, Paul A.

2014-01-01

In the High Temperature Gas-Cooled Reactor (HTGR) the TRISO particle fuel serves as the primary fission product containment. However the large number of TRISO particles present in proposed HTGRs dictates that there will be a small fraction (~10"-"4 to 10"-"5) of as manufactured defects and in-pile particle failures that will lead to some fission product release. The matrix material surrounding the TRISO particles in fuel compacts and the structural graphite holding the TRISO particles in place can also serve as sinks for containing any released fission products. However data on the migration of solid fission products through these materials is lacking. One of the primary goals of the AGR-3/4 experiment is to study fission product migration from intentionally failed TRISO particles in prototypic HTGR components such as structural graphite and compact matrix material. In this work, the potential for a Gamma Emission Computed Tomography (GECT) technique to non-destructively examine the fission product distribution in AGR-3/4 components and other irradiation experiments is explored. Specifically, the feasibility of using the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) Precision Gamma Scanner (PGS) system for this GECT application was considered. Previous experience utilizing similar techniques, the expected activities in AGR-3/4 rings, and analysis of this work indicate using GECT to evaluate AGR-3/4 will be feasible. The GECT technique was also applied to other irradiated nuclear fuel systems currently available in the HFEF hot cell, including oxide fuel pins, metallic fuel pins, and monolithic plate fuel. Results indicate GECT with the HFEF PGS is effective. (author)

Description of mathematical models and computer programs

International Nuclear Information System (INIS)

1977-01-01

The paper gives a description of mathematical models and computer programs for analysing possible strategies for spent fuel management, with emphasis on economic analysis. The computer programs developed, describe the material flows, facility construction schedules, capital investment schedules and operating costs for the facilities used in managing the spent fuel. The computer programs use a combination of simulation and optimization procedures for the economic analyses. Many of the fuel cycle steps (such as spent fuel discharges, storage at the reactor, and transport to the RFCC) are described in physical and economic terms through simulation modeling, while others (such as reprocessing plant size and commissioning schedules, interim storage facility commissioning schedules etc.) are subjected to economic optimization procedures to determine the approximate lowest-cost plans from among the available feasible alternatives

Qualification process of dispersion fuels in the IEAR1 research reactor

International Nuclear Information System (INIS)

Domingos, D.B.; Silva, A.T.; Silva, J.E.R.

2010-01-01

Neutronic, thermal-hydraulics and accident analysis calculations were developed to estimate the safety of a miniplate irradiation device (MID) to be placed in the IEA-R1 reactor core. The irradiation device will be used to receive miniplates of U 3 O 8 -Al and U 3 Si 2 -Al dispersion fuels, LEU type (19,9% of 235 U) with uranium densities of, respectively, 3.0 gU/cm 3 and 4.8 gU/cm 3 . The fuel miniplates will be irradiated to nominal 235 U burnup levels of 50% and 80%, in order to qualify the above high-density dispersion fuels to be used in the Brazilian Multipurpose Reactor (RMB), now in the conception phase. For the neutronic calculation, the computer codes CITATION and TWODB were utilized. The computer code FLOW was used to calculate the coolant flow rate in the irradiation device, allowing the determination of the fuel miniplate temperatures with the computer model MTRCR-IEA-R1. A postulated Loss of Coolant Accident (LOCA) was analyzed with the computer code LOSS and TEMPLOCA, allowing the calculation of the fuel miniplate temperatures after the reactor pool draining. This paper also presents a system designed for fuel swelling evaluation. The determination of the fuel swelling will be performed by means of the fuel miniplate thickness measurements along the irradiation time. (author)

Mathematical simulation of stressed-deformed state in rod cylindrical fuel elemnts KONDOR program

International Nuclear Information System (INIS)

Khmelevskij, M.Ya.; Malakhova, E.I.; Dolmatov, P.S.

1987-01-01

A mathematical model for numerical computation of stressed-deformed stae in a rod cylindrical fuel element is developed. The model is based on preliminary discretization of the design scheme and linearization of radial parameters as radius functions. The formulation generality enables to calculate strength parameter kinetics in any circular cylindrical fuel element (e.g. annular fuel element; solid or tubular core; ceramic, metallic or dispersion fuel) for arbitrary transient operating conditions and taking into account all possible loading factors. The method is realized in the KONDOR programm (FORTRAN, ES-1061 computer). An example illustrating computation of stress kinetics in a fast reactor fuel element during transient operation is given

Computer simulation of the behaviour and performance of a CANDU fuel rod

International Nuclear Information System (INIS)

Marino, A.C.

1997-01-01

At the Argentine Atomic Energy Commission (Comision Nacional de Energia Atomica, CNEA) the BACO code (for 'BArra COmbustible', fuel rod) was developed. It allows the simulation of the thermo-mechanical performance of a cylindrical fuel rod in a Pressurized Heavy Water Reactor (PHWR). The standard present version of the code (2.30), is a powerful tool for a relatively easy and complete evaluation of fuel behaviour predictions. Input parameters and, therefore, output ones may include statistical dispersion. As a demonstration of BACO capabilities we include a review of CANDU fuel applications, and the calculation and a parametric analysis of a characteristic CANDU fuel. (author)

Nuclear fuels

International Nuclear Information System (INIS)

Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F.

2009-01-01

, Bubbles and precipitates, Modeling fuel behavior); Modeling defects and fission products in UO 2 ceramic by ab initio computation (Ab initio computation, Point defects in uranium dioxide, Fission products in uranium dioxide, The indispensable coupling of modeling and experiment); Cladding and assembly materials (What is the purpose of cladding?, Zirconium alloys, Claddings: required to exhibit good mechanical strength, Mechanical behavior of irradiated Zr alloys, Claddings: required to prove corrosion resistant); Pellet-cladding interaction (The phenomena involved in pellet-cladding interaction (PCI), Experimental simulation of PCI and the lessons to be drawn from it, The requirement for an experimental basis, Numerical simulation of PCI, Towards a lifting of PCI-related operating constraints); Advanced UO 2 and MOX ceramics (Chromium oxide-doped UO 2 fuel, Novel MOX microstructures); Mechanical behavior of fuel assemblies (Assembly mechanical behavior in normal operating conditions, Assembly mechanical behavior in accident situations, Fuel in a loss of primary coolant accident (LOCA)); Introduction to LOCA-type accident transients (Overview of thermal-hydraulic and fuel-related aspects, Incidence of LOCA transients on the thermal-metallurgical-mechanical behavior of zirconium-base alloy cladding); Fuel in a reactivity insertion accident (RIA) (Safety criteria); Fuel in a severe accident (The VERCORS analytical program, The Phebus-FP global tests, Control of severe accidents in the EPR reactor); In-core fuel management (Relationships between cycle length, maximum burnup, and batch fraction Enrichment and burnable poisons, The impact of the nature of the fuel used, and its evolution, on the major parameters of core physics, and management Prospects for future trends in core management); Fuel cycle material balances (In-core evolution of materials, Decay heat and potential radiotoxicity, Plutonium management); Long-term behavior of spent fuel (The nature of spent nuclear

Nuclear fuels

Energy Technology Data Exchange (ETDEWEB)

Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F

2009-07-01

irradiation, Bubbles and precipitates, Modeling fuel behavior); Modeling defects and fission products in UO{sub 2} ceramic by ab initio computation (Ab initio computation, Point defects in uranium dioxide, Fission products in uranium dioxide, The indispensable coupling of modeling and experiment); Cladding and assembly materials (What is the purpose of cladding?, Zirconium alloys, Claddings: required to exhibit good mechanical strength, Mechanical behavior of irradiated Zr alloys, Claddings: required to prove corrosion resistant); Pellet-cladding interaction (The phenomena involved in pellet-cladding interaction (PCI), Experimental simulation of PCI and the lessons to be drawn from it, The requirement for an experimental basis, Numerical simulation of PCI, Towards a lifting of PCI-related operating constraints); Advanced UO{sub 2} and MOX ceramics (Chromium oxide-doped UO{sub 2} fuel, Novel MOX microstructures); Mechanical behavior of fuel assemblies (Assembly mechanical behavior in normal operating conditions, Assembly mechanical behavior in accident situations, Fuel in a loss of primary coolant accident (LOCA)); Introduction to LOCA-type accident transients (Overview of thermal-hydraulic and fuel-related aspects, Incidence of LOCA transients on the thermal-metallurgical-mechanical behavior of zirconium-base alloy cladding); Fuel in a reactivity insertion accident (RIA) (Safety criteria); Fuel in a severe accident (The VERCORS analytical program, The Phebus-FP global tests, Control of severe accidents in the EPR reactor); In-core fuel management (Relationships between cycle length, maximum burnup, and batch fraction Enrichment and burnable poisons, The impact of the nature of the fuel used, and its evolution, on the major parameters of core physics, and management Prospects for future trends in core management); Fuel cycle material balances (In-core evolution of materials, Decay heat and potential radiotoxicity, Plutonium management); Long-term behavior of spent fuel (The

Computer model for refinery operations with emphasis on jet fuel production. Volume 3: Detailed systems and programming documentation

Science.gov (United States)

Dunbar, D. N.; Tunnah, B. G.

1978-01-01

The FORTRAN computing program predicts flow streams and material, energy, and economic balances of a typical petroleum refinery, with particular emphasis on production of aviation turbine fuels of varying end point and hydrogen content specifications. The program has a provision for shale oil and coal oil in addition to petroleum crudes. A case study feature permits dependent cases to be run for parametric or optimization studies by input of only the variables which are changed from the base case.

Analysis of irradiation temperature in fuel rods of OGL-1 fuel assembly

International Nuclear Information System (INIS)

Fukuda, Kousaku; Kobayashi, Fumiaki; Minato, Kazuo; Ikawa, Katsuichi; Iwamoto, Kazumi

1984-10-01

Irradiation temperature in the fuel rods of 5th OGL-1 fuel assembly was analysed by the system composed by STPDSP2 and TRUMP codes. As the measured input-data, following parameters were allowed for; circumferential heating distribution around the fuel rod, which was measured in the JMTR critical assembly, axial heating distribution through the fuel rod, ratio of peak heatings of three fuel rods, and pre- and post-irradiation outer radii of the fuel compacts and inner radii of the graphite sleeves, which had been measured in PIE of the 5th OGL-1 fuel assembly. In computation the axial distributions of helium coolant temperature through the fuel rod and the heating value of each fuel rod were, firstly, calculated as input data for TRUMP. The TRUMP calculation yielded the temperatures which were fitted in those measured by all of the thermo-couples installed in the fuel rods, by adjusting only the value of the surface heat transfer coefficient, and consequently, the temperatures in all portions of the fuel rod were obtained. The apparent heat transfer coefficient changed to 60% of the initial values in the middle period of irradiation. For this reduction it was deduced that shoot had covered the surface of the fuel rod during irradiation, which was confirmed in PIE. Beside it, several things were found in this analysis. (author)

Characteristics of spent nuclear fuel

International Nuclear Information System (INIS)

Notz, K.J.

1988-04-01

The Office of Civilian Radioactive Waste Management (OCRWM) is responsible for the spent fuels and other wastes that will, or may, eventually be disposed of in a geological repository. The two major sources of these materials are commercial light-water reactor (LWR) spent fuel and immobilized high-level waste (HLW). Other wastes that may require long-term isolation include non-LWR spent fuels and miscellaneous sources such as activated metals. This report deals with spent fuels, but for completeness, the other sources are described briefly. Detailed characterizations are required for all of these potential repository wastes. These characteristics include physical, chemical, and radiological properties. The latter must take into account decay as a function of time. In addition, the present inventories and projected quantities of the various wastes are needed. This information has been assembled in a Characteristics Data Base which provides data in four formats: hard copy standard reports, menu-driven personal computer (PC) data bases, program-level PC data bases, and mainframe computer files. 5 refs., 3 figs., 4 tabs

Performance of candu-6 fuel bundles manufactured in romania nuclear fuel plant

International Nuclear Information System (INIS)

Bailescu, A.; Barbu, A.; Din, F.; Dinuta, G.; Dumitru, I.; Musetoiu, A.; Serban, G.; Tomescu, A.

2013-01-01

The purpose of this article is to present the performance of nuclear fuel produced by Nuclear Fuel Plant (N.F.P.) - Pitesti during 1995 - 2012 and irradiated in units U1 and U2 from Nuclear Power Plant (N.P.P.) Cernavoda and also present the Nuclear Fuel Plant (N.F.P.) - Pitesti concern for providing technology to prevent the failure causes of fuel bundles in the reactor. This article presents Nuclear Fuel Plant (N.F.P.) - Pitesti experience on tracking performance of nuclear fuel in reactor and strategy investigation of fuel bundles notified as suspicious and / or defectives both as fuel element and fuel bundle, it analyzes the possible defects that can occur at fuel bundle or fuel element and can lead to their failure in the reactor. Implementation of modern technologies has enabled optimization of manufacturing processes and hence better quality stability of achieving components (end caps, chamfered sheath), better verification of end cap - sheath welding. These technologies were qualified by Nuclear Fuel Plant (N.F.P.) - Pitesti on automatic and Computer Numerical Control (C.N.C.) programming machines. A post-irradiation conclusive analysis which will take place later this year (2013) in Institute for Nuclear Research Pitesti (the action was initiated earlier this year by bringing a fuel bundle which has been reported defective by pool visual inspection) will provide additional information concerning potential damage causes of fuel bundles due to manufacturing processes. (authors)

DIADEME: A computer code to assess in operation defective fuel characteristics and primary circuit contamination

Energy Technology Data Exchange (ETDEWEB)

Genin, J.B. [DEN/DEC/S3C, CEA Cadarache, 13 - Saint-Paul-lez-Durance (France); Harrer, A. [EdF/SEPTEN, 69 - Villeurbanne (France); Musante, Y. [FRAMATOME-ANP, 69 - Lyon (France)

2002-07-01

DIADEME is a computer code developed within the framework of R and D cooperation between the French Atomic Energy Commission (CEA), Electricite de France (EdF) and FRAMATOME-ANP. Its aim is to assess in operation defective fuel characteristics and primary circuit contamination for actinides and long half-life fission products involved in health physics problems as well as in waste and decommissioning studies. DIADEME has been developed and qualified for the EDF nuclear power plants. For many years, both theoretical and experimental studies have been carried out at the CEA on the release of fission products and actinides out of defective fuel rods in operation, their migration and deposition in PWR primary circuits. These studies have allowed defect characteristic diagnosis methods to be developed, based on radiochemical measurements of the primary coolant. These methods are generally used along with gamma spectrometry measurements on primary water sampling. In order to be completely efficient, these methods can also be used in connection with an on-line primary water gamma spectrometry device. This permits to obtain the most comprehensive data on fission product activity evolutions at steady state and during operation transients, and allows the on-line characterization of the defective fuel assemblies. For long half-life fission products and for actinides, DIADEME is also able to assess the activities of soluble and insoluble forms in the primary water and in the chemical and voluminal control system (CVCS) filters and resins, as well as those activities deposited on primary circuit surfaces. (author)

DIADEME: A computer code to assess in operation defective fuel characteristics and primary circuit contamination

International Nuclear Information System (INIS)

Genin, J.B.; Harrer, A.; Musante, Y.

2002-01-01

DIADEME is a computer code developed within the framework of R and D cooperation between the French Atomic Energy Commission (CEA), Electricite de France (EdF) and FRAMATOME-ANP. Its aim is to assess in operation defective fuel characteristics and primary circuit contamination for actinides and long half-life fission products involved in health physics problems as well as in waste and decommissioning studies. DIADEME has been developed and qualified for the EDF nuclear power plants. For many years, both theoretical and experimental studies have been carried out at the CEA on the release of fission products and actinides out of defective fuel rods in operation, their migration and deposition in PWR primary circuits. These studies have allowed defect characteristic diagnosis methods to be developed, based on radiochemical measurements of the primary coolant. These methods are generally used along with gamma spectrometry measurements on primary water sampling. In order to be completely efficient, these methods can also be used in connection with an on-line primary water gamma spectrometry device. This permits to obtain the most comprehensive data on fission product activity evolutions at steady state and during operation transients, and allows the on-line characterization of the defective fuel assemblies. For long half-life fission products and for actinides, DIADEME is also able to assess the activities of soluble and insoluble forms in the primary water and in the chemical and voluminal control system (CVCS) filters and resins, as well as those activities deposited on primary circuit surfaces. (author)

Development of fuel cycles with new fuel with 8.9 mm external diameter for VVER-440. Preliminary assessment of operating efficiency

Energy Technology Data Exchange (ETDEWEB)

Gagarinskiy, Alexey [National Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

2017-09-15

Since the introduction of VVERs-440, their fuel assemblies are subject to ongoing improvements. Until now, the basic structural parameters of fuel, such as rod diameter of 9.1 mm, have never changed. This paper focuses on computational estimates of basic neutronic parameters of the fuel cycle that involves assemblies consisting of fuel rods with diameter reduced to 8.9 mm.

Spent fuel heatup following loss of water during storage

International Nuclear Information System (INIS)

Benjamin, A.S.; McCloskey, D.J.; Powers, D.A.; Dupree, S.A.

1979-03-01

An analysis of spent fuel heatup following a hypothetical accident involving drainage of the storage pool is presented. Computations based upon a new computer code called SFUEL have been performed to assess the effect of decay time, fuel element design, storage rack design, packing density, room ventilation, drainage level, and other variables on the heatup characteristics of the spent fuel and to predict the conditions under which clad failure will occur. Possible storage pool design modifications and/or onsite emergency action have also been considered

Computational simulation of the microstructure of irradiation damaged regions for the plate type fuel of UO2 microspheres dispersed in stainless steel matrix

International Nuclear Information System (INIS)

Reis, S.C. dos; Lage, A.F.; Braga, D.; Ferraz, W.B.

2006-01-01

Plate type fuel elements have high efficiency of thermal transference what benefits the heat flux with high rates of power output. In reactor cores, fuel elements, in general, are subject to a high neutrons flux, high working temperatures, severe corrosion conditions, direct interference of fission products that result from nuclear reactions and radiation interaction-matter. For plate type fuels composed of ceramic particles dispersed in metallic matrix, one can observe the damage regions that arise due to the interaction fission products in the metallic matrix. Aiming at evaluating the extension of the damage regions in function of the particles and its diameters, in this paper, computational geometric simulations structure of plate type fuel cores, composed of UO 2 microspheres dispersed in stainless steel in several fractions of volume and diameters were carried out. The results of the simulations were exported to AutoCAD R where it was possible its visualization and analysis. (author)

The fuel performance code future

International Nuclear Information System (INIS)

Ronchi, C.; Van de Laar, J.

1988-01-01

The paper describes the LWR version of the fuel performance code FUTURE, which was recently developed to calculate the fuel response (swelling, cladding deformation, release) to reactor transient conditions, starting from a broad-based description of the processes of major concern. The main physical models assumed are presented together with the scheme of the computer program

SATURN-FS 1: A computer code for thermo-mechanical fuel rod analysis

International Nuclear Information System (INIS)

Ritzhaupt-Kleissl, H.J.; Heck, M.

1993-09-01

The SATURN-FS code was written as a general revision of the SATURN-2 code. SATURN-FS is capable to perform a complete thermomechanical analysis of a fuel pin, with all thermal, mechanical and irradiation-based effects. Analysis is possible for LWR and for LMFBR fuel pins. The thermal analysis consists of calculations of the temperature profile in fuel, gap and in the cladding. Pore migration, stoichiometry change of oxide fuel, gas release and diffusion effects are taken into account. The mechanical modeling allows the non steady-state analysis of elastic and nonelastic fuel pin behaviour, such as creep, strain hardening, recovery and stress relaxation. Fuel cracking and healing is taken into account as well as contact and friction between fuel and cladding. The modeling of the irradiation effects comprises swelling and fission gas production, Pu-migration and irradiation induced creep. The code structure, the models and the requirements for running the code are described in the report. Recommendations for the application are given. Program runs for verification and typical examples of application are given in the last part of this report. (orig.) [de

The state-of-the-art and problems of fuel element structural analysis

International Nuclear Information System (INIS)

Lassmann, K.

1980-02-01

This study of fuel element structural analysis is arranged in two parts: In the first, self-contained, part the general basic principles of deterministic computer programs for structural analysis of fuel elements are reviewed critically and an approach is shown which can be used to expand the system with respect to statistical investigations. The second part contains technical details summarized in 11 publications, all of which appeared in periodicals with reviewer teams. The major aspects of this study are thought to be the following ones: Contributions to the 'philosophy' of fuel element structural analysis. Critical analysis of the basic structure of computer programs. Critical analysis of the mechanical concept of integral fuel rod computer programs. Establishment of a comprehensive computer program system (URANUS). Expansion from purely deterministic information by statistical analyses. Methodological and computer program developments for the analysis of fast accidents. (orig.) 891 HP/orig. 892 MKO [de

Computer applications for the Fast Flux Test Facility

International Nuclear Information System (INIS)

Worth, G.A.; Patterson, J.R.

1976-01-01

Computer applications for the FFTF reactor include plant surveillance functions and fuel handling and examination control functions. Plant surveillance systems provide the reactor operator with a selection of over forty continuously updated, formatted displays of correlated data. All data are checked for limits and validity and the operator is advised of any anomaly. Data are also recorded on magnetic tape for historical purposes. The system also provides calculated variables, such as reactor thermal power and anomalous reactivity. Supplementing the basic plant surveillance computer system is a minicomputer system that monitors the reactor cover gas to detect and characterize absorber or fuel pin failures. In addition to plant surveillance functions, computers are used in the FFTF for controlling selected refueling equipment and for post-irradiation fuel pin examination. Four fuel handling or examination systems operate under computer control with manual monitoring and over-ride capability

Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

Science.gov (United States)

Williamson, R. L.; Capps, N. A.; Liu, W.; Rashid, Y. R.; Wirth, B. D.

2016-11-01

Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial ( R- Z) or plane radial-circumferential ( R- θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. In comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.

The sphere-PAC fuel code 'SPHERE-3'

Energy Technology Data Exchange (ETDEWEB)

Wallin, H

2000-07-01

Sphere-PAC fuel is an advanced nuclear fuel, in which the cladding tube is filled with small fuel spheres instead of the more usual fuel pellets. At PSI, the irradiation behaviour of sphere-PAC fuel is calculated using the computer code SPHERE-3. The paper describes the present status of the SPHERE-3 code, and some results of the qualification process against experimental data. (author)

Used fuel disposition in crystalline rocks

Energy Technology Data Exchange (ETDEWEB)

Wang, Y. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kalinina, Elena Arkadievna [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Jerden, James L. [Argonne National Lab. (ANL), Argonne, IL (United States); Copple, Jacqueline M. [Argonne National Lab. (ANL), Argonne, IL (United States); Cruse, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Ebert, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Buck, E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Eittman, R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Tinnacher, R. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Tournassat, Christophe. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Davis, J. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Viswanathan, H. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Chu, S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dittrich, T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hyman, F. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Karra, S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Makedonska, N. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reimus, P. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Zavarin, Mavrik [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Joseph, C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2016-09-01

The U.S. Department of Energy Office of Nuclear Energy, Office of Fuel Cycle Technology established the Used Fuel Disposition Campaign (UFDC) in fiscal year 2010 (FY10) to conduct the research and development (R&D) activities related to storage, transportation and disposal of used nuclear fuel and high level nuclear waste. The objective of the Crystalline Disposal R&D Work Package is to advance our understanding of long-term disposal of used fuel in crystalline rocks and to develop necessary experimental and computational capabilities to evaluate various disposal concepts in such media.

Safety analysis of spent fuel packaging

International Nuclear Information System (INIS)

Akamatsu, Hiroshi; Taniuchi, Hiroaki; Tai, Hideto

1987-01-01

Many types of spent fuel packagings have been manufactured and been used for transport of spent fuels discharged from nuclear power plant. These spent fuel packagings need to be assesed thoroughly about safety transportation because spent fuels loaded into the packaging have high radioactivity and generation of heat. This paper explains the outline of safety analysis of a packaging, Safety analysis is performed for structural, thermal, containment, shielding and criticality factors, and MARC-CDC, TRUMP, ORIGEN, QAD, ANISN, KENO, etc computer codes are used for such analysis. (author)

Perturbation theory in nuclear fuel management optimization

International Nuclear Information System (INIS)

Ho, L.W.; Rohach, A.F.

1982-01-01

Perturbation theory along with a binary fuel shuffling technique is applied to predict the effects of various core configurations and, hence, the optimization of in-core fuel management. The computer code FULMNT has been developed to shuffle the fuel assemblies in search of the lowest possible power peaking factor. An iteration approach is used in the search routine. A two-group diffusion theory method is used to obtain the power distribution for the iterations. A comparison of the results of this method with other methods shows that this approach can save computer time and obtain better power peaking factors. The code also has a burnup capability that can be used to check power peaking throughout the core life

Nuclear reactor fuel rod behavior modelling and current trends

International Nuclear Information System (INIS)

Colak, Ue.

2001-01-01

Safety assessment of nuclear reactors is carried out by simulating the events to taking place in nuclear reactors by realistic computer codes. Such codes are developed in a way that each event is represented by differential equations derived based on physical laws. Nuclear fuel is an important barrier against radioactive fission gas release. The release of radioactivity to environment is the main concern and this can be avoided by preserving the integrity of fuel rod. Therefore, safety analyses should cover an assessment of fuel rod behavior with certain extent. In this study, common approaches for fuel behavior modeling are discussed. Methods utilized by widely accepted computer codes are reviewed. Shortcomings of these methods are explained. Current research topics to improve code reliability and problems encountered in fuel rod behavior modeling are presented

Characterizing Gaseous Fuels for Their Knock Resistance based on the Chemical and Physical Properties of the Fuel

NARCIS (Netherlands)

Levinsky, Howard; Gersen, Sander; van Essen, Martijn; van Dijk, Gerco

2016-01-01

A method is described to characterize the effects of changes in the composition of gaseous fuels on engine knock by computing the autoignition process during the compression and burn periods of the engine cycle. To account for the effects of fuel composition on the in-cylinder pressure and

Spent fuel reprocessing system availability definition by process simulation

International Nuclear Information System (INIS)

Holder, N.; Haldy, B.B.; Jonzen, M.

1978-05-01

To examine nuclear fuel reprocessing plant operating parameters such as maintainability, reliability, availability, equipment redundancy, and surge storage requirements and their effect on plant throughput, a computer simulation model of integrated HTGR fuel reprocessing plant operations is being developed at General Atomic Company (GA). The simulation methodology and the status of the computer programming completed on reprocessing head end systems is reported

Nuclear fuel utilization at the Kozloduy NPP

Energy Technology Data Exchange (ETDEWEB)

Boyadzhiev, Z [Kombinat Atomna Energetika, Kozloduj (Bulgaria); Kharalampieva, Ts; Pejchinov, Ts

1996-12-31

Data on core loading and operation conditions during past fuel cycles of the Units 1 to 6 at the Kozloduy NPP are presented. The Units 1 and 2 have reached average discharge fuel burn-up of 31 MW d/kg U, the Unit 3 - 34 MW d/kg U and the Unit 4 - 36.5 MW d/kg U. By use of dummies and low-leakage core loading patterns for WWER-440 cores an effective reduction in reactor pressure vessel irradiation is obtained. By increasing the enrichment level and improving the characteristics of the Units 3 and 4, a design fuel cycles duration has been reduced by 5-10% in number of assemblies. Core loading design has been modelled using computer codes SPPS-1, BIPR-7, ALBOM, PROROC. A 3-year fuel cycle utilizing 4.4% enriched fuel proved to be more efficient for WWER-1000 by 15% reduction in fuel cost compared to the 2-year cycle. Future developments include improvements of the in-core monitoring system and process on-line simulation based on more accurate computer codes. 7 refs., 6 figs., 10 tabs.

Examination of irradiated fuel elements using gamma scanning technique

International Nuclear Information System (INIS)

Ichim, O.; Mincu, M.; Man, I.; Stanica, M.

2016-01-01

The purpose of this paper is to validate the gamma scanning technique used to calculate the activity of gamma fission products from CANDU/TRIGA irradiated fuel elements. After a short presentation of the equipments used and their characteristics, the paper describes the calibration technique for the devices and how computed tomography reconstruction is done. Following the previously mentioned steps is possible to obtain the axial and radial profiles and the computed tomography reconstruction for calibration sources and for the irradiated fuel elements. The results are used to validate the gamma scanning techniques as a non-destructive examination method. The gamma scanning techniques will be used to: identify the fission products in the irradiated CANDU/TRIGA fuel elements, construct the axial and radial distributions of fission products, get the distribution in cross section through computed tomography reconstruction, and determine the nuclei number and the fission products activity of the irradiated CANDU/TRIGA fuel elements. (authors)

Production process and quality control for the HTTR fuel

International Nuclear Information System (INIS)

Yoshimuta, S.; Suzuki, N.; Kaneko, M.; Fukuda, K.

1991-01-01

Development of the production and inspection technology for High Temperature Engineering Test Reactor (HTTR) fuel has been carried out by cooperative work between Japan Atomic Energy Research Institute (JAERI) and Nuclear Fuel Industries, Ltd (NFI). The performance and the quality level of the developed fuel are well established to meet the design requirements of the HTTR. For the commercial scale production of the fuel, statistical quality control and quality assurance must be carefully considered in order to assure the safety of the HTTR. It is also important to produce the fuel under well controlled process condition. To meet these requirements in the production of the HTTR fuel, a new production process and quality control system is to be introduced in the new facilities. The main feature of the system is a computer integrated control system. Process control data at each production stage of products and semi-products are all gathered by terminal computers and processed by a host computer. The processed information is effectively used for the production, quality and accountancy control. With the aid of this system, all the products will be easily traceable from starting materials to final stages and the statistical evaluation of the quality of products becomes more reliable. (author). 8 figs

National Science Bowl | NREL

Science.gov (United States)

Science Bowl National Science Bowl The Department of Energy's Office of Science sponsors the National Science Bowl competition. This fun, fast-paced academic tournament tests the brainpower of middle and high school student teams on science and math topics. The National Science Bowl provides an

MODELING OF FUEL SPRAY CHARACTERISTICS AND DIESEL COMBUSTION CHAMBER PARAMETERS

Directory of Open Access Journals (Sweden)

G. M. Kukharonak

2011-01-01

Full Text Available The computer model for coordination of fuel spray characteristics with diesel combustion chamber parameters has been created in the paper.  The model allows to observe fuel sprays  develоpment in diesel cylinder at any moment of injection, to calculate characteristics of fuel sprays with due account of a shape and dimensions of a combustion chamber, timely to change fuel injection characteristics and supercharging parameters, shape and dimensions of a combustion chamber. Moreover the computer model permits to determine parameters of holes in an injector nozzle that provides the required fuel sprays characteristics at the stage of designing a diesel engine. Combustion chamber parameters for 4ЧН11/12.5 diesel engine have been determined in the paper.

Automatic fuel charging machine for nuclear power plants

International Nuclear Information System (INIS)

Neda, Toshikazu; Aoki, Shigeo.

1978-01-01

Purpose: To enable smooth, safety and rapid refueling by automatically conducting a series of fuel exchange steps through the combination of calculations for fuel control and refueling works. Constitution: Processing operations are conducted based on various data from the memory unit of a computer that processes and stores various data inputted from a nuclear power plant, calculation systems stored in the memory unit of another computer, peripheral units such as typewriters and process input units. A refueling platform is operated by way of a platform control device and a platform driving device, and fuel exchange is conducted by the operation of a channel mounting and demounting device. (Yoshino, Y.)

Fuel performance computer code simulation of steady-state and transient regimes of the stainless steel fuel rods

International Nuclear Information System (INIS)

Gomes, Daniel de Souza

2014-01-01

The immediate cause of the accident at the Fukushima Daiichi nuclear plant in March 2011 was the meltdown of the reactor core. During this process, the zirconium cladding of the fuel reacts with water, producing a large amount of hydrogen. This hydrogen, combined with volatile radioactive materials leaked from the containment vessel and entered the building of the reactor, resulting in explosions. In the past, stainless steel was used as the coating in many pressurized water reactors (PWR) under irradiation and their performance was excellent, however, the stainless steel was replaced by a zirconium-based alloy as a coating material mainly due to its lower section shock-absorbing neutrons. Today, the stainless steel finish appears again as a possible solution for security issues related to the explosion and hydrogen production. The objective of this thesis is to discuss the performance under irradiation of fuel rods using stainless steel as a coating material. The results showed that stainless steel rods exhibit lower temperatures and higher fuel pellet width of the gap - coating the coated rods Zircaloy and this gap does not close during the irradiation. The thermal performance of the two fuel rods is very similar, and the penalty of increased absorption of neutrons due to the use of stainless steel can be offset by the combination of a small increase in the enrichment of U- 235 and changes in the size of the spacing between the fuel rods. (author)

Vibration of fuel bundles

International Nuclear Information System (INIS)

Chen, S.S.

1975-06-01

Several mathematical models have been proposed for calculating fuel rod responses in axial flows based on a single rod consideration. The spacing between fuel rods in liquid metal fast breeder reactors is small; hence fuel rods will interact with one another due to fluid coupling. The objective of this paper is to study the coupled vibration of fuel bundles. To account for the fluid coupling, a computer code, AMASS, is developed to calculate added mass coefficients for a group of circular cylinders based on the potential flow theory. The equations of motion for rod bundles are then derived including hydrodynamic forces, drag forces, fluid pressure, gravity effect, axial tension, and damping. Based on the equations, a method of analysis is presented to study the free and forced vibrations of rod bundles. Finally, the method is applied to a typical LMFBR fuel bundle consisting of seven rods

Management number identification method for nuclear fuel assembly

International Nuclear Information System (INIS)

Furuya, Nobuo; Mori, Kazuma.

1995-01-01

In the present invention, a management number indicated to appropriate portions of a fuel assembly can be read with no error for the management of nuclear fuel materials in the nuclear fuel assembly (counting management) and physical protection: PP. Namely, bar codes as a management number are printed by electrolytic polishing to one or more portions of a side surface of an upper nozzle of the assembly, an upper surface of a clamp and a side surface of a lower nozzle. The bar codes are read by a reader at one or more portions in a transporting path for transporting the fuel assembly and at a fuel detection device disposed in a fuel storage pool. The read signals are inputted to a computer. With such procedures, the nuclear fuel assembly can be identified with no error by reading the bar codes and without applying no danger to a human body. Since the reader is disposed in the course of the transportation and test for the assembly, and the read signals are inputted to the computer, the management for the counting number and PP is facilitated. (I.S.)

The scale analysis sequence for LWR fuel depletion

International Nuclear Information System (INIS)

Hermann, O.W.; Parks, C.V.

1991-01-01

The SCALE (Standardized Computer Analyses for Licensing Evaluation) code system is used extensively to perform away-from-reactor safety analysis (particularly criticality safety, shielding, heat transfer analyses) for spent light water reactor (LWR) fuel. Spent fuel characteristics such as radiation sources, heat generation sources, and isotopic concentrations can be computed within SCALE using the SAS2 control module. A significantly enhanced version of the SAS2 control module, which is denoted as SAS2H, has been made available with the release of SCALE-4. For each time-dependent fuel composition, SAS2H performs one-dimensional (1-D) neutron transport analyses (via XSDRNPM-S) of the reactor fuel assembly using a two-part procedure with two separate unit-cell-lattice models. The cross sections derived from a transport analysis at each time step are used in a point-depletion computation (via ORIGEN-S) that produces the burnup-dependent fuel composition to be used in the next spectral calculation. A final ORIGEN-S case is used to perform the complete depletion/decay analysis using the burnup-dependent cross sections. The techniques used by SAS2H and two recent applications of the code are reviewed in this paper. 17 refs., 5 figs., 5 tabs

DATING: A computer code for determining allowable temperatures for dry storage of spent fuel in inert and nitrogen gases

International Nuclear Information System (INIS)

Simonen, E.P.; Gilbert, E.R.

1988-12-01

The DATING (Determining Allowable Temperatures in Inert and Nitrogen Gases) code can be used to calculate allowable initial temperatures for dry storage of light-water-reactor spent fuel. The calculations are based on the life fraction rule using both measured data and mechanistic equations as reported by Chin et al. (1986). The code is written in FORTRAN and utilizes an efficient numerical integration method for rapid calculations on IBM-compatible personal computers. This report documents the technical basis for the DATING calculations, describes the computational method and code statements, and includes a user's guide with examples. The software for the DATING code is available through the National Energy Software Center operated by Argonne National Laboratory, Argonne, Illinois 60439. 5 refs., 8 figs., 5 tabs

Modern approach to the problem of fossil gas fuels replacement by alternative fuels

Energy Technology Data Exchange (ETDEWEB)

Soroka, Boris [Gas Institute, National Academy of Sciences, Kiev (Ukraine)

2013-07-01

New scientific and engineering fundamentals of fuels substitution have been developed instead of obsolete methodology “Interchangeability of Fuel Gases” developed in USA and existing from the middle of XX{sup th} century. To perform the complex prediction of total or partial substitution of given flow rate of natural gas NG for alternative gases AG the following parameters are to be predicted: plant utilization efficiencies – regarding fuel and energy utilization, the last in form of heat Ș{sub H} and exergy Ș{sub eff} efficiencies, saving or overexpenditure of the NG flow rate in the gas mixture with AG, specific fuel consumption b f and specific issue of harmful substances C{sub t} – pollutants in the combustion products (C{sub NO{sub x}} ) and greenhouse gases (C {sub CO{sub 2}} ). Certification of alternative gas fuels and fuel mixtures as a commodity products is carried out in frame of our approach with necessary set of characteristics, similar to those accepted in the world practice. Key words: alternative fuel, fuel replacement (substitution), natural gas, process gases, theoretical combustion temperature, thermodynamic equilibrium computations, total enthalpy.

On the effects of fuel properties and injection timing in partially premixed compression ignition of low octane fuels

KAUST Repository

Naser, Nimal

2017-06-29

A better understanding on the effects of fuel properties and injection timing is required to improve the performance of advanced engines based on low temperature combustion concepts. In this work, an experimental and computational study was conducted to investigate the effects of physical and chemical kinetic properties of low octane fuels and their surrogates in partially premixed compression ignition (PPCI) engines. The main objective was to identify the relative importance of physical versus chemical kinetic properties in predicting practical fuel combustion behavior across a range of injection timings. Two fuel/surrogate pairs were chosen for comparison: light naphtha (LN) versus the primary reference fuel (PRF) with research octane number of 65 (PRF 65), and FACE (fuels for advanced combustion engines) I gasoline versus PRF 70. Two sets of parametric studies were conducted: the first varied the amount of injected fuel mass at different injection timings to match a fixed combustion phasing, and the second maintained the same injected fuel mass at each injection timing to assess resulting combustion phasing changes. Full-cycle computational fluid dynamic engine simulations were conducted by accounting for differences in the physical properties of the original and surrogate fuels, while employing identical chemical kinetics. The simulations were found to capture trends observed in the experiments, while providing details on spatial mixing and chemical reactivity for different fuels and injection timings. It was found that differences in physical properties become increasingly important as injection timing was progressively delayed from premixed conditions, and this was rationalized by analysis of mixture stratification patterns resulting from injection of fuels with different physical properties. The results suggest that accurate descriptions of both physical and chemical behavior of fuels are critical in predictive simulations of PPCI engines for a wide range of

On the effects of fuel properties and injection timing in partially premixed compression ignition of low octane fuels

KAUST Repository

Naser, Nimal; Jaasim, Mohammed; Atef, Nour; Chung, Suk-Ho; Im, Hong G.; Sarathy, Mani

2017-01-01

A better understanding on the effects of fuel properties and injection timing is required to improve the performance of advanced engines based on low temperature combustion concepts. In this work, an experimental and computational study was conducted to investigate the effects of physical and chemical kinetic properties of low octane fuels and their surrogates in partially premixed compression ignition (PPCI) engines. The main objective was to identify the relative importance of physical versus chemical kinetic properties in predicting practical fuel combustion behavior across a range of injection timings. Two fuel/surrogate pairs were chosen for comparison: light naphtha (LN) versus the primary reference fuel (PRF) with research octane number of 65 (PRF 65), and FACE (fuels for advanced combustion engines) I gasoline versus PRF 70. Two sets of parametric studies were conducted: the first varied the amount of injected fuel mass at different injection timings to match a fixed combustion phasing, and the second maintained the same injected fuel mass at each injection timing to assess resulting combustion phasing changes. Full-cycle computational fluid dynamic engine simulations were conducted by accounting for differences in the physical properties of the original and surrogate fuels, while employing identical chemical kinetics. The simulations were found to capture trends observed in the experiments, while providing details on spatial mixing and chemical reactivity for different fuels and injection timings. It was found that differences in physical properties become increasingly important as injection timing was progressively delayed from premixed conditions, and this was rationalized by analysis of mixture stratification patterns resulting from injection of fuels with different physical properties. The results suggest that accurate descriptions of both physical and chemical behavior of fuels are critical in predictive simulations of PPCI engines for a wide range of

Advanced Fuel Cycle Cost Basis

Energy Technology Data Exchange (ETDEWEB)

D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

2009-12-01

This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

Advanced Fuel Cycle Cost Basis

Energy Technology Data Exchange (ETDEWEB)

D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert

2007-04-01

This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 26 cost modules—24 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, and high-level waste.

Advanced Fuel Cycle Cost Basis

Energy Technology Data Exchange (ETDEWEB)

D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

2008-03-01

This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

Thermal-hydraulic analysis of spent fuel storage systems

International Nuclear Information System (INIS)

Rector, D.R.; Wheeler, C.L.; Lombardo, N.J.

1987-01-01

This paper describes the COBRA-SFS (Spent Fuel Storage) computer code, which is designed to predict flow and temperature distributions in spent nuclear fuel storage and transportation systems. The decay heat generated by spent fuel in a dry storage cask is removed through a combination of conduction, natural convection, and thermal radiation. One major advantage of COBRA-SFS is that fluid recirculation within the cask is computed directly by solving the mass and momentum conservation equations. In addition, thermal radiation heat transfer is modeled using detailed radiation exchange factors based on quarter-rod segments. The equations governing mass, momentum, and energy conservation for incompressible flows are presented, and the semi-implicit solution method is described. COBRA-SFS predictions are compared to temperature data from a spent fuel storage cask test and the effect of different fill media on the cladding temperature distribution is discussed. The effect of spent fuel consolidation on cask thermal performance is also investigated. 16 refs., 6 figs., 2 tabs

Interim report spent nuclear fuel retrieval system fuel handling development testing

Energy Technology Data Exchange (ETDEWEB)

Ketner, G.L.; Meeuwsen, P.V.; Potter, J.D.; Smalley, J.T.; Baker, C.P.; Jaquish, W.R.

1997-06-01

Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project at the Hanford Site. The project will retrieve spent nuclear fuel, clean and remove fuel from canisters, repackage fuel into baskets, and load fuel into a multi-canister overpack (MCO) for vacuum drying and interim dry storage. The FRS is required to retrieve basin fuel canisters, clean fuel elements sufficiently of uranium corrosion products (or sludge), empty fuel from canisters, sort debris and scrap from whole elements, and repackage fuel in baskets in preparation for MCO loading. The purpose of fuel handling development testing was to examine the systems ability to accomplish mission activities, optimization of equipment layouts for initial process definition, identification of special needs/tools, verification of required design changes to support performance specification development, and validation of estimated activity times/throughput. The test program was set up to accomplish this purpose through cold development testing using simulated and prototype equipment; cold demonstration testing using vendor expertise and systems; and graphical computer modeling to confirm feasibility and throughput. To test the fuel handling process, a test mockup that represented the process table was fabricated and installed. The test mockup included a Schilling HV series manipulator that was prototypic of the Schilling Hydra manipulator. The process table mockup included the tipping station, sorting area, disassembly and inspection zones, fuel staging areas, and basket loading stations. The test results clearly indicate that the Schilling Hydra arm cannot effectively perform the fuel handling tasks required unless it is attached to some device that can impart vertical translation, azimuth rotation, and X-Y translation. Other test results indicate the importance of camera locations and capabilities, and of the jaw and end effector tool design. 5 refs., 35 figs., 3 tabs.

Development and application of methods and computer codes of fuel management and nuclear design of reload cycles in PWR

International Nuclear Information System (INIS)

Ahnert, C.; Aragones, J.M.; Corella, M.R.; Esteban, A.; Martinez-Val, J.M.; Minguez, E.; Perlado, J.M.; Pena, J.; Matias, E. de; Llorente, A.; Navascues, J.; Serrano, J.

1976-01-01

Description of methods and computer codes for Fuel Management and Nuclear Design of Reload Cycles in PWR, developed at JEN by adaptation of previous codes (LEOPARD, NUTRIX, CITATION, FUELCOST) and implementation of original codes (TEMP, SOTHIS, CICLON, NUDO, MELON, ROLLO, LIBRA, PENELOPE) and their application to the project of Management and Design of Reload Cycles of a 510 Mwt PWR, including comparison with results of experimental operation and other calculations for validation of methods. (author) [es

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

International Nuclear Information System (INIS)

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

1989-02-01

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

PWR fuel physico chemistry. Improvements of the Sage code to compute thermochemical balance in PWR fuel

International Nuclear Information System (INIS)

Garcia, P.; Baron, D.; Piron, J.P.

1993-02-01

A physicochemical survey of high burnup fuel has been undertaken in the context of a 3-party action (CEA Cadarache - EDF/DER - FRAMATOME). One of the tasks involved consists in adapting the SAGE code for assessment of the thermochemical equilibria of fission products in solution in the fuel matrix. This paper describes the first stage of this task. Even if other improvements are planned, the oxid oxygen potentials are yet properly reproduced for the simulated burnup. (authors). 63 figs., 4 tabs., 41 refs

Fuel pellet fracture and relocation

International Nuclear Information System (INIS)

Walton, L.A.; Husser, D.L.

1983-01-01

The model used to describe fuel pellet fracture and relocation is an important feature of a fuel performance computer code. This model becomes especially important if the computer code is principally to be used for the evaluation of pellet clad interaction. The fracture and relocation model being developed for the B and W fuel performance code FUMAC was derived from an extensive data base. Cross sections of irradiated fuel rods were photographically magnified and measured to determine the configuration of the fragments of the fractured fuel pellets. Data, representing a wide range of LWR fuel designs and as-manufactured mechanical configurations, were catalogued and systematically reduced and then correlated as a function of the likely independent variables. These correlations define the key phenomenological behavior patterns which the relocation model must duplicate and indicate which mechanistic approaches are viable explanations of this behavior. The data base covers the burnup range from approximately one to 35 GWd/mtU and linear heat rates from less than 100 to nearly 700 W/Cm. This paper presents the correlated data base and the methods used to derive and interpret it. It was determined from this data base that pellet cracking is initially both power level and burnup dependent but tends to saturate eventually with continued steady irradiation. Fuel pellet relocation was found to be much more extensive than would be deduced from thermal considerations alone. Even at very low burnups fuel fragments were found to move outward until restrained by the cladding. The results also suggest that changes in internal resistance to heat flow within the pellets due to the opening of cracks may be as important as peripheral gap changes to the thermal modeler. The transient response and thermal implications of this model are recommended as primary areas for future investigation

Fuel channel refilling

International Nuclear Information System (INIS)

Shoukri, M.; Abdul-Razzak, A.

1990-04-01

Analysis of existing data on fuel channel refilling is presented. The analysis focuses on the data obtained using the Stern Laboratories Cold Water Injection Test (CWIT) Facility. The two-fluid model thermal-hydraulics computer code CATHENA is also used to simulate experimental results on fuel channel refilling from both the CWIT and RD-14 facilities. Conclusions related to single and double break tests, including the effect of non-condensible gases, are presented. A set of recommendations is given for further analysis and separate effect experiments. (67 figs., 5 tabs., 24 refs.)

Uncertainty Analyses of Advanced Fuel Cycles

International Nuclear Information System (INIS)

Miller, Laurence F.; Preston, J.; Sweder, G.; Anderson, T.; Janson, S.; Humberstone, M.; MConn, J.; Clark, J.

2008-01-01

The Department of Energy is developing technology, experimental protocols, computational methods, systems analysis software, and many other capabilities in order to advance the nuclear power infrastructure through the Advanced Fuel Cycle Initiative (AFDI). Our project, is intended to facilitate will-informed decision making for the selection of fuel cycle options and facilities for development

Uncertainty Analyses of Advanced Fuel Cycles

Energy Technology Data Exchange (ETDEWEB)

Laurence F. Miller; J. Preston; G. Sweder; T. Anderson; S. Janson; M. Humberstone; J. MConn; J. Clark

2008-12-12

The Department of Energy is developing technology, experimental protocols, computational methods, systems analysis software, and many other capabilities in order to advance the nuclear power infrastructure through the Advanced Fuel Cycle Initiative (AFDI). Our project, is intended to facilitate will-informed decision making for the selection of fuel cycle options and facilities for development.

Development and validation of a fuel performance analysis code

International Nuclear Information System (INIS)

Majalee, Aaditya V.; Chaturvedi, S.

2015-01-01

CAD has been developing a computer code 'FRAVIZ' for calculation of steady-state thermomechanical behaviour of nuclear reactor fuel rods. It contains four major modules viz., Thermal module, Fission Gas Release module, Material Properties module and Mechanical module. All these four modules are coupled to each other and feedback from each module is fed back to others to get a self-consistent evolution in time. The computer code has been checked against two FUMEX benchmarks. Modelling fuel performance in Advance Heavy Water Reactor would require additional inputs related to the fuel and some modification in the code.(author)

Numerical simulator of the CANDU fueling machine driving desk

International Nuclear Information System (INIS)

Doca, Cezar

2008-01-01

As a national and European premiere, in the 2003 - 2005 period, at the Institute for Nuclear Research Pitesti two CANDU fueling machine heads, no.4 and no.5, for the Nuclear Power Plant Cernavoda - Unit 2 were successfully tested. To perform the tests of these machines, a special CANDU fueling machine testing rig was built and was (and is) available for this goal. The design of the CANDU fueling machine test rig from the Institute for Nuclear Research Pitesti is a replica of the similar equipment operating in CANDU 6 type nuclear power plants. High technical level of the CANDU fueling machine tests required the using of an efficient data acquisition and processing Computer Control System. The challenging goal was to build a computer system (hardware and software) designed and engineered to control the test and calibration process of these fuel handling machines. The design takes care both of the functionality required to correctly control the CANDU fueling machine and of the additional functionality required to assist the testing process. Both the fueling machine testing rig and staff had successfully assessed by the AECL representatives during two missions. At same the time, at the Institute for Nuclear Research Pitesti was/is developed a numerical simulator for the CANDU fueling machine operators training. The paper presents the numerical simulator - a special PC program (software) which simulates the graphics and the functions and the operations at the main desk of the computer control system. The simulator permits 'to drive' a CANDU fueling machine in two manners: manual or automatic. The numerical simulator is dedicated to the training of operators who operate the CANDU fueling machine in a nuclear power plant with CANDU reactor. (author)

Harnessing the Power of Intellectual Capital.

Science.gov (United States)

Bassi, Laurie J.

1997-01-01

Describes intellectual capital--employees' brainpower, know-how, knowledge, and processes--and knowledge management--the processes by which a company creates and leverages intellectual capital--as the primary sources of competitive advantage in many industries. Offers ways to measure intellectual capital, a glossary, and additional resources. (JOW)

Programs for Students and Teachers | NREL

Science.gov (United States)

competition that tests the brainpower of middle and high school teams on science and math topics. Model Car and Education Programs promote science, technology, engineering, and mathematics (STEM) using Science educators provide hands-on experiences for students in grades 4-12 to learn about renewable energy

Analyses with the FSTATE code: fuel performance in destructive in-pile experiments

International Nuclear Information System (INIS)

Bauer, T.H.; Meek, C.C.

1982-01-01

Thermal-mechanical analysis of a fuel pin is an essential part of the evaluation of fuel behavior during hypothetical accident transients. The FSTATE code has been developed to provide this required computational ability in situations lacking azimuthal symmetry about the fuel-pin axis by performing 2-dimensional thermal, mechanical, and fission gas release and redistribution computations for a wide range of possible transient conditions. In this paper recent code developments are described and application is made to in-pile experiments undertaken to study fast-reactor fuel under accident conditions. Three accident simulations, including a fast and slow ramp-rate overpower as well as a loss-of-cooling accident sequence, are used as representative examples, and the interpretation of STATE computations relative to experimental observations is made

Fuel-to-cladding heat transfer coefficient into reactor fuel element

International Nuclear Information System (INIS)

Lassmann, K.

1979-01-01

Models describing the fuel-to-cladding heat transfer coefficient in a reactor fuel element are reviewed critically. A new model is developed with contributions from solid, fluid and radiation heat transfer components. It provides a consistent description of the transition from an open gap to the contact case. Model parameters are easily available and highly independent of different combinations of material surfaces. There are no restrictions for fast transients. The model parameters are fitted to 388 data points under reactor conditions. For model verification another 274 data points of steel-steel and aluminium-aluminium interfaces, respectively, were used. The fluid component takes into account peak-to-peak surface roughnesses and, approximatively, also the wavelengths of surface roughnesses. For minor surface roughnesses normally prevailing in reactor fuel elements the model asymptotically yields Ross' and Stoute's model for the open gap, which is thus confirmed. Experimental contact data can be interpreted in very different ways. The new model differs greatly from Ross' and Stoute's contact term and results in better correlation coefficients. The numerical algorithm provides an adequate representation for calculating the fuel-to-cladding heat transfer coefficient in large fuel element structural analysis computer systems. (orig.) [de

CANDU-PHW fuel management

International Nuclear Information System (INIS)

Frescura, G.M.; Wight, A.L.

1982-01-01

This report covers the material presented in a series of six lectures at the Winter College on Nuclear Physics and Reactors held at the International Centre for Theoretical Physics in Trieste, Italy, Jan 22 - March 28, 1980. The report deals with fuel management in natural uranium fuelled CANDU-PHW reactors. Assuming that the reader has a basic knowledge of CANDU core physics, some of the reactor systems which are more closely related to fuelling are described. This is followed by a discussion of the methods used to calculate the power distribution and perform fuel management analyses for the equilibrium core. A brief description of some computer codes used in fuel management is given, together with an overview of the calculations required to provide parameters for core design and support the accident analysis. Fuel scheduling during approach to equilibrium and equilibrium is discussed. Fuel management during actual reactor operation is discussed with a review of the operating experience for some of the Ontario Hydro CANDU reactors. (author)

Efficient characterization of fuel depletion in boiling water reactor

International Nuclear Information System (INIS)

Kim, S.H.

1980-01-01

An efficient fuel depletion method for boiling water reactor (BWR) fuel assemblies has been developed for fuel cycle analysis. A computer program HISTORY based on this method was designed to carry out accurate and rapid fuel burnup calculation for the fuel assembly. It has been usefully employed to study the depletion characteristics of the fuel assemblies for the preparation of nodal code input data and the fuel management study. The adequacy and the effectiveness of the assessment of this method used in HISTORY were demonstrated by comparing HISTORY results with more detailed CASMO results. The computing cost of HISTORY typically has been less than one dollar for the fuel assembly-level depletion calculations over the full life of the assembly, in contrast to more than $1000 for CASMO. By combining CASMO and HISTORY, a large number of expensive CASMO calculations can be replaced by inexpensive HISTORY. For the depletion calculations via CASMO/HISTORY, CASMO calculations are required only for the reference conditions and just at the beginning of life for other cases such as changes in void fraction, control rod condition and temperature. The simple and inexpensive HISTORY is sufficienty accurate and fast to be used in conjunction with CASMO for fuel cycle analysis and some BWR design calculations

ADELA - user interface for fuel charge design

International Nuclear Information System (INIS)

Havluj, Frantisek

2010-01-01

ADELA is a supporting computer code - ANDREA code add-on - for fuel batch designing and optimization. It facilitates fuel batch planning, evaluation and archival by using graphical user interface. ADELA simplifies and automates the design process and is closely linked to the QUADRIGA system for data library creation. (author)

An introduction to current modeling techniques in nuclear fuel performance analysis

International Nuclear Information System (INIS)

Faya, S.C.S.

1981-10-01

An overview of several important phenomena considered in fuel performance analysis, such as fuel restructuring, fuel swelling, fission gas release, pellet cracking, etc... are provided. All these phenomena are independent. For the purpose of illustration of integral fuel rod analysis the LIFE-1 computer code is discussed. (Author) [pt

Operational experience using the OSTR flip fuel self-protection program

International Nuclear Information System (INIS)

Dodd, B.; Ringle, J.C.; Anderson, T.V.; Johnson, A.G.

1982-01-01

Recent changes in NRC Physical Security regulations make it highly desirable for a small number of TRIGA research reactor establishments to maintain each of the fuel elements in their reactor core above the self-protection dose rate criterion. OSTR operations personnel have written a computer program (SPOOF) which calculates the exposure rate (in Rhr -1 ) from an irradiated fuel element at 3 feet in air using the actual operating history of the reactor. The purpose of this current paper is to describe the operational experience gained over the last year and a half while using the SPOOF computer program, and while performing the quarterly dose rate measurements needed to confirm the continuing accuracy of the program, and, most importantly, the self-protection status of the OSTR fuel. The computer program in association with the quarterly dose rate measurements have been accepted by the NRC, and allow the OSTR to take credit for self-protecting FLIP fuel under the current physical security regulations

Measurement of the fuel temperature and the fuel-to-coolant heat transfer coefficient of Super Phenix 1 fuel elements

International Nuclear Information System (INIS)

Edelmann, M.

1995-12-01

A new measurement method for measuring the mean fuel temperature as well as the fuel-to-coolant heat transfer coefficient of fast breeder reactor subassemblies (SA) is reported. The method is based on the individual heat balance of fuel SA's after fast reactor shut-downs and uses only the plants normal SA outlet temperature and neutron power signals. The method was used successfully at the french breeder prototype Super Phenix 1. The mean SA fuel temperature as well as the heat transfer coefficient of all SPX SA's have been determined at power levels between 15 and 90% of nominal power and increasing fuel burn-up from 3 to 83 EFPD (Equivalent of Full Power-Days). The measurements also provided fuel and whole SA time constants. The estimated accuracy of measured fuel parameters is in the order of 10%. Fuel temperatures and SA outlet temperature transients were also calculated with the SPX1 systems code DYN2 for exactly the same fuel and reactor operating parameters as in the experiments. Measured fuel temperatures were higher than calculated ones in all cases. The difference between measured and calculated core mean values increases from 50 K at low power to 180 K at 90% n.p. This is about the double of the experimental error margins. Measured SA heat transfer coefficients are by nearly 20% lower than corresponding heat transfer parameters used in the calculations. Discrepancies found between measured and calculated results also indicate that either the transient heat transfer in the gap between fuel and cladding (gap conductance) might not be exactly reproduced in the computer code or that the gap in the fresh fuel was larger than assumed in the calculations. (orig.) [de

Automated manufacturing of breeder reactor fuels

International Nuclear Information System (INIS)

Nyman, D.H.; Benson, E.M.; Bennett, D.W.

1983-09-01

The Secure Automated Fabrication (SAF) line is an automated, remotely controlled breeder fuel pin fabrication process which is to be installed in the Fuels and Materials Examination Facility (FMEF). The FMEF is presently under construction at Hanford and is scheduled for completion in 1984. The SAF line is scheduled for startup in 1987 and will produce mixed uranium-plutonium oxide fuel pins for the Fast Flux Test Facility (FFTF). Radiological protection requirements, computer control equipment, use of robotics, and the fabrication process is described

The Information Age?

Science.gov (United States)

MacVicar, Margaret L. A.

1985-01-01

The author examines the concept of information and how it relates to our future. She discusses the various waves of information about education: the decay of technology and the raid on educational institutions for brainpower made by business and industry, foreign competition, and the quality of education. (CT)

Determination of the vibration characteristics of nuclear fuel rods in a fluid flow using multiphysics computation

International Nuclear Information System (INIS)

Sbragio, Ricardo

1999-01-01

The determination of natural frequencies and displacement Power Spectrum Density (PSD) of fuel rods in a fluid using Computational Fluid Dynamics and Finite Element Methods is presented. The rods are modeled as slender beams subjected to small displacements in a fluid using three-dimensional mesh. The incompressible Navier-Stokes and linear momentum balance equations are solved simultaneously using Spectrum code. Two examples from literature are analyzed. The first consists in one rod in a fluid. The excitation is an impulse force at the rod central node. The second example is a two rod system in a fluid. In this case, the excitation force is random and is determined from a PSD. (author)

Pre-irradiation testing of experimental fuel elements

International Nuclear Information System (INIS)

Basova, B.G.; Davydov, E.F.; Dvoretskij, V.G.; Ivanov, V.B.; Syuzev, V.N.; Timofeev, G.A.; Tsykanov, V.A.

1979-01-01

The problems of testing of experimental fuel elements of nuclear reactors on the basis of complex accountancy of the factors defining operating capacity of the fuel elements are considered. The classification of the parameters under control and the methods of initial technological testing, including testing of the fuel product, cladding and fished fuel element, is given. The requirements to the apparatus used for complex testing are formulated. One of the possible variants of representation of the information obtained in the form of the input certificate of a single fuel element under study is proposed. The processing flowsheet of the gathered information using the computer is given. The approach under consideration is a methodological basis of investigation of fuel element operating life at the testing stage of the experimental fuel elements

The lumped parameter model for fuel pins

Energy Technology Data Exchange (ETDEWEB)

Liu, W S [Ontario Hydro, Toronto, ON (Canada)

1996-12-31

The use of a lumped fuel-pin model in a thermal-hydraulic code is advantageous because of computational simplicity and efficiency. The model uses an averaging approach over the fuel cross section and makes some simplifying assumptions to describe the transient equations for the averaged fuel, fuel centerline and sheath temperatures. It is shown that by introducing a factor in the effective fuel conductivity, the analytical solution of the mean fuel temperature can be modified to simulate the effects of the flux depression in the heat generation rate and the variation in fuel thermal conductivity. The simplified analytical method used in the transient equation is presented. The accuracy of the lumped parameter model has been compared with the results from the finite difference method. (author). 4 refs., 2 tabs., 4 figs.

Neutron-photon energy deposition in CANDU reactor fuel channels: a comparison of modelling techniques using ANISN and MCNP computer codes

International Nuclear Information System (INIS)

Bilanovic, Z.; McCracken, D.R.

1994-12-01

In order to assess irradiation-induced corrosion effects, coolant radiolysis and the degradation of the physical properties of reactor materials and components, it is necessary to determine the neutron, photon, and electron energy deposition profiles in the fuel channels of the reactor core. At present, several different computer codes must be used to do this. The most recent, advanced and versatile of these is the latest version of MCNP, which may be capable of replacing all the others. Different codes have different assumptions and different restrictions on the way they can model the core physics and geometry. This report presents the results of ANISN and MCNP models of neutron and photon energy deposition. The results validate the use of MCNP for simplified geometrical modelling of energy deposition by neutrons and photons in the complex geometry of the CANDU reactor fuel channel. Discrete ordinates codes such as ANISN were the benchmark codes used in previous work. The results of calculations using various models are presented, and they show very good agreement for fast-neutron energy deposition. In the case of photon energy deposition, however, some modifications to the modelling procedures had to be incorporated. Problems with the use of reflective boundaries were solved by either including the eight surrounding fuel channels in the model, or using a boundary source at the bounding surface of the problem. Once these modifications were incorporated, consistent results between the computer codes were achieved. Historically, simple annular representations of the core were used, because of the difficulty of doing detailed modelling with older codes. It is demonstrated that modelling by MCNP, using more accurate and more detailed geometry, gives significantly different and improved results. (author). 9 refs., 12 tabs., 20 figs

A comparative study of fuel management in PWR reactors

International Nuclear Information System (INIS)

Barroso, D.E.G.

1980-01-01

A comparitive study of fuel recycling in Pressurized Water Reactors was developed, considering not only the conventional uranium cycle, but also the use of thorium as an alternative. The use of thorium was done by varying its conoentration in the homogeneous mixture with uranium in the fuel from 30% up to 90%. The U-233 produced is incorporated within the isotopic composition of irradiated uranium. Various fractions of irradiated recycled fuel to be reprocessed and recycled was considered. Various alternatives of recycling were outlined and a final comparison in the tests done, is furnished in terms of U 3 O 8 and UTS requirements and approximated costs of fuel cycle stages involved. The recycled fuel is considered to be uniformly distributed in the fuel element rods introduced in the nucleus. The influence of the utilization of thorium was also considered for the development of an optimum fuel cycle, regarding the safeguards against nuclear proliferation when utilizing plutonium. A zero-dimensional cellular model was adopted to represent the reactor and the calculus of microscopic cross-sections for the homogenized cell was done by the computer code LEOPARD. A digital computer program was develped for neutronic and fuel depletion calculus and to simulate the refueling of various cycles. (Author) [pt

Practical constraints on fuel management a utility perspective

International Nuclear Information System (INIS)

Grier, C.A.

1986-01-01

The practical and potential constraints of performing fuel management at a large utility are reviewed. Based on approximately six years of experience in performing fuel management the constraints due to commercial, technical, utility system, design methods, and personnel and computer resources are discussed in detail

Fuel cycle management in Finland

International Nuclear Information System (INIS)

Vaeyrynen, H.; Mikkola, I.

1987-01-01

Both Finnish utilities producing nuclear power - Imatran Voima Oy (IVO) and Teollisuuden Voima Oy (Industrial Power Co. Ltd, TVO) - have created efficient fuel cycle management systems. The systems however differ in almost all respects. The reason is that the principal supplier for IVO is the Soviet Union and for TVO is Sweden. A common feature of both systems at the front end of the cycle is the building of stockpiles in order to provide for interruptions in fuel deliveries. Quality assurance supervision at the fuel factory for IVO is regulated by the Soviet Chamber of Commerce and Industry and a final control is made in Finland. The in-core fuel management is done by IVO using codes developed in Finland. The whole IVO fuel cycle is basically a leasing arrangement. The spent fuel is returned to the USSR after five years cooling. TVO carries out the in-core fuel management using a computer code system supplied by Asea-Atom. TVO is responsable for the back end of the cycle and makes preparations for the final disposal of the spent fuel in Finland. 6 refs., 2 figs

Fuels planning: science synthesis and integration; forest structure and fire hazard fact sheet 03: visualizing forest structure and fuels

Science.gov (United States)

Rocky Mountain Research Station USDA Forest Service

2004-01-01

The software described in this fact sheet provides managers with tools for visualizing forest and fuels information. Computer-based landscape simulations can help visualize stand and landscape conditions and the effects of different management treatments and fuel changes over time. These visualizations can assist forest planning by considering a range of management...

FEMAXI-III, a computer code for fuel rod performance analysis

International Nuclear Information System (INIS)

Ito, K.; Iwano, Y.; Ichikawa, M.; Okubo, T.

1983-01-01

This paper presents a method of fuel rod thermal-mechanical performance analysis used in the FEMAXI-III code. The code incorporates the models describing thermal-mechanical processes such as pellet-cladding thermal expansion, pellet irradiation swelling, densification, relocation and fission gas release as they affect pellet-cladding gap thermal conductance. The code performs the thermal behavior analysis of a full-length fuel rod within the framework of one-dimensional multi-zone modeling. The mechanical effects including ridge deformation is rigorously analyzed by applying the axisymmetric finite element method. The finite element geometrical model is confined to a half-pellet-height region with the assumption that pellet-pellet interaction is symmetrical. The 8-node quadratic isoparametric ring elements are adopted for obtaining accurate finite element solutions. The Newton-Raphson iteration with an implicit algorithm is applied to perform the analysis of non-linear material behaviors accurately and stably. The pellet-cladding interaction mechanism is exactly treated using the nodal continuity conditions. The code is applicable to the thermal-mechanical analysis of water reactor fuel rods experiencing variable power histories. (orig.)

Flow field measurements using LDA and numerical computation for rod bundle of reactor fuel assembly

International Nuclear Information System (INIS)

Hu Jun; Zou Zunyu

1995-02-01

Local mean velocity and turbulence intensity measurements were obtained with DANTEC 55 X two-dimensional Laser Dopper Anemometry (LDA) for rod bundle of reactor fuel assembly test model which was a 4 x 4 rod bundle. The data were obtained from different experimental cross-sections both upstream and downstream of the model support plate. Measurements performed at test Reynolds numbers of 1.8 x 10 4 ∼3.6 x 10 4 . The results described the local and gross effects of the support plate on upstream and downstream flow distributions. A numerical computation was also given, the experimental results are in good agreement with the numerical one and the others in references. Finally, a few suggestions were proposed for how to use the LDA system well. (11 figs.)

Recent BWR fuel management reactor physics advances

International Nuclear Information System (INIS)

Crowther, R.L.; Congdon, S.P.; Crawford, B.W.; Kang, C.M.; Martin, C.L.; Reese, A.P.; Savoia, P.J.; Specker, S.R.; Welchly, R.

1982-01-01

Improvements in BWR fuel management have been under development to reduce uranium and separative work (SWU) requirements and reduce fuel cycle costs, while also maintaining maximal capacity factors and high fuel reliability. Improved reactor physics methods are playing an increasingly important role in making such advances feasible. The improved design, process computer and analysis methods both increase knowledge of the thermal margins which are available to implement fuel management advance, and improve the capability to reliably and efficiently analyze and design for fuel management advances. Gamma scan measurements of the power distributions of advanced fuel assembly and advanced reactor core designs, and improved in-core instruments also are important contributors to improving 3-d predictive methods and to increasing thermal margins. This paper is an overview of the recent advances in BWR reactor physics fuel management methods, coupled with fuel management and core design advances. The reactor physics measurements which are required to confirm the predictions of performance fo fuel management advances also are summarized

The Development Of A Computer Program For Thermohydraulic Analysis Of Kartini Reactor

International Nuclear Information System (INIS)

Ischaq, Ma'sum; Syarip; BS, Edi Trijono; Suyamto

1996-01-01

A computer programming that could be used to calculate the fuel temperature and the reactor core, has been developed. The inside fuel temperature was calculated by explicit method. First, the differential partial equations were arranged the temperature as a function of the fuel radius and lime. T = f(r.t), and then the equations were transformed into the finite difference equations that could be solved numerically by the computer. The convection heal transfer coefficient between the fuel and the coolant was calculated basically by the free convection phenomena that followed the equation Nu = f (Gr, Pr). By this computer programming, the fuel and the core temperature in a certain condition of the reactor power and the fluid could be predicted

Commit to Building on New England's "Employability" Advantage!

Science.gov (United States)

Gittell, Ross

2017-01-01

Higher education has provided New England with an economic advantage, as the region without strong natural resource advantages has relied on its higher education institutions (HEIs) and brainpower. A higher education-based economic advantage has enabled the region to develop strong well-paying technology and knowledge-based industries tied to New…

Studies on modeling to failed fuel detection system response in LMFBR

International Nuclear Information System (INIS)

Miyazawa, T.; Saji, G.; Mitsuzuku, N.; Hikichi, T.; Odo, T.; Rindo, H.

1981-05-01

Failed Fuel Detection (FFD) system with Fission Products (FP) detection is considered to be the most promissing method, since FP provides direct information against fuel element failure. For designing FFD system and for evaluating FFD signals, some adequate FFD signal response to fuel failure have been required. But few models are available in nowadays. Thus Power Reactor and Nuclear Fuel Development Corporation (PNC) had developed FFD response model with computer codes, based on several fundamental investigations on FP release and FP behavior, and referred to foreign country experiences on fuel failure. In developing the model, noble gas and halogen FP release and behavior were considered, since FFD system would be composed of both cover gas monitoring and delayed neutron monitoring. The developed model can provide typical fuel failure response and detection limit which depends on various background signals at cover gas monitoring and delayed neutron monitoring. According to the FFD response model, we tried to assume fuel failure response and detection limit at Japan experimental fast reactor ''JOYO''. The detection limit of JOYO FFD system was estimated by measuring the background signals. Followed on the studies, a complete computer code has been now made with some improvement. On the paper, the details of the model, out line of developed computer code, status of JOYO FFD system, and trial assumption of JOYO FFD response and detection limit. (author)

Development of a Computer Program for the Integrated Control of the Fuel Homogeneity Measurement System

International Nuclear Information System (INIS)

Shin, H. S.; Jang, J. W.; Lee, Y. H.; Oh, S. J.; Park, H. D.; Kim, C. K.

2005-11-01

The computer program is developed based on Visual C++, which is equipped with a user-friendly interface of the input/output(I/O) and a display function for the measuring conditions. This program consists of three parts which are the port communication, PLC(Programmable Logic Controller) and the MCA(Multi Channel Analyzer) control parts. The communication type between the CPU of the PLC module box and the computer is selected as be the Rs-232 asynchronous type and the thread method is adapted in the development of the first part of the program. The PLC-related program has been developed so that the data communication between the PLC CPU and the computer could be harmonized with the unique commands which have already been defined in the PLC. The measuring space and time intervals, the start and end ROI(region of interest) values, and the allowable error limitation are input at each measurement in this program. Finally the controlling MCA program has been developed by using Canberra's programming library which contains several files including the head files in which the variable and the function of C++ are declared according to the MCA function. The performance test has been carried out through an application of the developed computer program to the homogeneity measurement system. The gamma counts at 28 measuring points of a fuel rod of 700 mm in length are measured for 50 sec at each point. It was revealed that the measurement results are better than the previous ones in respects of the measurement accuracy and a measurement time saving could be achieved. It was concluded that the gamma measurement system can be improved through equipping it with the developed control program

Micromechanical modelling of fuel viscoplastic behaviour

International Nuclear Information System (INIS)

Masson, R.; Blanc, V.; Gatt, J.M.; Julien, J.; Michel, B.; Largenton, R.

2015-01-01

To identify the effect of microstructural parameters on the viscoplastic behaviour of nuclear fuels, micromechanical (also called homogenisation) approaches are used. These approaches aim at deriving effective properties of heterogeneous material from the properties of their constituents. They stand on full-field computations of representative volume elements of microstructures as well as on mean-field semi-analytical models. For light water reactor fuels, these approaches have been applied to the modelling of the effect of two microstructural parameters: the porosity effects on the thermal creep of dioxide uranium fuels (transient conditions of irradiation) as well as the plutonium content effect on the viscoplastic behaviour (nominal conditions of irradiations) of mixed oxide fuels (MOX). (authors)

Rapid scanning system for fuel drawers

International Nuclear Information System (INIS)

Caldwell, J.T.; Fehlau, P.E.; France, S.W.

1981-01-01

A nondestructive method for uniqely distinguishing among and quantifying the mass of individual fuel plates in situ in fuel drawers utilized in nuclear reactors is described. The method is both rapid and passive, eliminating the personnel hazard of the commonly used irradiation techniques which require that the analysis be performed in proximity to an intense neutron source such as a reactor. In the present technique, only normally decaying nuclei are observed. This allows the analysis to be performed anywhere. This feature, combined with rapid scanning of a given fuel drawer (in approximately 30 s), and the computer data analysis allows the processing of large numbers of fuel drawers efficiently in the event of a loss alert

Models for fuel rod behaviour at high burnup

Energy Technology Data Exchange (ETDEWEB)

Jernkvist, Lars O.; Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park, Uppsala (Sweden)

2004-12-01

This report deals with release of fission product gases and irradiation-induced restructuring in uranium dioxide nuclear fuel. Waterside corrosion of zirconium alloy clad tubes to light water reactor fuel rods is also discussed. Computational models, suitable for implementation in the FRAPCON-3.2 computer code, are proposed for these potentially life-limiting phenomena. Hence, an integrated model for the calculation or thermal fission gas release by intragranular diffusion, gas trapping in grain boundaries, irradiation-induced re-solution, grain boundary saturation, and grain boundary sweeping in UO{sub 2} fuel, under time varying temperature loads, is formulated. After a brief review of the status of thermal fission gas release modelling, we delineate the governing equations for the aforementioned processes. Grain growth kinetic modelling is briefly reviewed and pertinent data on grain growth of high burnup fuel obtained during power ramps in the Third Risoe Fission Gas Release Project are evaluated. Sample computations are performed, which clearly show the connection between fission gas release and gram growth as a function of time at different isotherms. Models are also proposed for the restructuring of uranium dioxide fuel at high burnup, the so-called rim formation, and its effect on fuel porosity build-up, fuel thermal conductivity and fission gas release. These models are assessed by use of recent experimental data from the High Burnup Rim Project, as well as from post irradiation examinations of high-burnup fuel, irradiated in power reactors. Moreover, models for clad oxide growth and hydrogen pickup in PWRs, applicable to Zircaloy-4, ZIRLO or M5 cladding, are formulated, based on recent in-reactor corrosion data for high-burnup fuel rods. Our evaluation of these data indicates that the oxidation rate of ZIRLO-type materials is about 20% lower than for standard Zircaloy-4 cladding under typical PWR conditions. Likewise, the oxidation rate of M5 seems to be

Automated Fuel Element Closure Welding System

International Nuclear Information System (INIS)

Wahlquist, D.R.

1993-01-01

The Automated Fuel Element Closure Welding System is a robotic device that will load and weld top end plugs onto nuclear fuel elements in a highly radioactive and inert gas environment. The system was developed at Argonne National Laboratory-West as part of the Fuel Cycle Demonstration. The welding system performs four main functions, it (1) injects a small amount of a xenon/krypton gas mixture into specific fuel elements, and (2) loads tiny end plugs into the tops of fuel element jackets, and (3) welds the end plugs to the element jackets, and (4) performs a dimensional inspection of the pre- and post-welded fuel elements. The system components are modular to facilitate remote replacement of failed parts. The entire system can be operated remotely in manual, semi-automatic, or fully automatic modes using a computer control system. The welding system is currently undergoing software testing and functional checkout

Growth of the interaction layer around fuel particles in dispersion fuel

International Nuclear Information System (INIS)

Olander, D.

2009-01-01

Corrosion of uranium particles in dispersion fuel by the aluminum matrix produces interaction layers (an intermetallic-compound corrosion product) around the shrinking fuel spheres. The rate of this process was modeled as series resistances due to Al diffusion through the interaction layer and reaction of aluminum with uranium in the fuel particle to produce UAl x . The overall kinetics are governed by the relative rates of these two steps, the slowest of which is reaction at the interface between Al in the interaction layer and U in the fuel particle. The substantial volume change as uranium is transferred from the fuel to the interaction layer was accounted for. The model was compared to literature data on in-reactor growth of the interaction layer and the Al/U gradient in this layer, the latter measured in ex-reactor experiments. The rate constant of the Al-U interface reaction and the diffusivity of Al in the interaction layer were obtained from this fitting procedure. The second feature of the corrosion process is the transfer of fission products from the fuel particle to the interaction layer due to the reaction. It is commonly assumed that the observed swelling of irradiated fuel elements of this type is due to release of fission gas in the interaction layer to form large bubbles. This hypothesis was tested by using the model to compute the quantity of fission gas available from this source and comparing the pressure of the resulting gas with the observed swelling of fuel plates. It was determined that the gas pressure so generated is too small to account for the observed delamination of the fuel

On techniques of ATR lattice computation

International Nuclear Information System (INIS)

1997-08-01

Lattice computation is to compute the average nuclear constants of unit fuel lattice which are required for computing core nuclear characteristics such as core power distribution and reactivity characteristics. The main nuclear constants are infinite multiplying rate, neutron movement area, cross section for diffusion computation, local power distribution and isotope composition. As for the lattice computation code, WIMS-ATR is used, which is based on the WIMS-D code developed in U.K., and for the purpose of heightening the accuracy of analysis, which was improved by adding heavy water scattering cross section considering the temperature dependence by Honeck model. For the computation of the neutron absorption by control rods, LOIEL BLUE code is used. The extrapolation distance of neutron flux on control rod surfaces is computed by using THERMOS and DTF codes, and the lattice constants of adjoining lattices are computed by using the WIMS-ATR code. As for the WIMS-ATR code, the computation flow and nuclear data library, and as for the LOIEL BLUE code, the computation flow are explained. The local power distribution in fuel assemblies determined by the WIMS-ATR code was verified with the measured data, and the results are reported. (K.I.)

Advances in PEM fuel cells with CFD techniques

Energy Technology Data Exchange (ETDEWEB)

Robalinho, Eric; Cunha, Edgar Ferrari da; Zararya, Ahmed; Linardi, Marcelo [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)], Email: eric@ipen.br; Cekinski, Efrain [Instituto de Pesquisas Tecnologicas (IPT), Sao Paulo, SP (Brazil)

2010-07-01

This paper presents some applications of computational fluid dynamics techniques in the optimization of Proton Exchange Membrane Fuel Cell (PEMFC) designs. The results concern: modeling of gas distribution channels, the study for both porous anode and cathode and the three-dimensional modeling of a partial geometry layer containing catalytic Gas Diffusion Layers (GDL) and membrane. Numerical results of the simulations of graphite plates flow channels, using ethanol as fuel, are also presented. Some experimental results are compared to the corresponding numerical ones for several cases, demonstrating the importance and usefulness of this computational tool. (author)

Elastic plastic analysis of fuel element assemblies - hexagonal claddings and fuel rods

International Nuclear Information System (INIS)

Mamoun, M.M.; Wu, T.S.; Chopra, P.S.; Rardin, D.C.

1979-01-01

Analytical studies have been conducted to investigate the structural, thermal, and mechanical behavior of fuel rods, claddings and fuel element assemblies of several designs for a conceptual Safety Test Facility (STF). One of the design objectives was to seek a geometrical configuration for a clad by maximizing the volume fraction of fuel and minimizing the resultant stresses set-up in the clad. The results of studies conducted on various geometrical configurations showed that the latter design objective can be achieved by selecting a clad of an hexagonal geometry. The analytical studies necessitated developing solutions for determining the stresses, strains, and displacements experienced by fuel rods and an hexagonal cladding subjected to thermal fuel-bowing loads acting on its internal surface, the external pressure of the coolant, and elevated temperatures. This paper presents some of the initially formulated analytical methods and results. It should be emphasized that the geometrical configuration considered in this paper may not necessarily be similar to that of the final design. Several variables have been taken into consideration including cladding thickness, the dimensions of the fuel rod, the temperature of the fuel and cladding, the external pressure of the cooling fluid, and the mechanical strength properties of fuel and cladding. A finite-element computer program, STRAW Code, has also been employed to generate several numerical results which have been compared with those predicted by employing the initially formulated solutions. The theoretically predicted results are in good agreement with those of the STRAW Code. (orig.)

Special equipment for the fabrication and quality control of nuclear fuel elements

International Nuclear Information System (INIS)

Guse, K.; Herbert, W.; Jaeger, K.

1989-01-01

For the fabrication of LWR fuel elements, columns are packed of up to 4 m in length, consisting of fuel pellets with different uranium enrichment, their weight and total length to be measured prior to further processing to fuel rods. An automated column packing device has been developed for this purpose. The packing jobs and other tasks are computer-controlled, measured data are stored and are printed out for quality documentation. The forces in the springs of fuel spacers of LWR fuel elements are to be measured and compared with the standard requirements, deviations to be documented. For this task, another computer-controlled, automated device has been developed for measuring the spring forces at all required positions after positioning and fixation of the spacers. (orig./DG) [de

Science based integrated approach to advanced nuclear fuel development - vision, approach, and overview

Energy Technology Data Exchange (ETDEWEB)

Unal, Cetin [Los Alamos National Laboratory; Pasamehmetoglu, Kemal [IDAHO NATIONAL LAB; Carmack, Jon [IDAHO NATIONAL LAB

2010-01-01

Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Rcactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems is critical. In order to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. The purpose of this paper is to identify the modeling and simulation approach in order to deliver predictive tools for advanced fuels development. The coordination between experimental nuclear fuel design, development technical experts, and computational fuel modeling and simulation technical experts is a critical aspect of the approach and naturally leads to an integrated, goal-oriented science-based R & D approach and strengthens both the experimental and computational efforts. The Advanced Fuels Campaign (AFC) and Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Integrated Performance and Safety Code (IPSC) are working together to determine experimental data and modeling needs. The primary objective of the NEAMS fuels IPSC project is to deliver a coupled, three-dimensional, predictive computational platform for modeling the fabrication and both normal and abnormal operation of nuclear fuel pins and assemblies, applicable to both existing and future reactor fuel designs. The science based program is pursuing the development of an integrated multi-scale and multi-physics modeling and simulation platform for nuclear fuels. This overview paper discusses the vision, goals and approaches how to develop and implement the new approach.

Science based integrated approach to advanced nuclear fuel development - vision, approach, and overview

International Nuclear Information System (INIS)

Unal, Cetin; Pasamehmetoglu, Kemal; Carmack, Jon

2010-01-01

Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Rcactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems is critical. In order to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. The purpose of this paper is to identify the modeling and simulation approach in order to deliver predictive tools for advanced fuels development. The coordination between experimental nuclear fuel design, development technical experts, and computational fuel modeling and simulation technical experts is a critical aspect of the approach and naturally leads to an integrated, goal-oriented science-based R and D approach and strengthens both the experimental and computational efforts. The Advanced Fuels Campaign (AFC) and Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Integrated Performance and Safety Code (IPSC) are working together to determine experimental data and modeling needs. The primary objective of the NEAMS fuels IPSC project is to deliver a coupled, three-dimensional, predictive computational platform for modeling the fabrication and both normal and abnormal operation of nuclear fuel pins and assemblies, applicable to both existing and future reactor fuel designs. The science based program is pursuing the development of an integrated multi-scale and multi-physics modeling and simulation platform for nuclear fuels. This overview paper discusses the vision, goals and approaches how to develop and implement the new approach.

Analysis of uncertainty propagation in nuclear fuel cycle scenarios

International Nuclear Information System (INIS)

Krivtchik, Guillaume

2014-01-01

Nuclear scenario studies model nuclear fleet over a given period. They enable the comparison of different options for the reactor fleet evolution, and the management of the future fuel cycle materials, from mining to disposal, based on criteria such as installed capacity per reactor technology, mass inventories and flows, in the fuel cycle and in the waste. Uncertainties associated with nuclear data and scenario parameters (fuel, reactors and facilities characteristics) propagate along the isotopic chains in depletion calculations, and through out the scenario history, which reduces the precision of the results. The aim of this work is to develop, implement and use a stochastic uncertainty propagation methodology adapted to scenario studies. The method chosen is based on development of depletion computation surrogate models, which reduce the scenario studies computation time, and whose parameters include perturbations of the depletion model; and fabrication of equivalence model which take into account cross-sections perturbations for computation of fresh fuel enrichment. Then the uncertainty propagation methodology is applied to different scenarios of interest, considering different options of evolution for the French PWR fleet with SFR deployment. (author) [fr

EBR-II high-ramp transients under computer control

International Nuclear Information System (INIS)

Forrester, R.J.; Larson, H.A.; Christensen, L.J.; Booty, W.F.; Dean, E.M.

1983-01-01

During reactor run 122, EBR-II was subjected to 13 computer-controlled overpower transients at ramps of 4 MWt/s to qualify the facility and fuel for transient testing of LMFBR oxide fuels as part of the EBR-II operational-reliability-testing (ORT) program. A computer-controlled automatic control-rod drive system (ACRDS), designed by EBR-II personnel, permitted automatic control on demand power during the transients

SunFast: A sun workstation based, fuel analysis scoping tool for pressurized water reactors

International Nuclear Information System (INIS)

Bohnhoff, W.J.

1991-05-01

The objective of this research was to develop a fuel cycle scoping program for light water reactors and implement the program on a workstation class computer. Nuclear fuel management problems are quite formidable due to the many fuel arrangement options available. Therefore, an engineer must perform multigroup diffusion calculations for a variety of different strategies in order to determine an optimum core reload. Standard fine mesh finite difference codes result in a considerable computational cost. A better approach is to build upon the proven reliability of currently available mainframe computer programs, and improve the engineering efficiency by taking advantage of the most useful characteristic of workstations: enhanced man/machine interaction. This dissertation contains a description of the methods and a user's guide for the interactive fuel cycle scoping program, SunFast. SunFast provides computational speed and accuracy of solution along with a synergetic coupling between the user and the machine. It should prove to be a valuable tool when extensive sets of similar calculations must be done at a low cost as is the case for assessing fuel management strategies. 40 refs

Seismic analysis of freestanding fuel racks

International Nuclear Information System (INIS)

Gilmore, C.B.

1982-01-01

This paper presents a nonlinear transient dynamic time-history analysis of freestanding spent fuel storage racks subjected to seismic excitation. This type of storage rack is structurally unrestrained and submerged in water in the spent fuel pool of a nuclear power complex, holds (spent) fuel assemblies which have been removed from the reactor core. Nonlinearities in the fuel rack system include impact between the fuel assembly and surrounding cell due to clearances between them, friction due to sliding between the fuel rack support structure and spent fuel pool floor, and the lift-off of the fuel rack support structure from the spent fuel pool floor. The analysis of the fuel rack system includes impacting due to gap closures, energy losses due to impacting bodies, Coulomb damping between sliding surfaces, and hydrodynamic mass effects. Acceleration time history excitation development is discussed. Modeling considerations, such as the initial status of nonlinear elements, number of mode shapes to include in the analysis, modal damping, and integration time-step size are presented. The response of the fuel rack subjected to two-dimensional seismic excitation is analyzed by the modal superposition method, which has resulted in significant computer cost savings when compared to that of direct integration

Notion Of Artificial Labs Slow Global Warming And Advancing Engine Studies Perspectives On A Computational Experiment On Dual-Fuel Compression-Ignition Engine Research

Directory of Open Access Journals (Sweden)

Tonye K. Jack

2017-06-01

Full Text Available To appreciate clean energy applications of the dual-fuel internal combustion engine D-FICE with pilot Diesel fuel to aid public policy formulation in terms of present and future benefits to the modern transportation stationary power and promotion of oil and gas green- drilling the brief to an engine research team was to investigate the feasible advantages of dual-fuel compression-ignition engines guided by the following concerns i Sustainable fuel and engine power delivery ii The requirements for fuel flexibility iii Low exhausts emissions and environmental pollution iv Achieving low specific fuel consumption and economy for maximum power v The comparative advantages over the conventional Diesel engines vi Thermo-economic modeling and analysis for the optimal blend as basis for a benefitcost evaluation Planned in two stages for reduced cost and fast turnaround of results - initial preliminary stage with basic simple models and advanced stage with more detailed complex modeling. The paper describes a simplified MATLAB based computational experiment predictive model for the thermodynamic combustion and engine performance analysis of dual-fuel compression-ignition engine studies operating on the theoretical limited-pressure cycle with several alternative fuel-blends. Environmental implications for extreme temperature moderation are considered by finite-time thermodynamic modeling for maximum power with predictions for pollutants formation and control by reaction rates kinetics analysis of systematic reduced plausible coupled chemistry models through the NCN reaction pathway for the gas-phase reactions classes of interest. Controllable variables for engine-out pollutants emissions reduction and in particular NOx elimination are identified. Verifications and Validations VampV through Performance Comparisons were made using a clinical approach in selection of StrokeBore ratios greater-than and equal-to one amp88051 low-to-high engine speeds and medium

BWROPT: A multi-cycle BWR fuel cycle optimization code

Energy Technology Data Exchange (ETDEWEB)

Ottinger, Keith E.; Maldonado, G. Ivan, E-mail: Ivan.Maldonado@utk.edu

2015-09-15

Highlights: • A multi-cycle BWR fuel cycle optimization algorithm is presented. • New fuel inventory and core loading pattern determination. • The parallel simulated annealing algorithm was used for the optimization. • Variable sampling probabilities were compared to constant sampling probabilities. - Abstract: A new computer code for performing BWR in-core and out-of-core fuel cycle optimization for multiple cycles simultaneously has been developed. Parallel simulated annealing (PSA) is used to optimize the new fuel inventory and placement of new and reload fuel for each cycle considered. Several algorithm improvements were implemented and evaluated. The most significant of these are variable sampling probabilities and sampling new fuel types from an ordered array. A heuristic control rod pattern (CRP) search algorithm was also implemented, which is useful for single CRP determinations, however, this feature requires significant computational resources and is currently not practical for use in a full multi-cycle optimization. The PSA algorithm was demonstrated to be capable of significant objective function reduction and finding candidate loading patterns without constraint violations. The use of variable sampling probabilities was shown to reduce runtime while producing better results compared to using constant sampling probabilities. Sampling new fuel types from an ordered array was shown to have a mixed effect compared to random new fuel type sampling, whereby using both random and ordered sampling produced better results but required longer runtimes.

A computer program for calculation of the fuel cycle in pressurized water reactors

International Nuclear Information System (INIS)

Solanilla, R.

1976-01-01

The purpose of the FUCEFURE program is two-fold: first, it is designed to solve the problem of nuclear fuel cycle cost in one pressurized light water reactor calculation. The code was developed primarily for comparative and sensitivity studies. The program contains simple correlations between exposure and available depletion data used to predict the uranium and plutonium content of the fuel as a function of the fuel initial enrichment. Second, it has been devised to evaluate the nuclear fuel demand associated with an expanding nuclear power system. Evaluation can be carried out at any time and stage in the fuel cycle. The program can calculate the natural uranium and separate work requirements of any final and tails enrichment. It also can determine the nuclear power share of each reactor in the system when a decision has been made about the long-term nuclear power installations to be used and the types of PWR and fast breeder reactor characteristics to be involved in them. (author)

Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks

Science.gov (United States)

Poulson, D.; Durham, J. M.; Guardincerri, E.; Morris, C. L.; Bacon, J. D.; Plaud-Ramos, K.; Morley, D.; Hecht, A. A.

2017-01-01

Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This paper describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon scattering imaging. The specific application to monitoring spent nuclear fuel in dry storage casks is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, simulations indicate that missing fuel bundles can be detected with a statistical significance of ∼ 18 σ in less than two days exposure and a sensitivity at 1σ to a 5% missing portion of a fuel bundle. Potential detector technologies and geometries are discussed.

Hydrogen and fuel cell research: Institute for Integrated Energy Systems (IESVic)

International Nuclear Information System (INIS)

Pitt, L.

2006-01-01

Vision: IESVic's mission is to chart feasible paths to sustainable energy. Current research areas of investigation: 1. Energy system analysis 2. Computational fuel cell engineering; Fuel cell parameter measurement; Microscale fuel cells 3. Hydrogen dispersion studies for safety codes 4. Active magnetic refrigeration for hydrogen liquifaction and heat transfer in metal hydrides 5. Hydrogen and fuel cell system integration (author)

Simplified probabilistic risk assessment in fuel reprocessing

International Nuclear Information System (INIS)

Solbrig, C.W.

1993-01-01

An evaluation was made to determine if a backup mass tracking computer would significantly reduce the probability of criticality in the fuel reprocessing of the Integral Fast Reactor. Often tradeoff studies, such as this, must be made that would greatly benefit from a Probably Risk Assessment (PRA). The major benefits of a complete PRA can often be accrued with a Simplified Probabilistic Risk Assessment (SPRA). An SPRA was performed by selecting a representative fuel reprocessing operation (moving a piece of fuel) for analysis. It showed that the benefit of adding parallel computers was small compared to the benefit which could be obtained by adding parallelism to two computer input steps and two of the weighing operations. The probability of an incorrect material moves with the basic process is estimated to be 4 out of 100 moves. The actual values of the probability numbers are considered accurate to within an order of magnitude. The most useful result of developing the fault trees accrue from the ability to determine where significant improvements in the process can be made. By including the above mentioned parallelism, the error move rate can be reduced to 1 out of 1000

Development of nuclear fuel materials for research reactor

International Nuclear Information System (INIS)

Kim, Chang Kyu; Park, H. D.; Kim, K. H.; Lee, J. T.; Ryu, W. S.; Hwang, W.; Kim, H. N.; Kim, H. I.; Kwon, H. I.; Park, C.; Lee, B. C.; Park, J. M.; Lee, C. S.; Chae, H. T.; Im, N. J.; Cho, M. S.; Im, I. C.; Nam, C.; Lee, D. B.; Goh, Y. M.; Kim, J. D.; Ahn, H. S.; Woo, Y. M.; Chang, S. J.; Cho, H. D.

1997-09-01

This project has aimed at the development of U 3 Si dispersion fuel for the localization of HANARO fuel and the application of atomization process to advanced RERTR fuel development. The design criteria were established through the modified computer codes. Design documents were prepared and issued. The acceptable co-extrusion cladding was achieved. The electron beam welding technology has been developed and the sealing of the end plug and cladding was accomplished without defects. The atomization fuel meats have about 200% higher elongation and about 20% higher than comminution fuel meats. The thermal compatibility test showed that atomization fuel have about 30% higher stability that the comminution fuel. The pressure drops of 18 rods fuel assembly and 36 rods fuel assembly were measured to have 213 kPa and 205 kPa respectively. Apparent wear was not found in endurance test. The irradiation fuel was designed and fabricated by using low enriched uranium metal following the developed Q/A system. The safety analysis of irradiation fuel assembly was performed through linear power calculation by using MCNP4A code and centerline temperature calculation by using DIFAIR code. The quality assurance system has been established. The quality inspection technologies were developed. By acquiring the license, low enriched uranium of 100 kg as well as depleted uranium can be used. U 3 Si 2 -Al fuel swelled less than comminution fuel irrespective of temperature and fuel fraction in a compatibility test. The atomized U-10wt.%Mo powder were found to have gamma phase of isotropic structure. Gamma structure remained with a little swelling without any structure change at 400 deg C for 100 hours. Irradiation miniplate and test rig were designed preliminary manufactured. Thermal hydraulic and linear power calculations were performed by using PLTEMP and MCNP4A computer codes respectively. The hydraulic test showed that the pressure drop met the HANARO requirement. The vibration

The cost of fuel economy in the Indian passenger vehicle market

International Nuclear Information System (INIS)

Chugh, Randy; Cropper, Maureen; Narain, Urvashi

2011-01-01

To investigate how fuel economy is valued in the Indian car market, we compute the cost to Indian consumers of purchasing a more fuel-efficient vehicle and compare it to the benefit of lower fuel costs over the life of the vehicle. We estimate hedonic price functions for four market segments (petrol hatchbacks, diesel hatchbacks, petrol sedans, and diesel sedans) to compute 95% confidence intervals for the marginal cost to the consumer for an increase in fuel economy. We find that the associated present value of fuel savings falls within the 95% confidence interval for most specifications, in all market segments, for the years 2002 through 2006. Thus, we fail to consistently reject the hypothesis that consumers appropriately value fuel economy. - Highlights: → We examine the tradeoffs faced by new vehicle consumers in India. → We use hedonic price functions and instrumental variables. → We find no support for the hypothesis that consumers undervalue fuel economy. → Some consumers are willing to forgo substantial potential savings to own their preferred vehicle.

Chemistry and radiation in oxy-fuel combustion

DEFF Research Database (Denmark)

Yin, Chungen; Rosendahl, Lasse; Kær, Søren Knudsen

2011-01-01

In order to investigate the role of combustion chemistry and radiation heat transfer in oxy-fuel combustion modeling, a computational fluid dynamics (CFD) modeling study has been performed for two different oxy-fuel furnaces. One is a lab-scale 0.8MW oxy-natural gas flame furnace whose detailed in....... Among the key issues in combustion modeling, e.g., mixing, radiation and chemistry, this paper derives useful guidelines on radiation and chemistry implementation for reliable CFD analyses of oxy-fuel combustion, particularly for industrial applications....

Calculation device for fuel power history in BWR type reactors

International Nuclear Information System (INIS)

Sakagami, Masaharu.

1980-01-01

Purpose: To enable calculations for power history and various variants of power change in the power history of fuels in a BWR type reactor or the like. Constitution: The outputs of the process computation for the nuclear reactor by a process computer are stored and the reactor core power distribution is judged from the calculated values for the reactor core power distribution based on the stored data. Data such as for thermal power, core flow rate, control rod position and power distribution are recorded where the changes in the power distribution exceed a predetermined amount, and data such as for thermal power and core flow rate are recorded where the changes are within the level of the predetermined amount, as effective data excluding unnecessary data. Accordingly, the recorded data are taken out as required and the fuel power history and the various variants in the fuel power are calculated and determined in a calculation device for fuel power history and variants for fuel power fluctuation. (Furukawa, Y.)

How well does ORIGEN predict spent LWR fuel characteristics

International Nuclear Information System (INIS)

Mailen, J.C.; Roddy, J.W.

1987-01-01

The ORIGEN computer code is widely used to estimate the radionuclide content (actinides, activation and fission products) of irradiated reactor fuel and the resultant heat generation and radiation levels associated with such fuel. These estimates are used as source terms in safety evaluations of operating reactors, for evaluation of fuel behavior and regulation of the at-reactor storage, for transportation studies, and for evaluation of the ultimate geologic storage of the fuel. This survey summarizes the fuel data available in the open literature and, where given, the calculated values by ORIGEN. Plans for additional analyses of well-characterized reactor fuel samples to improve the validation of ORIGEN2 are discussed

Calculation of Heat-Bearing Agent’s Steady Flow in Fuel Bundle

Science.gov (United States)

Amosova, E. V.; Guba, G. G.

2017-11-01

This paper introduces the result of studying the heat exchange in the fuel bundle of the nuclear reactor’s fuel magazine. The article considers the fuel bundle of the infinite number of fuel elements, fuel elements are considered in the checkerboard fashion (at the tops of a regular triangle a fuel element is a plain round rod. The inhomogeneity of volume energy release in the rod forms the inhomogeneity of temperature and velocity fields, and pressure. Computational methods for studying hydrodynamics in magazines and cores with rod-shape fuel elements are based on a significant simplification of the problem: using basic (averaged) equations, isobaric section hypothesis, porous body model, etc. This could be explained by the complexity of math description of the three-dimensional fluid flow in the multi-connected area with the transfer coefficient anisotropy, curved boundaries and technical computation difficulties. Thus, calculative studying suggests itself as promising and important. There was developed a method for calculating the heat-mass exchange processes of inter-channel fuel element motions, which allows considering the contribution of natural convection to the heat-mass exchange based on the Navier-Stokes equations and Boussinesq approximation.

CASKETSS-DYNA2D: a nonlinear impact analysis computer program for nuclear fuel transport casks in two dimensional geometries

International Nuclear Information System (INIS)

Ikushima, Takeshi

1988-10-01

A nonlinear impact analysis computer program DYNA2D, which was developed by Hallquist, has been introduced from Lawrence Livermore National Laboratory for the purpose of using impact analysis of nuclear fuel transport casks. DYNA2D has been built in CASKETSS code system (CASKETSS means a modular code system for CASK Evaluation code system for Thermal and Structural Safety). Main features of DYNA2D are as follows; (1) This program has been programmed to provide near optimal speed on vector processing computers. (2) An explicit time integration method is used for fast calculation. (3) Many material models are available in the program. (4) A contact-impact algorithm permits gap and sliding along structural interfaces. (5) A rezoner has been embedded in the program. (6) The graphic program for representations of calculation is provided. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

Research works at the Physics Institute nuclear reactor for the nuclear power engineering

International Nuclear Information System (INIS)

Gavars, V.V.; Kalnin'sh, D.O.; Lapenas, A.A.; Tomsons, E.Ya.; Ulmanis, U.A.

1985-01-01

Methods for neutron spectra determination in the nuclear reactor core and vessel have been developed. On their base the neutron spectra at the Novo-Voronezh and kola NPPs have been measured. Such measurements are necessary for the determination of the nuclear fuel reprocessing coefficients, for the evaluation of the construction radiation-damage stability and the NPP economical efficiency on the whole. A new type of the reactor regulator - a liquid metal one - has been created. Such regulators are promising in respect of their use at the NPPs. The base for studying new radiation-damage-stable insulators has been founded. The materials obtained are now applied to designing the reactors of the second (fast) and the third (thermonuclear H) generations. There have developed and by a long-time exploitation checked a hot loop, used for materials irradiation. the nuclear reactor in Salaspils provides training of students being the new brain-power for the nuclear power engineering

WWER expert system for fuel failure analysis using the RTOP-CA code

International Nuclear Information System (INIS)

Likhanskii, V.; Evdokimov, I.; Sorokin, A.; Khromov, A.; Kanukova, V.; Apollonova, O.; Ugryumov, A.

2008-01-01

The computer expert system for fuel failure analysis of WWER during operation is presented. The diagnostics is based on the measurement of specific activity of reference nuclides in reactor primary coolant and application of a computer code for the data interpretation. The data analysis includes an evaluation of tramp uranium mass in reactor core, detection of failures by iodine and caesium spikes, evaluation of burnup of defective fuel. Evaluation of defective fuel burnup was carried out by applying the relation of caesium nuclides activity in spikes and relations of activities of gaseous fission products for steady state operational conditions. The method of burnup evaluation of defective fuel by use of fission gas activity is presented in details. The neural-network analysis is performed for determination of failed fuel rod number and defect size. Results of the expert system application are illustrated for several fuel campaigns on operating WWER NPPs. (authors)

Microprobe analysis of PuO2--UO2 nuclear fuel

International Nuclear Information System (INIS)

Clark, W.I.; Rasmussen, D.E.; Carlson, R.L.; Highley, D.M.

1977-01-01

For the preirradiation characterization of FFTF UO 2 --PuO 2 fuel, a program was developed to determine the preirradiation porosity, grain structure, and microcomposition of the fuel. Two computer programs, MITRAN and MERIT, were developed to evaluate the homogeneity of the fuel. These programs use elemental composition data generated by the electron microprobe. MITRAN determines information on the size and frequency of individual regions, whereas MERIT provides an index of the thermal performance of the fuel and calculated statistical data for comparison to other fuel batches

Pulse Detonation Assessment for Alternative Fuels

Directory of Open Access Journals (Sweden)

Muhammad Hanafi Azami

2017-03-01

Full Text Available The higher thermodynamic efficiency inherent in a detonation combustion based engine has already led to considerable interest in the development of wave rotor, pulse detonation, and rotating detonation engine configurations as alternative technologies offering improved performance for the next generation of aerospace propulsion systems, but it is now important to consider their emissions also. To assess both performance and emissions, this paper focuses on the feasibility of using alternative fuels in detonation combustion. Thus, the standard aviation fuels Jet-A, Acetylene, Jatropha Bio-synthetic Paraffinic Kerosene, Camelina Bio-synthetic Paraffinic Kerosene, Algal Biofuel, and Microalgae Biofuel are all asessed under detonation combustion conditions. An analytical model accounting for the Rankine-Hugoniot Equation, Rayleigh Line Equation, and Zel’dovich–von Neumann–Doering model, and taking into account single step chemistry and thermophysical properties for a stoichiometric mixture, is applied to a simple detonation tube test case configuration. The computed pressure rise and detonation velocity are shown to be in good agreement with published literature. Additional computations examine the effects of initial pressure, temperature, and mass flux on the physical properties of the flow. The results indicate that alternative fuels require higher initial mass flux and temperature to detonate. The benefits of alternative fuels appear significant.

Modular approach to LWR in-core fuel management

International Nuclear Information System (INIS)

Urli, N.; Pevec, D.; Coffou, E.; Petrovic, B.

1980-01-01

The most important methods in the LWR in-core fuel management are reviewed. A modular approach and optimization by use of infinite multiplication factor and power form-factor are favoured. A computer program for rotation of fuel assemblies at reloads has been developed which improves further fuel economy and reliability of nuclear power plants. The program has been tested on the PWR core and showed to decrease the power form-factors and flatten the radial power distribution. (author)

Computed isotopic inventory and dose assessment for SRS fuel and target assemblies

International Nuclear Information System (INIS)

Chandler, M.C.; Ketusky, E.T.; Thoman, D.C.

1995-01-01

Past studies have identified and evaluated important radionuclide contributors to dose from reprocessed spent fuel sent to waste for Mark 16B and 22 fuel assemblies and for Mark 31 A and 31B target assemblies. Fission-product distributions after a 5- and 15-year decay time were calculated for a ''representative'' set of irradiation conditions (i.e., reactor power, irradiation time, and exposure) for each type of assembly. The numerical calculations were performed using the SHIELD/GLASS system of codes. The sludge and supernate source terms for dose were studied separately with the significant radionuclide contributors for each identified and evaluated. Dose analysis considered both inhalation and ingestion pathways: The inhalation pathway was analyzed for both evaporative and volatile releases. Analysis of evaporative releases utilized release fractions for the individual radionuclides as defined in the ICRP-30 by DOE guidance. A release fraction of unity was assumed for each radionuclide under volatile-type releases, which would encompass internally initiated events (e.g., fires, explosions), process-initiated events, and externally initiated events. Radionuclides which contributed at least 1% to the overall dose were designated as significant contributors. The present analysis extends and complements the past analyses through considering a broader spectrum of fuel types and a wider range of irradiation conditions. The results provide for a more thorough understanding of the influences of fuel composition and irradiation parameters on fission product distributions (at 2 years or more). Additionally, the present work allows for a more comprehensive evaluation of radionuclide contributions to dose and an estimation of the variability in the radionuclide composition of the dose source term that results from the spent fuel sent to waste encompassing a broad spectrum of fuel compositions and irradiation conditions

Computation fluid dynamic modelling of natural convection heat flow in unpumped molten salt fuel tubes

International Nuclear Information System (INIS)

Leefe, S.; Jackson-Laver, P.; Scott, I.R.

2015-01-01

Use of static molten salt nuclear fuel in simple tubes was discarded in 1949 without considering how convection could affect its utility. This poster describes CFD studies showing that such tubes are practical as fuel elements in essentially conventional fuel assemblies. They can achieve power densities above 250kW per liter of fuel salt (higher than PWR's) and do so without causing the tube wall to heat to dangerous levels. This discovery enables the achievement of the many benefits of molten salt fuel while utilizing the highly developed technology, regulatory, non proliferation and safety benefits of current fuel assembly technology. (author)

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)

CFD prediction of flow and phase distribution in fuel assemblies with spacers

Energy Technology Data Exchange (ETDEWEB)

Anglart, H.; Nylund, O. [ABB Atom AB, Vasteras (Switzerland); Kurul, N. [Rensselaer Polytechnic Institute, Troy, NY (United States)] [and others

1995-09-01

This paper is concerned with the modeling and computation of multi-dimensional two-phase flows in BWR fuel assemblies. The modeling principles are presented based on using a two-fluid model in which lateral interfacial effects are accounted for. This model has been used to evaluate the velocity fields of both vapor and liquid phases, as well as phase distribution, between fuel elements in geometries similar to BWR fuel bundles. Furthermore, this model has been used to predict, in a detailed mechanistic manner, the effects of spacers on flow and phase distribution between, and pressure drop along, fuel elements. The related numerical simulations have been performed using a CFD computer code, CFDS-FLOW3D.

Structural analysis and modeling of water reactor fuel rod behavior

International Nuclear Information System (INIS)

Roshan Zamir, M.

2000-01-01

An important aspect of the design and analysis of nuclear reactor is the ability to predict the behavior of fuel elements in the adverse environment of a reactor system under normal and emergency operating conditions. To achieve these objectives and in order to provide a suitable computer code based on fundamental material properties for design and study of the thermal-mechanical behavior of water reactor fuel rods during their irradiation life and also to demonstrate the fuel rod design and modeling for students, The KIANA-1 computer program has been developed by the writer at Amir-Kabir university of technology with support of Atomic Energy Organization of Iran. KIANA-1 is an integral one-dimensional computer program for the thermal and mechanical analysis in order to predict fuel rods performance and also parameter study of Zircaloy-clad UO 2 fuel rod during steady state conditions. The code has been designed for the following main objectives: To give a solution for the steady state heat conduction equation for fuel as a heat source and clad by using finite difference, control volume and semi-analytical methods in order to predict the temperature profile in the fuel and cladding. To predict the inner gas pressures due to the filling gases and released gaseous fission products. To predict the fission gas production and release by using a simple diffusion model based on the Booth models and an empirical model. To calculate the fuel-clad gap conductance for cracked fuel with partial contact zones to a closed gap with strong contact. To predict the distribution of stress in three principal directions in the fuel and sheet by assuming one-dimensional plane strain and asymmetric idealization. To calculate the strain distribution in three principal directions and the corresponding deformation in the fuel and cladding. For this purpose the permanent strain such as creep or plasticity as well as the thermoelastic deformation and also the swelling, densification, cracking

Nuclear fuel burnup calculation in a Voronezh type reactor; Analiza izgaranja nuklearnog goriva u reaktoru tipa Voronjez

Energy Technology Data Exchange (ETDEWEB)

Matausek, M; Marinkovic, N; Kocic, A [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

1977-07-01

In order to summarize and present our abilities to perform a complex computation of the nuclear fuel burn-up, a systematic review of the available methods, algorithms and computer programmes is given in this paper. The computer programmes quoted have all been developed, modified and tested in our department, so that they can be successfully used in the analysis of nuclear power plants from both physics and economic points of view. For a commercially proven nuclear reactor - reactor of the Voronezh type - an illustrative computation of the fuel burn-up is performed. The typical results are presented and discussed. The conclusion concerns the completion of a modular scheme for the fuel burn-up calculation and the fuel cycle analysis (author)

The Preliminary Study for Numerical Computation of 37 Rod Bundle in CANDU Reactor

International Nuclear Information System (INIS)

Jeon, Yu Mi; Bae, Jun Ho; Park, Joo Hwan

2010-01-01

A typical CANDU 6 fuel bundle consists of 37 fuel rods supported by two endplates and separated by spacer pads at various locations. In addition, the bearing pads are brazed to each outer fuel rod with the aim of reducing the contact area between the fuel bundle and the pressure tube. Although the recent progress of CFD methods has provided opportunities for computing the thermal-hydraulic phenomena inside of a fuel channel, it is yet impossible to reflect the detailed shape of rod bundle on the numerical computation due to a lot of computing mesh and memory capacity. Hence, the previous studies conducted a numerical computation for smooth channels without considering spacers, bearing pads. But, it is well known that these components are an important factor to predict the pressure drop and heat transfer rate in a channel. In this study, the new computational method is proposed to solve the complex geometry such as a fuel rod bundle. In front of applying the method to the problem of 37 rod bundle, the validity and the accuracy of the method are tested by applying the method to the simple geometry. Based on the present result, the calculation for the fully shaped 37-rod bundle is scheduled for the future works

ELOCA: fuel element behaviour during high temperature transients

International Nuclear Information System (INIS)

Sills, H.E.

1979-03-01

The ELOCA computer code was developed to simulate the uniform thermal-mechanical behaviour of a fuel element during high-temperature transients such as a loss-of-coolant accident (LOCA). Primary emphasis is on the diametral expansion of the fuel sheath. The model assumed is a single UO2/zircaloy-clad element with axisymmetric properties. Physical effects considered by the code are fuel expansion, cracking and melting; variation, during the transient, of internal gas pressure; changing fuel/sheath heat transfer; thermal, elastic and plastic sheath deformation (anisotropic); Zr/H 2 O chemical reaction effects; and beryllium-assisted crack penetration of the sheath. (author)

AREVA modeling and predictive capacities to support PWR fuel assembly upgrading

International Nuclear Information System (INIS)

Canat, J. N.; Mollard, P.; Gentet, G.; Uyeda, G.

2008-01-01

The first goal of the fuel designer is to closely address the customers' expectations, with the aim of providing them in the shortest possible time a flawless product fully addressing their needs. However, the designer knows from experience that designing a new fuel assembly is a task which always lasts a long time. Depending on the extent and innovative dimension of the performed changes, development and qualification of new products have lasted from a few years to as much as roughly 15 years. Experience feedback proves that developing and qualifying a cladding material is the longest-term process, requiring the determination of its behavior laws under irradiation and also under accident conditions. Regarding fuel assembly structure, new development generally requires the irradiation of Lead Test Assemblies during a period of time representative of the fuel operating conditions. This explains the critical importance of high powered, top quality modeling to adequately support the fuel assembly design development and the behavior assessment. Advanced calculation codes and methods, improved modeling tools and test facilities, are key contributors to reinforced reliability, robustness, thermal hydraulic performance and maneuverability of nuclear fuel under ever more demanding operational conditions. Sophisticated, high powered modeling tools and representative test capacities are cutting the time necessary for AREVA to develop a new product, license it and load it in the core of a reactor. This trend towards greater modeling capability has been backed up by the upgrading of computing means over the last few years, allowing the consideration of a large number of factors and a higher accuracy in the representation of the modeled phenomena. This article details how predictive tools currently play a more and more important role in the design developments engaged by AREVA. They have led to a more physical approach to finding technical solutions and allowed their analytical

Resolving the three-dimensional microstructure of polymer electrolyte fuel cell electrodes using nanometer-scale X-ray computed tomography

Energy Technology Data Exchange (ETDEWEB)

Epting, William K.; Gelb, Jeff; Litster, Shawn

2012-02-08

The electrodes of a polymer electrolyte fuel cell (PEFC) are composite porous layers consisting of carbon and platinum nanoparticles and a polymer electrolyte binder. The proper composition and arrangement of these materials for fast reactant transport and high electrochemical activity is crucial to achieving high performance, long lifetimes, and low costs. Here, the microstructure of a PEFC electrode using nanometer-scale X-ray computed tomography (nano-CT) with a resolution of 50 nm is investigated. The nano-CT instrument obtains this resolution for the low-atomic-number catalyst support and binder using a combination of a Fresnel zone plate objective and Zernike phase contrast imaging. High-resolution, non-destructive imaging of the three-dimensional (3D) microstructures provides important new information on the size and form of the catalyst particle agglomerates and pore spaces. Transmission electron microscopy (TEM) and mercury intrusion porosimetry (MIP) is applied to evaluate the limits of the resolution and to verify the 3D reconstructions. The computational reconstructions and size distributions obtained with nano-CT can be used for evaluating electrode preparation, performing pore-scale simulations, and extracting effective morphological parameters for large-scale computational models. (Copyright copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

Feasibility studies of computed tomography in partial defect detection of spent BWR fuel

International Nuclear Information System (INIS)

Levai, F.; Tikkinen, J.; Tarvainen, M.; Arlt, R.

1990-10-01

Feasibility studies were made for tomographic reconstruction of a cross-sectional activity distribution of a spent nuclear fuel assembly. The purpose was to determine the number of fuel rods (pins) and localize the positisons where pins are missing. The activity distribution map showing the locations of fuel rods in the assembly was reconstructed. The theoretical part of this work consists of simulation of image reconstruction based on theoretically calculated data from a reference assembly model. Evaluation of different image reconstruction techniques was made. Measurements were made in real facility conditions. Gamma radiation from an irradiated 8 x 8 - 1 BWR fuel assembly was measured through a narrow custom made collimator from different angles and positions. The measured data set was used as projections for reconstructing the activity profile of the assembly in cross-sectional plane

Computational comparison of the effect of mixing grids of 'swirler' and 'run-through' types on flow parameters and the behavior of steam phase in WWER fuel assemblies

International Nuclear Information System (INIS)

Shcherbakov, S.; Sergeev, V.

2011-01-01

The results obtained using the TURBOFLOW computer code are presented for the numerical calculations of space distributions of coolant flow, heating and boiling characteristics in WWER fuel assemblies with regard to the effect of mixing grids of 'Swirler' and 'Run-through' types installed in FA on the above processes. The nature of the effect of these grids on coolant flow was demonstrated to be different. Thus, the relaxation length of cross flows after passing a 'Run-through' grid is five times as compared to a 'Swirler'-type grid, which correlates well with the experimental data. At the same time, accelerations occurring in the flow downstream of a 'Swirler'-type grid are by an order of magnitude greater than those after a 'Run-through' grid. As a result, the efficiency of one-phase coolant mixing is much higher for the grids of 'Run-through' type, while the efficiency of steam removal from fuel surface is much higher for 'Swirler'-type grids. To achieve optimal removal of steam from fuel surface it has been proposed to install into fuel assemblies two 'Swirler'-type grids in tandem at a distance of about 10 cm from each other with flow swirling in opposite directions. 'Run-through' grids would be appropriate for use for mixing in fuel assemblies with a high non-uniformity of fuel-by-fuel power generation. (authors)

Fuel analysis code FAIR and its high burnup modelling capabilities

International Nuclear Information System (INIS)

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

1995-01-01

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

Out-of-core nuclear fuel cycle optimization utilizing an engineering workstation

International Nuclear Information System (INIS)

Turinsky, P.J.; Comes, S.A.

1986-01-01

Within the past several years, rapid advances in computer technology have resulted in substantial increases in their performance. The net effect is that problems that could previously only be executed on mainframe computers can now be executed on micro- and minicomputers. The authors are interested in developing an engineering workstation for nuclear fuel management applications. An engineering workstation is defined as a microcomputer with enhanced graphics and communication capabilities. Current fuel management applications range from using workstations as front-end/back-end processors for mainframe computers to completing fuel management scoping calculations. More recently, interest in using workstations for final in-core design calculations has appeared. The authors have used the VAX 11/750 minicomputer, which is not truly an engineering workstation but has comparable performance, to complete both in-core and out-of-core fuel management scoping studies. In this paper, the authors concentrate on our out-of-core research. While much previous work in this area has dealt with decisions concerned with equilibrium cycles, the current project addresses the more realistic situation of nonequilibrium cycles

Optimal pin enrichment distributions in nuclear reactor fuel bundles

International Nuclear Information System (INIS)

Lim, E.Y.

1976-01-01

A methodology has been developed to determine the fuel pin enrichment distribution that yields the best approximation to a prescribed power distribution in nuclear reactor fuel bundles. The problem is formulated as an optimization problem in which the optimal pin enrichments minimize the sum of squared deviations between the actual and prescribed fuel pin powers. A constant average enrichment constraint is imposed to ensure that a suitable value of reactivity is present in the bundle. When constraints are added that limit the fuel pins to a few enrichment types, one must determine not only the optimal values of the enrichment types but also the optimal distribution of the enrichment types amongst the pins. A matrix of boolean variables is used to describe the assignment of enrichment types to the pins. This nonlinear mixed integer programming problem may be rigorously solved with either exhaustive enumeration or branch and bound methods using a modification of the algorithm from the continuous problem as a suboptimization. Unfortunately these methods are extremely cumbersome and computationally overwhelming. Solutions which require only a moderate computational effort are obtained by assuming that the fuel pin enrichments in this problem are ordered as in the solution to the continuous problem. Under this assumption search schemes using either exhaustive enumeration or branch and bound become computationally attractive. An adaptation of the Hooke--Jeeves pattern search technique is shown to be especially efficient

Modelling of fuel spray and combustion in diesel engines

Energy Technology Data Exchange (ETDEWEB)

Huttunen, M T; Kaario, O T [VTT Energy, Espoo (Finland)

1998-12-31

Fuel spray and air motion characteristics and combustion in direct injection (DI) diesel engines was studied using computational models of the commercial CFD-code FIRE. Physical subprocesses modelled included Lagrangian spray droplet movement and behaviour (atomisation, evaporation and interaction of spray droplets) and combustion of evaporated liquid spray in the gas phase. Fuel vapour combustion rate was described by the model of Magnussen and Hjertager. The standard k,{epsilon}-model was used for turbulence. In order to be able to predict combustion accurately, the fuel spray penetration should be predicted with reasonable accuracy. In this study, the standard drag coefficient had to be reduced in order to match the computed penetration to the measured one. In addition, the constants in the submodel describing droplet breakup also needed to be adjusted for closer agreement with the measurements. The characteristic time scale of fuel consumption rate k/C{sub R} {epsilon} strongly influenced the heat release and in-cylinder pressure. With a value around 2.0 to 5.0 for C{sub R}, the computed in-cylinder pressure during the compression stroke agreed quite well with the measurements. On the other hand, the in-cylinder pressure was underpredicted during the expansion stroke. This is partly due to the fact that hydrocarbon fuel combustion was modelled as a one-step reaction reading to CO{sub 2} and H{sub 2}O and inadequate description of the mixing of reactants and combustion products. (author) 16 refs.

Modelling of fuel spray and combustion in diesel engines

Energy Technology Data Exchange (ETDEWEB)

Huttunen, M.T.; Kaario, O.T. [VTT Energy, Espoo (Finland)

1997-12-31

Fuel spray and air motion characteristics and combustion in direct injection (DI) diesel engines was studied using computational models of the commercial CFD-code FIRE. Physical subprocesses modelled included Lagrangian spray droplet movement and behaviour (atomisation, evaporation and interaction of spray droplets) and combustion of evaporated liquid spray in the gas phase. Fuel vapour combustion rate was described by the model of Magnussen and Hjertager. The standard k,{epsilon}-model was used for turbulence. In order to be able to predict combustion accurately, the fuel spray penetration should be predicted with reasonable accuracy. In this study, the standard drag coefficient had to be reduced in order to match the computed penetration to the measured one. In addition, the constants in the submodel describing droplet breakup also needed to be adjusted for closer agreement with the measurements. The characteristic time scale of fuel consumption rate k/C{sub R} {epsilon} strongly influenced the heat release and in-cylinder pressure. With a value around 2.0 to 5.0 for C{sub R}, the computed in-cylinder pressure during the compression stroke agreed quite well with the measurements. On the other hand, the in-cylinder pressure was underpredicted during the expansion stroke. This is partly due to the fact that hydrocarbon fuel combustion was modelled as a one-step reaction reading to CO{sub 2} and H{sub 2}O and inadequate description of the mixing of reactants and combustion products. (author) 16 refs.

Feasibility study on AFR-100 fuel conversion from uranium-based fuel to thorium-based fuel

Energy Technology Data Exchange (ETDEWEB)

Heidet, F.; Kim, T.; Grandy, C. (Nuclear Engineering Division)

2012-07-30

best core performance characteristics for each of them. With the exception of the fuel type and enrichment, the reference AFR-100 core design characteristics were kept unchanged, including the general core layout and dimensions, assembly dimensions, materials and power rating. In addition, the mass of {sup 235}U required was kept within a reasonable range from that of the reference AFR-100 design. The core performance characteristics, kinetics parameters and reactivity feedback coefficients were calculated using the ANL suite of fast reactor analysis code systems. Orifice design calculations and the steady-state thermal-hydraulic analyses were performed using the SE2-ANL code. The thermal margins were evaluated by comparing the peak temperatures to the design limits for parameters such as the fuel melting temperature and the fuel-cladding eutectic temperature. The inherent safety features of AFR-100 cores proposed were assessed using the integral reactivity parameters of the quasi-static reactivity balance analysis. The design objectives and requirements, the computation methods used as well as a description of the core concept are provided in Section 2. The three major approaches considered are introduced in Section 3 and the neutronics performances of those approaches are discussed in the same section. The orifice zoning strategies used and the steady-state thermal-hydraulic performance are provided in Section 4. The kinetics and reactivity coefficients, including the inherent safety characteristics, are provided in Section 5, and the Conclusions in Section 6. Other scenarios studied and sensitivity studies are provided in the Appendix section.

Improvement of computer programs 'BAMBOO' and 'ASFRE-IV' for coupling analysis of deformation and thermal-hydraulics in a high burn-up fuel subassembly of fast reactor

International Nuclear Information System (INIS)

Uwaba, Tomoyuki; Ohshima, Hiroyuki; Imai, Yasutomo

2003-04-01

A simulation system of a deformed fuel subassembly is being developed for the structure integrity of high burn-up wire-spacer-type fuel subassemblies of sodium-cooled fast breeder reactors. This report describes a computer program improvement work for coupling analyses of deformation and thermal-hydraulics in a fuel subassembly as part of the simulation system development. In this work, a function of data conversion as an interface between a bundle deformation analysis program BAMBOO and a thermal hydraulic analysis program ASFRE-IV was incorporated to each program. BAMBOO was improved to accept the coolant temperature data from ASFRE-IV and to offer bundle deformation data to ASFRE-IV. ASFRE-IV was also improved to offer the coolant temperature data to BAMBOO and to obtain the bundle deformation data from BAMBOO. Improved BAMBOO and ASFRE-IV were applied to an analysis of 169-pin bundle for the program verification. It was confirmed that the coupling analysis gave the physically reasonable results on both deformation and thermal hydraulic behaviors in the fuel subassembly. (author)

Fuel fragmentation model advances using TEXAS-V

Energy Technology Data Exchange (ETDEWEB)

Corradini, M.L.; El-Beshbeeshy, M.; Nilsuwankowsit, S.; Tang, J. [Wisconsin Univ., Madison, WI (United States). Dept. of Nuclear Engineering and Engineering Physics

1998-01-01

Because an energetic fuel-coolant interaction may be a safety hazard, experiments are being conducted to investigate the fuel-coolant mixing/quenching process (FARO) as well as the energetics of vapor explosion propagation for high temperature fuel melt simulants (KROTOS, WFCI, ZrEX). In both types of experiments, the dynamic breakup of the fuel is one of the key aspects that must be fundamentally understood to better estimate the magnitude of the mixing/quenching process or the explosion energetics. To aid our understanding the TEXAS fuel-coolant interaction computer model has been developed and is being used to analyze these experiments. Recently, the models for dynamic fuel fragmentation during the mixing and explosion phases of the FCI have been improved by further insights into these processes. The purpose of this paper is to describe these enhancements and to demonstrate their improvements by analysis of particular JRC FCI data. (author)

Computer simulation of thermal-hydraulics of MNSR fuel-channel assembly using LabView

International Nuclear Information System (INIS)

Gadri, L. A.

2013-07-01

A LabView simulator of thermal hydraulics has been developed to demonstrate the temperature profile of coolant flow in the reactor core during normal operation. The simulator could equally be used for any transient behaviour of the reactor. Heat generation, transfer and the associated temperature profile in the fuel-channel elements viz: the coolant, cladding and fuel were studied and the corresponding analytical temperature equations in the axial and radial directions for the coolant, outer surface of the cladding, fuel surface and fuel center were obtained for the simulation using LabView. Tables of values for the equations were constructed by MATLAB and excel software programs. Plots of the equations with LabView were verified and validated with the graphs drawn by the MATLAB. In this thesis, an analysis of the effects of the coolant inlet temperature of 24.5°C and exit temperature of 70.0° on the temperature distribution in fuel-channel elements of the reactor core of cylindrical geometry was carried out. Other parameters, including the total fuel channel power, mass flow rate and convective heat transfer coefficient were varied to study the effects on the temperature profile. The analytical temperature equations in the fuel channel elements of the reactor core were obtained. MATLAB and Excel software were used to construct data for the equations. The plots by MATLAB were used to benchmark the LabVIEW simulation. Excellent agreement was obtained between the MATLAB plots and the LabView simulation results with an error margin of 0.001. The analysis of the results by comparing gradients of inlet temperature, total reactor channel power and mass flow indicated that inlet temperature gradient is one of the key parameters in determining the temperature profile in the MNSR core. (au)

Lumped-parameter fuel rod model for rapid thermal transients

International Nuclear Information System (INIS)

Perkins, K.R.; Ramshaw, J.D.

1975-07-01

The thermal behavior of fuel rods during simulated accident conditions is extremely sensitive to the heat transfer coefficient which is, in turn, very sensitive to the cladding surface temperature and the fluid conditions. The development of a semianalytical, lumped-parameter fuel rod model which is intended to provide accurate calculations, in a minimum amount of computer time, of the thermal response of fuel rods during a simulated loss-of-coolant accident is described. The results show good agreement with calculations from a comprehensive fuel-rod code (FRAP-T) currently in use at Aerojet Nuclear Company

On the problem of in-core fuel management in power reactors

International Nuclear Information System (INIS)

Marinkovic, N.; Matausek, M.V.

1985-01-01

Within the scope of in-core fuel management including refuelling schedule and reactivity control it is indispensable to define nuclear fuel worth, optimal depletion of the spent fuel assemblies as well as isotopic composition of the spent fuel. This paper shows the computed values of the mentioned parameters in case of different reactor types, PWR, WWER, HWR and BWR of 1000 MWe as well as the intensity of radiation of the spent fuel 3 and 1 years after fission.(author)

Reactor core T-H characteristics determination in case of parallel operation of different fuel assembly types

International Nuclear Information System (INIS)

Hermansky, J.; Petenyi, V.; Zavodsky, M.

2009-01-01

The WWER-440 nuclear fuel vendor permanently improve the assortment of produced nuclear fuel assemblies for achieving better fuel cycle economy and reactor operation safety. Therefore it is necessary to have the skilled methodology and computing code for analyzing factors which affecting the accuracy of flow redistributed determination through reactor on flows through separate parts of reactor core in case of parallel operation different assembly types. Whereas the geometric parameters of new manufactured assemblies were changed recently, the calculated flows through the fuel parts of different type of assemblies are depended also on their real position in reactor core. Therefore the computing code CORFLO was developed in VUJE Trnava for carrying out stationary analyses of T-H characteristics of reactor core within 60 deg symmetry. The CORFLO code deals the area of the active core which consists of 312 fuel assemblies and 37 control assemblies. Regarding the rotational 60 deg symmetry of reactor core only 1/6 of reactor core with 59 fuel assemblies is calculated. Computing code is verified and validated at this time. Paper presents the short description of computing code CORFLO with some calculated results. (Authors)

Design and fuel management of PWR cores to optimize the once-through fuel cycle

International Nuclear Information System (INIS)

Fujita, E.K.; Driscoll, M.J.; Lanning, D.D.

1978-08-01

The once-through fuel cycle has been analyzed to see if there are substantial prospects for improved uranium ore utilization in current light water reactors, with a specific focus on pressurized water reactors. The types of changes which have been examined are: (1) re-optimization of fuel pin diameter and lattice pitch, (2) axial power shaping by enrichment gradation in fresh fuel, (3) use of 6-batch cores with semi-annual refueling, (4) use of 6-batch cores with annual refueling, hence greater extended (approximately doubled) burnup, (5) use of radial reflector assemblies, (6) use of internally heterogeneous cores (simple seed/blanket configurations), (7) use of power/temperature coastdown at the end of life to extend burnup, (8) use of metal or diluted oxide fuel, (9) use of thorium, and (10) use of isotopically separated low sigma/sub a/ cladding material. State-of-the-art LWR computational methods, LEOPARD/PDQ-7/FLARE-G, were used to investigate these modifications

Automated uncertainty analysis methods in the FRAP computer codes

International Nuclear Information System (INIS)

Peck, S.O.

1980-01-01

A user oriented, automated uncertainty analysis capability has been incorporated in the Fuel Rod Analysis Program (FRAP) computer codes. The FRAP codes have been developed for the analysis of Light Water Reactor fuel rod behavior during steady state (FRAPCON) and transient (FRAP-T) conditions as part of the United States Nuclear Regulatory Commission's Water Reactor Safety Research Program. The objective of uncertainty analysis of these codes is to obtain estimates of the uncertainty in computed outputs of the codes is to obtain estimates of the uncertainty in computed outputs of the codes as a function of known uncertainties in input variables. This paper presents the methods used to generate an uncertainty analysis of a large computer code, discusses the assumptions that are made, and shows techniques for testing them. An uncertainty analysis of FRAP-T calculated fuel rod behavior during a hypothetical loss-of-coolant transient is presented as an example and carried through the discussion to illustrate the various concepts

Automated breeder fuel fabrication

International Nuclear Information System (INIS)

Goldmann, L.H.; Frederickson, J.R.

1983-01-01

The objective of the Secure Automated Fabrication (SAF) Project is to develop remotely operated equipment for the processing and manufacturing of breeder reactor fuel pins. The SAF line will be installed in the Fuels and Materials Examination Facility (FMEF). The FMEF is presently under construction at the Department of Energy's (DOE) Hanford site near Richland, Washington, and is operated by the Westinghouse Hanford Company (WHC). The fabrication and support systems of the SAF line are designed for computer-controlled operation from a centralized control room. Remote and automated fuel fabriction operations will result in: reduced radiation exposure to workers; enhanced safeguards; improved product quality; near real-time accountability, and increased productivity. The present schedule calls for installation of SAF line equipment in the FMEF beginning in 1984, with qualifying runs starting in 1986 and production commencing in 1987. 5 figures

Development of methodology for the analysis of fuel behavior in light water reactor in design basis accidents

International Nuclear Information System (INIS)

Salatov, A. A.; Goncharov, A. A.; Eremenko, A. S.; Kuznetsov, V. I.; Bolnov, V. A.; Gusev, A. S.; Dolgov, A. B.; Ugryumov, A. V.

2013-01-01

The report attempts to analyze the current experience of the safety fuel for light-water reactors (LWRs) under design-basis accident conditions in terms of its compliance with international requirements for licensing nuclear power plants. The components of fuel behavior analysis methodology in design basis accidents in LWRs were considered, such as classification of design basis accidents, phenomenology of fuel behavior in design basis accidents, system of fuel safety criteria and their experimental support, applicability of used computer codes and input data for computational analysis of the fuel behavior in accidents, way of accounting for the uncertainty of calculation models and the input data. A brief history of the development of probabilistic safety analysis methodology for nuclear power plants abroad is considered. The examples of a conservative approach to safety analysis of VVER fuel and probabilistic approach to safety analysis of fuel TVS-K are performed. Actual problems in development of the methodology of analyzing the behavior of VVER fuel at the design basis accident conditions consist, according to the authors opinion, in following: 1) Development of a common methodology for analyzing the behavior of VVER fuel in the design basis accidents, implementing a realistic approach to the analysis of uncertainty - in the future it is necessary for the licensing of operating VVER fuel abroad; 2) Experimental and analytical support to the methodology: experimental studies to identify and study the characteristics of the key uncertainties of computational models of fuel and the cladding, development of computational models of key events in codes, validation code on the basis of integral experiments

Preparation for commissioning of nuclear plant with reference to British Nuclear Fuels Plc fuel handling plant project

International Nuclear Information System (INIS)

Bamber, D.R.

1987-01-01

The new Fuel Handling Plant at British Nuclear Fuels Sellafield is part of a Pound 550M complex which provides facilities for the receipt, storage and mechanical preparation of both Magnox and A.G.R. fuel. The plant is very large and complex with considerable use of computer based process control systems, providing for physical and nuclear safety. The preparation of such plant for active commissioning necessitates a great many physical checks and technical evaluations in support of its safety case. This paper describes arrangements for plant commissioning checks, against the regulatory framework and explains the physical preparations necessary for their timely accomplishment. (author)

Preparation for commissioning of nuclear plant with reference to British Nuclear Fuels Plc fuel handling plant project

International Nuclear Information System (INIS)

Bamber, D.

1987-01-01

The new Fuel Handing Plant at British Nuclear Fuels Sellafield is part of a Pound 550M complex which provides facilities for the receipt, storage and mechanical preparation of both magnox and A.G.R. fuel. The plant is very large and complex with considerable use of computer based process control systems, providing for physical and nuclear safety. The preparation of such plant for ''active'' commissioning necessitates a great many physical checks and technical evaluations in support of its safety case. This paper describes arrangements for plant commissioning checks, against the regulatory framework and explains the physical preparations necessary for their timely accomplishment. (author)

DUPIC fuel irradiation test and performance evaluation; the performance analysis of pellet-cladding contact fuel

Energy Technology Data Exchange (ETDEWEB)

Ho, K. I.; Kim, H. M.; Yang, K. B.; Choi, S. J. [Suwon University, Whasung (Korea)

2002-04-01

Thermal and mechanical models were reviewed, and selected for the analysis of nuclear fuel performance in reactor. 2 dimensional FEM software was developed. Thermal models-gap conductances, thermal conductivity of pellets, fission gas release, temperature distribution-were set and packaged into a software. Both thermal and mechanical models were interrelated to each other, and the final results, fuel performance during irradiation is obtained by iteration calculation. Also, the contact phenomena between pellet and cladding was analysed by mechanical computer software which was developed during this work. dimensional FEM program was developed which estimate the mechanical behavior and the thermal behaviors of nuclear fuel during irradiation. Since there is a importance during the mechanical deformation analysis in describing pellet-cladding contact phenomena, simplified 2 dimensional calculation method is used after the contact. The estimation of thermal fuel behavior during irradiation was compared with the results of other. 8 refs., 17 figs. (Author)

Ciclon: A neutronic fuel management program for PWR's consecutive cycles

International Nuclear Information System (INIS)

Aragones, J.M.

1977-01-01

The program description and user's manual of a new computer code is given. Ciclon performs the neutronic calculation of consecutive reload cycles for PWR's fuel management optimization. Fuel characteristics and burnup data, region or batch sizes, loading schemes and state of previously irradiated fuel are input to the code. Cycle lengths or feed enrichments and burnup sharing for each region or batch are calculate using different core neutronic models and printed or punched in standard fuel management format. (author) [es

Fuel Management of WWER-1000 Reactors of Kudankulam Nuclear Power Plant, India

International Nuclear Information System (INIS)

Pandey, Y.; Chauhan, A.

2008-01-01

Two units of WWER-1000 reactors of Russian design are under construction at Kudankulam site in India. These reactors are expected to be commissioned in 2008. The fuel management services for these reactors shall be carried out using Russian Computer codes. This paper includes a brief description of the core, fuel assembly lattice and physics modeling of the lattice and core for these reactors. Presented in this paper are the salient features of the core load pattern designs and fuel performance for 8 operating cycles of these reactors. The paper describes key improvements in the core load pattern designs to enhance the fuel utilization and its thermal behaviour. Presented in the paper are also the on site fuel management strategies with regard to fuel inventory and nuclear material accounting. A computer code for Fuel Inventory and Nuclear Material Accounting (FINMAC) has been developed for this purpose. The code FINMAC takes care of receipt of fresh fuel, flow between various accounting sub areas (ASAs), burnup or production of nuclear isotopes in the reactor cores and discharge from the reactor core. The code generates Material Balance Reports (MBRs) and Composition of Ending Inventory Reports (COEIs) as per the IAEA standards. (authors)

Fission product release by fuel oxidation after water ingress

International Nuclear Information System (INIS)

Schreiber.

1990-01-01

On the basis of data obtained by a literature search, a computer code has been established for the calculation of the degree of oxidation of the fuel in the damaged fuel particles, and hence of the fission product release as a function of the time period of steam ingress. (orig.) [de

Stand for visual ultrasonic testing of spent fuel

International Nuclear Information System (INIS)

Czajkowski, W.; Borek-Kruszewska, E.

2001-01-01

A stand for visual and ultrasonic testing of spent fuel, constructed under Strategic Governmental Programme for management of spent fuel and radioactive waste, is presented in the paper. The stand, named 'STEND-1', built up at the Institute of Atomic Energy in Swjerk, is appointed for underwater visual testing of spent fuel elements type MR6 and WWR by means of TV-CCD camera and image processing system and for ultrasonic scanning of external surface of these elements by means of video scan immersion transducer and straight UHT connector. 'STEND-1' is built using flexible in use, high-tensile, anodized aluminum profiles. All the profiles feature longitudinal grooves to accommodate connecting elements and for the attachment of accessories at any position. They are also characterised by straight-through core bores for use with standard fastening elements and to accommodate accessory components. Stand, equipped with automatic control and processing system based on personal computer, may be manually or automatically controlled. Control system of movements of the camera in the vertical axis and rotational movement of spent fuel element permits to fix chosen location of fuel element with accuracy better than 0.1 mm. High resolution of ultrasonic method allows to record damages of outer surface of order 0.1 mm. The results of visual testing of spent fuel are recorded on video tape and then may be stored on the hard disc of the personal computer and presented in shape of photo or picture. Only selected damage surfaces of spent fuel elements are tested by means of ultrasonic scanning. All possibilities of the stand and results of visual testing of spent fuel type WWR are presented in the paper. (author)

MTR radiological database for SRS spent nuclear fuel facilities

International Nuclear Information System (INIS)

Blanchard, A.

2000-01-01

A database for radiological characterization of incoming Material Test Reactor (MTR) fuel has been developed for application to the Receiving Basin for Offsite Fuels (RBOF) and L-Basin spent fuel storage facilities at the Savannah River Site (SRS). This database provides a quick quantitative check to determine if SRS bound spent fuel is radiologically bounded by the Reference Fuel Assembly used in the L-Basin and RBOF authorization bases. The developed database considers pertinent characteristics of domestic and foreign research reactor fuel including exposure, fuel enrichment, irradiation time, cooling time, and fuel-to-moderator ratio. The supplied tables replace the time-consuming studies associated with authorization of SRS bound spent fuel with simple hand calculations. Additionally, the comprehensive database provides the means to overcome resource limitations, since a series of simple, yet conservative, hand calculations can now be performed in a timely manner and replace computational and technical staff requirements

Burn up Theoretical Analysis of A Thorium Fuel Rod in Light Water Reactor

International Nuclear Information System (INIS)

Gaber, F.A.; Aziz, M.; Elsheikh, B.

2008-01-01

A computer model was designed to analyze the burn up and irradiation of both Th-Pu and Th-U fuel rod in a typical light water reactors conditions. MCNP computer model was designed to simulate the fuel rod burnup and evaluate neutron flux and group constants . A system of ordinary differential equations were solved numerically to evaluate the isotopic concentrations for both the two types of fuel using the previous calculated data from MCNP model. The results are analyzed and compared with published data where satisfactory agreement was found

Current developments of fuel fabrication technologies at the plutonium fuel production facility, PFPF

International Nuclear Information System (INIS)

Asakura, K.; Aono, S.; Yamaguchi, T.; Deguchi, M.

2000-01-01

The Japan Nuclear Cycle Development Institute, JNC, designed, constructed and has operated the Plutonium Fuel Production Facility, PFPF, at the JNC Tokai Works to supply MOX fuels to the proto-type Fast Breeder Reactor, FBR, 'MONJU' and the experimental FBR 'JOYO' with 5 tonMOX/year of fabrication capability. Reduction of personal radiation exposure to a large amount of plutonium is one of the most important subjects in the development of MOX fabrication facility on a large scale. As the solution of this issue, the PFPF has introduced automated and/or remote controlled equipment in conjunction with computer controlled operation scheme. The PFPF started its operation in 1988 with JOYO reload fuel fabrication and has demonstrated MOX fuel fabrication on a large scale through JOYO and MONJU fuel fabrication for this decade. Through these operations, it has become obvious that several numbers of equipment initially installed in the PFPF need improvements in their performance and maintenance for commercial utilization of plutonium in the future. Furthermore, fuel fabrication of low density MOX pellets adopted in the MONJU fuel required a complete inspection because of difficulties in pellet fabrication compared with high density pellet for JOYO. This paper describes new pressing equipment with a powder recovery system, and pellet finishing and inspection equipment which has multiple functions, such as grinding measurements of outer diameter and density, and inspection of appearance to improve efficiency in the pellet finishing and inspection steps. Another development of technology concerning an annular pellet and an innovative process for MOX fuel fabrication are also described in this paper. (author)

The Preliminary Study for Numerical Computation of 37 Rod Bundle in CANDU Reactor

International Nuclear Information System (INIS)

Jeon, Yu Mi; Park, Joo Hwan

2010-09-01

A typical CANDU 6 fuel bundle consists of 37 fuel rods supported by two endplates and separated by spacer pads at various locations. In addition, the bearing pads are brazed to each outer fuel rod with the aim of reducing the contact area between the fuel bundle and the pressure tube. Although the recent progress of CFD methods has provided opportunities for computing the thermal-hydraulic phenomena inside of a fuel channel, it is yet impossible to reflect numerical computations on the detailed shape of rod bundle due to challenges with computing mesh and memory capacity. Hence, the previous studies conducted a numerical computation for smooth channels without considering spacers and bearing pads. But, it is well known that these components are an important factor to predict the pressure drop and heat transfer rate in a channel. In this study, the new computational method is proposed to solve complex geometry such as a fuel rod bundle. Before applying a solution to the problem of the 37 rod bundle, the validity and the accuracy of the method are tested by applying the method to simple geometry. The split channel method has been proposed with the aim of computing the fully shaped CANDU fuel channel with detailed components. The validity was tested by applying the method to the single channel problem. The average temperature have similar values for the considered two methods, while the local temperature shows a slight difference by the effect of conduction heat transfer in the solid region of a rod. Based on the present result, the calculation for the fully shaped 37-rod bundle is scheduled for future work

Analysis of dismantling possibility and unloading efforts of fuel assemblies from core of WWER

International Nuclear Information System (INIS)

Danilov, V.; Dobrov, V.; Semishkin, V.; Vasilchenko, I.

2006-01-01

The computation methods of optimal dismantling sequence of fuel assemblies (FA) from core of WWER after different operating periods and accident conditions are considered. The algorithms of fuel dismantling sequence are constructed both on the basis of analysis of mutual spacer grid overlaps of adjacent fuel assemblies and numerical structure analysis of efforts required for FA removal as FA heaving from the core. Computation results for core dismantling sequence after 3-year operating period and LB LOCA are presented in the paper

Evaluation of the SCANAIR Computer Code

International Nuclear Information System (INIS)

Jernkvist, Lars Olof; Massih, Ali

2001-11-01

The SCANAIR computer code, version 3.2, has been evaluated from the standpoint of its capability to analyze, simulate and predict nuclear fuel behavior during severe power transients. SCANAIR calculates the thermal and mechanical behavior of a pressurized water reactor (PWR) fuel rod during a postulated reactivity initiated accident (RIA), and our evaluation indicates that SCANAIR is a state of the art computational tool for this purpose. Our evaluation starts by reviewing the basic theoretical models in SCANAIR, namely the governing equations for heat transfer, the mechanical response of fuel and clad, and the fission gas release behavior. The numerical methods used to solve the governing equations are briefly reviewed, and the range of applicability of the models and their limitations are discussed and illustrated with examples. Next, the main features of the SCANAIR user interface are delineated. The code requires an extensive amount of input data, in order to define burnup-dependent initial conditions to the simulated RIA. These data must be provided in a special format by a thermal-mechanical fuel rod analysis code. The user also has to supply the transient power history under RIA as input, which requires a code for neutronics calculation. The programming structure and documentation of the code are also addressed in our evaluation. SCANAIR is programmed in Fortran-77, and makes use of several general Fortran-77 libraries for handling input/output, data storage and graphical presentation of computed results. The documentation of SCANAIR and its helping libraries is generally of good quality. A drawback with SCANAIR in its present form, is that the code and its pre- and post-processors are tied to computers running the Unix or Linux operating systems. As part of our evaluation, we have performed a large number of computations with SCANAIR, some of which are documented in this report. The computations presented here include a hypothetical RIA in a high

Innovative nuclear fuels: results and strategy

International Nuclear Information System (INIS)

Stan, Marius

2009-01-01

To facilitate the discovery and design of innovative nuclear fuels, multi-scale models and simulations are used to predict irradiation effects on the thermal conductivity, oxygen diffusivity, and thermal expansion of oxide fuels. The multi-scale approach is illustrated using results on ceramic fuels with a focus on predictions of point defect concentrations, stoichiometry, and phase stability. The high performance computer simulations include coupled heat transport, diffusion, and thermal expansion, gas bubble formation and temperature evolution in a fuel element consisting of UO2 fuel and metallic cladding. The second part of the talk is dedicated to a discussion of an international strategy for developing advanced, innovative nuclear fuels. Four initiative are proposed to accelerate the discovery and design of new materials: (a) Develop an international pool of experts, (b) Create Institutes for Materials Discovery and Design, (c) Create an International Knowledge base for experimental data, models (mathematical expressions), and simulations (codes) and (d) Organize international workshops and conference sessions. The paper ends with a discussion of existing and emerging international collaborations.

X-ray vision of fuel sprays

International Nuclear Information System (INIS)

Wang, J.

2005-01-01

With brilliant synchrotron X-ray sources, microsecond time-resolved synchrotron X-ray radiography and tomography have been used to elucidate the detailed three-dimensional structure and dynamics of high-pressure high-speed fuel sprays in the near-nozzle region. The measurement allows quantitative determination of the fuel distribution in the optically impenetrable region owing to the multiple scattering of visible light by small atomized fuel droplets surrounding the jet. X-radiographs of the jet-induced shock waves prove that the fuel jets become supersonic under appropriate injection conditions and that the quantitative analysis of the thermodynamic properties of the shock waves can also be derived from the most direct measurement. In other situations where extremely axial-asymmetric sprays are encountered, mass deconvolution and cross-sectional fuel distribution models can be computed based on the monochromatic and time-resolved X-radiographic images collected from various rotational orientations of the sprays. Such quantitative analysis reveals the never-before-reported characteristics and most detailed near-nozzle mass distribution of highly transient fuel sprays

CANDU type fuel behavior evaluation - a probabilistic approach

International Nuclear Information System (INIS)

Moscalu, D.R.; Horhoianu, G.; Popescu, I.A.; Olteanu, G.

1995-01-01

In order to realistically assess the behavior of the fuel elements during in-reactor operation, probabilistic methods have recently been introduced in the analysis of fuel performance. The present paper summarizes the achievements in this field at the Institute for Nuclear Research (INR), pointing out some advantages of the utilized method in the evaluation of CANDU type fuel behavior in steady state conditions. The Response Surface Method (RSM) has been selected for the investigation of the effects of the variability in fuel element computer code inputs on the code outputs (fuel element performance parameters). A new developed version of the probabilistic code APMESRA based on RSM is briefly presented. The examples of application include the analysis of the results of an in-reactor fuel element experiment and the investigation of the calculated performance parameter distribution for a new CANDU type extended burnup fuel element design. (author)

A user input manual for single fuel rod behaviour analysis code FEMAXI-III

International Nuclear Information System (INIS)

Saito, Hiroaki; Yanagisawa, Kazuaki; Fujita, Misao.

1983-03-01

Principal objectives of Safety related research in connection with lighr water reactor fuel rods under normal operating condition are mainly addressed 1) to assess fuel integrity under steady state condition and 2) to generate initial condition under hypothetical accident. These assessments have to be relied principally upon steady state fuel behaviour computing code that is able to calculate fuel conditions to tbe occurred in a various manner. To achieve these objectives, efforts have been made to develope analytical computer code that calculates in-reactor fuel rod behaviour in best estimate manner. The computer code developed for the prediction of the long-term burnup response of single fuel rod under light water reactor condition is the third in a series of code versions:FEMAMI-III. The code calculates temperature, rod internal gas pressure, fission gas release and pellet-cladding interaction related rod deformation as a function of time-dependent fuel rod power and coolant boundary conditions. This document serves as a user input manual for the code FEMAMI-III which has opened to the public in year of 1982. A general description of the code input and output are included together with typical examples of input data. A detailed description of structures, analytical submodels and solution schemes in the code shall be given in the separate document to be published. (author)

Thermohydraulic analysis of assemblies containing up to 2/7 fuel rods

International Nuclear Information System (INIS)

Ferreira, W.J.; Luz, M.

1985-01-01

The COBRA IV-I computer code was tested using data from the Fast Flux Test Facility. Then this code was applied to the analysis of fuel assemblies from the Binary Breeder Reactor. Previously this analysis was carried out using the COBRA III-C code which allows only for the calculations of fuel assemblies having seven fuel pins. The COBRA IV-I permits the calculation of fuel assemblies containing up to 217 fuel pins and the inclusion of blanket and shielding effects. (F.E.) [pt

FREC-4A: a computer program to predict fuel rod performance under normal reactor operation

International Nuclear Information System (INIS)

Harayama, Yasuo; Izumi, Fumio

1981-10-01

The program FREC-4A (Fuel Reliability Evaluation Code-version 4A) is used for predicting fuel rod performance in normal reactor operation. The performance is calculated in accordance with the irradiation history of fuel rods. Emphasis is placed on the prediction of the axial elongation of claddings induced by pellet-cladding mechanical interaction, including the influence of initially preloaded springs inserted in fuel rod lower plenums. In the FREC-4A, an fuel rod is divided into axial segments. In each segment, it is assumed that the temperature, stress and strain are axi-symmetrical, and the axial strain in constant in fuel pellets and in a cladding, though the values in the pellets and in the cladding are different. The calculation of the contact load and the clearance along the length of a fuel rod and the stress and strain in each segment is explained. The method adopted in the FREC-4A is simple, and suitable to predict the deformation of fuel rods over their full length. This report is described on the outline of the program, the method of solving the stiffness equations, the calculation models, the input data such as irradiation history, output distribution, material properties and pores, the printing-out of input data and calculated results. (Kako, I.)

Mathematical modeling of solid oxide fuel cells

Science.gov (United States)

Lu, Cheng-Yi; Maloney, Thomas M.

1988-01-01

Development of predictive techniques, with regard to cell behavior, under various operating conditions is needed to improve cell performance, increase energy density, reduce manufacturing cost, and to broaden utilization of various fuels. Such technology would be especially beneficial for the solid oxide fuel cells (SOFC) at it early demonstration stage. The development of computer models to calculate the temperature, CD, reactant distributions in the tubular and monolithic SOFCs. Results indicate that problems of nonuniform heat generation and fuel gas depletion in the tubular cell module, and of size limitions in the monolithic (MOD 0) design may be encountered during FC operation.

Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes

International Nuclear Information System (INIS)

Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

2016-01-01

To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

FISS: a computer program for reactor systems studies

International Nuclear Information System (INIS)

Tamm, H.; Sherman, G.R.; Wright, J.H.; Nieman, R.E.

1979-08-01

ΣFISSΣ is a computer code for use in investigating alternative fuel cycle strategies for Canadian and world nuclear programs. The code performs a system simulation accounting for dynamic effects of growing nuclear systems. Facilities in the model include storage for irradiated fuel, mines, plants for enrichment, fuel fabrication, fuel reprocessing and heavy water, and reactors. FISS is particularly useful for comparing various reactor strategies and studying sensitivities of resource consumption, capital investment and energy costs with changes in fuel cycle parameters, reactor parameters and financial variables. (author)

Annual report of the working group 'fuel pin and fuel element mechanics' of the Institut fuer Reaktortechnik (IRT) of the Technische Hochschule Darmstadt for the Fast Breeder Project

International Nuclear Information System (INIS)

Fabian, H.; Humbach, W.; Lassmann, K.; Mueller, J.J.; Preusser, T.; Schmelz, K.

1978-09-01

This report comprises six single lectures given at an information meeting organized by the Institut fuer Reaktortechnik der Technischen Hochschule Darmstadt (IRT) in Darmstadt on April 24, 1978. The lectures are an account of work performed at IRT on the mechanics of fuel pins and fuel elements and supported by the Fast Breeder Project (PSB) of KfK. These activities can be broken down into studies of the integral fuel pin (URANUS computer code) and into multidimensional studies of the fuel pin using the finite-element method (FINEL and ZIDRIG computer codes). Moreover, a report is presented of the status of the test facility for simulation of out-of-pile cladding tube loads and of the IRT project on the simulation and analysis of radiation damage. (orig./GL) [de

A scheme of better utilization of PWR spent fuels

International Nuclear Information System (INIS)

Chung, Bum Jin; Kang, Chang Soon

1991-01-01

The recycle of PWR spent fuels in a CANDU reactor, so called the tandem fuel cycle is investigated in this study. This scheme of utilizing PWR spent fuels will ease the shortage of spent fuel storage capacity as well as will improve the use of uranium resources. The minimum modification the design of present CANDU reactor is seeked in the recycle. Nine different fuel types are considered in this work and are classified into two categories: refabrication and reconfiguration. For refabrication, PWR spent fuels are processed and refabricated into the present 37 rod lattice structure of fuel bundle, and for reconfiguration, meanwhile, spent fuels are simply disassembled and rods are cut to fit into the present grid configuration of fuel bundle without refabrication. For each fuel option, the neutronics calculation of lattice was conducted to evaluate the allowable burn up and distribution. The fuel cycle cost of each option was also computed to assess the economic justification. The results show that most tandem fuel cycle option considered in this study are technically feasible as well as economically viable. (Author)

User's manual for the CC3 computer models of the concept for disposal of Canada's nuclear fuel waste

International Nuclear Information System (INIS)

Dougan, K.D.; Wojciechowski, L.C.

1995-06-01

Atomic Energy of Canada Limited (AECL) is assessing a concept for disposing of CANDU reactor fuel waste in a vault deep in plutonic rock of the Canadian Shield. A computer program called the Systems Variability Analysis Code (SYVAC) has been developed as an analytical tool for the postclosure (long-term) assessment of the concept, and for environmental assessments of other systems. SYVAC3, the third generation of the code, is an executive program that directs repeated simulation of the disposal system, which is represented by the CC3 (Canadian Concept, generation 3) models comprising a design-specific vault, a site-specific geosphere and a biosphere typical of the Canadian Shield. (author). 23 refs., 7 tabs., 21 figs

Review of the IAEA Nuclear Fuel Cycle Materials Section activities related to WWER fuel

International Nuclear Information System (INIS)

Killeen, J.

2003-01-01

The IAEA Nuclear Fuel Cycle Programme, designated as Programme B, has the main objective of supporting Member States in policy making, strategic planning, developing technology and addressing issues with respect to safe, reliable, economically efficient, proliferation resistant and environmentally sound nuclear fuel cycle. This paper is concentrated on describing the work within Sub-programme B.2 'Fuel Performance and Technology'. Two Technical Working Groups assist in the preparation of the IAEA programme in the nuclear fuel cycle area - Technical Working Group on Water Reactor Fuel Performance and Technology and Technical Working Group on Nuclear Fuel Cycle Options. The activities of the Unit within the Nuclear Fuel Cycle and Materials Section working on Fuel Performance and Technology are given, based on the sub-programme structure of the Agency programme and budget for 2002-2003. Within the framework of Co-ordinated Research Projects a study of the delayed hydride cracking (DHC) of the zirconium alloys used in pressurised heavy water reactors (PHWR) involving 10 countries has been completed. It achieved very effective transfer of know-how at the laboratory level in three technologically important areas: 1) Controlled hydriding of samples to predetermined levels; 2) Accurate measurement of hydrogen concentrations at the relatively low levels found in pressure tubes and RBMK channel tubes; and 3) In the determination of DHC rates under various conditions of temperature and stress. A new project has been started on the 'Improvement of Models used for Fuel Behaviour Simulation' (FUMEX II) to assist Member States in improving the predictive capabilities of computer codes used in modelling fuel behaviour for extended burnup. The IAEA also collaborates with organisations in the Member States to support activities and meetings on nuclear fuel cycle related topics

Nuclear fuel waste disposal

International Nuclear Information System (INIS)

1982-01-01

This film for a general audience deals with nuclear fuel waste management in Canada, where research is concentrating on land based geologic disposal of wastes rather than on reprocessing of fuel. The waste management programme is based on cooperation of the AECL, various universities and Ontario Hydro. Findings of research institutes in other countries are taken into account as well. The long-term effects of buried radioactive wastes on humans (ground water, food chain etc.) are carefully studied with the help of computer models. Animated sequences illustrate the behaviour of radionuclides and explain the idea of a multiple barrier system to minimize the danger of radiation hazards

Computer generated movies at Hanford

International Nuclear Information System (INIS)

Lewis, C.H.; Fox, G.L.

1979-10-01

The message contained in the results of a large computer program is often difficult to present to large groups of people. This difficulty may be overcome by using 16mm color movie techniques. This presentation shows the results of directly using computer output to explain a story about fuel behavior during a power transient

Burn-up credit criticality safety benchmark phase VII - UO2 fuel: study of spent fuel compositions for long-term disposal

International Nuclear Information System (INIS)

2012-01-01

After spent nuclear fuel (SNF) is discharged from a nuclear reactor, fuel composition and reactivity continue to vary as a function of time due to the decay of unstable nuclides. Accurate predictions of the concentrations of long-lived radionuclides in SNF, which represent a significant potential hazard to human beings and to the environment over a very long period, are particularly necessary for radiological dose assessments. This report assesses the ability of existing computer codes and associated nuclear data to predict isotopic compositions and their corresponding neutron multiplication factor (k eff ) values for pressurised-water-reactor (PWR) UO 2 fuel at 50 GWd/MTU burn-up in a generic spent fuel cask configuration. Fuel decay compositions and k eff values have been calculated for 30 post-irradiation time steps out to one million years

Thermochemistry of nuclear fuels in advanced reactors

International Nuclear Information System (INIS)

Agarwal, Renu

2015-01-01

The presence of a large number of elements, accompanied with steep temperature gradient results in dynamic chemistry during nuclear fuel burn-up. Understanding this chemistry is very important for efficient and safe usage of nuclear fuels. The radioactive nature of these fuels puts lot of constraint on regulatory bodies to ensure their accident free operation in the reactors. One of the common aims of advanced fuels is to achieve high burn-up. As burn-up of the fuel increases, chemistry of fission-products becomes increasingly more important. To understand different phenomenon taking place in-pile, many out of-pile experiments are carried out. Extensive studies of thermodynamic properties, phase analysis, thermophysical property evaluation, fuel-fission product clad compatibility are carried out with relevant compounds and simulated fuels (SIMFUEL). All these data are compiled and jointly evaluated using different computational methods to predict fuel behaviour during burn-up. Only when this combined experimental and theoretical information confirms safe operation of the pin, a test pin is prepared and burnt in a test reactor. Every fuel has a different chemistry and different constraints associated with it. In this talk, various thermo-chemical aspects of some of the advanced fuels, mixed carbide, mixed nitride, 'Pu' rich MOX, 'Th' based AHWR fuels and metallic fuels will be discussed. (author)

Fission gas behaviour in water reactor fuels

International Nuclear Information System (INIS)

2002-01-01

During irradiation, nuclear fuel changes volume, primarily through swelling. This swelling is caused by the fission products and in particular by the volatile ones such as krypton and xenon, called fission gas. Fission gas behaviour needs to be reliably predicted in order to make better use of nuclear fuel, a factor which can help to achieve the economic competitiveness required by today's markets. These proceedings communicate the results of an international seminar which reviewed recent progress in the field of fission gas behaviour in light water reactor fuel and sought to improve the models used in computer codes predicting fission gas release. State-of-the-art knowledge is presented for both uranium-oxide and mixed-oxide fuels loaded in water reactors. (author)

Software to improve spent fuel measurements using the FDET

International Nuclear Information System (INIS)

Staples, P.; Beddingfield, D.; Lestone, J.; Pelowitz, D.; Sprinkle, J.; Bytchkov, V.; Starovich, Z.; Harizanov, I.; Vellejo-Luna, J.; Lavender, C.

2001-01-01

Full text: Vast quantities of spent fuel are available for safeguard measurements, primarily in Commonwealth of Independent States (CIS) of the former Soviet Union. This spent fuel, much of which consists of long cooling time material, is going to become less unique in the world safeguards arena as reprocessing projects or permanent repositories continue to be delayed or postponed. The long cooling time of many of the spent fuel assemblies in the CIS countries being prepared for intermediate term storage promotes the possibility of increased accuracy in spent fuel assays. An important point to consider for the future that could advance safeguards measurements for re-verification and inspection measurements would be to determine what safeguards requirements should be imposed upon this 'new' class of spent fuel. Improvements in measurement capability will obviously affect the safeguards requirements. What most significantly enables this progress in spent fuel measurements is the improvement in computer processing power and software enhancements leading to user-friendly Graphical User Interfaces (GUI's). The software used for these projects significantly reduces the IAEA inspector's time both learning and operating computer and data acquisition systems. While at the same time by standardizing the spent fuel measurements it is possible to increase reproducibility and reliability of the measurement data. The inclusion of various analysis algorithms into the operating environment, which can be performed in real time upon the measurement data, can also lead to increases in safeguard reliability and improvements in efficiency to plant operations. (author)

An interactive computer code for calculation of gas-phase chemical equilibrium (EQLBRM)

Science.gov (United States)

Pratt, B. S.; Pratt, D. T.

1984-01-01

A user friendly, menu driven, interactive computer program known as EQLBRM which calculates the adiabatic equilibrium temperature and product composition resulting from the combustion of hydrocarbon fuels with air, at specified constant pressure and enthalpy is discussed. The program is developed primarily as an instructional tool to be run on small computers to allow the user to economically and efficiency explore the effects of varying fuel type, air/fuel ratio, inlet air and/or fuel temperature, and operating pressure on the performance of continuous combustion devices such as gas turbine combustors, Stirling engine burners, and power generation furnaces.

State of art in FE-based fuel performance codes

International Nuclear Information System (INIS)

Kim, Hyo Chan; Yang, Yong Sik; Kim, Dae Ho; Bang, Je Geon; Kim, Sun Ki; Koo, Yang Hyun

2013-01-01

Fuel performance codes approximate this complex behavior using an axisymmetric, axially-stacked, one-dimensional radial representation to save computation cost. However, the need for improved modeling of PCMI and, particularly, the importance of multidimensional capability for accurate fuel performance simulation has been identified as safety margin decreases. Finite element (FE) method that is reliable and proven solution in mechanical field has been introduced into fuel performance codes for multidimensional analysis. The present state of the art in numerical simulation of FE-based fuel performance predominantly involves 2-D axisymmetric model and 3-D volumetric model. The FRAPCON and FRAPTRAN own 1.5-D and 2-D FE model to simulate PCMI and cladding ballooning. In 2-D simulation, the FALCON code, developed by EPRI, is a 2-D (R-Z and R-θ) fully thermal-mechanically coupled steady-state and transient FE-based fuel behavior code. The French codes TOUTATIS and ALCYONE which are 3-D, and typically used to investigate localized behavior. In 2008, the Idaho National Laboratory (INL) has been developing multidimensional (2-D and 3-D) nuclear fuel performance code called BISON. In this paper, the current state of FE-based fuel performance code and their models are presented. Based on investigation into the codes, requirements and direction of development for new FE-based fuel performance code can be discussed. Based on comparison of models in FE-based fuel performance code, status of art in the codes can be discussed. A new FE-based fuel performance code should include typical pellet and cladding models which all codes own. In particular, specified pellet and cladding model such as gaseous swelling and high burnup structure (HBS) model should be developed to improve accuracy of code as well as consider AC condition. To reduce computation cost, the approximated gap and the optimized contact model should be also developed

Economic analysis of thorium-uranium fuel cycle introduced into PWRs

International Nuclear Information System (INIS)

Fan Li; Sun Qian

2014-01-01

Using PWR of Daya Bay Unit l as the reference reactor, a validated computer code was used to calculate the fuel cycle costs for uranium fuel cycle and thorium-uranium fuel cycle over the following 20 0perational years respectively. The calculation results show that the thorium-uranium fuel cycle is economically competitive with the uranium fuel cycle when reprocessing mode is adopted. For thorium-uranium fuel cycle, if the price of natural uranium is higher than 120 $ /pound U_3O_8, the fuel cycle cost of the direct disposal mode is greater than that of the reprocessing mode. Therefore, when the uranium price may maintain a high level long-termly, adopting reprocessing mode will benefit the economic advantage for the thorium-uranium fuel cycle introduced into PWRs. (authors)

Qualification of FEAST 3.0 and FEAT 4.0 computer codes

International Nuclear Information System (INIS)

Xu, Z.; Lai, L.; Sim, K.-S.; Huang, F.; Wong, B.

2005-01-01

FEAST (Finite Element Analysis for Stresses) is an AECL computer code used to assess the structural integrity of the CANDU fuel element. FEAST models the thermo-elastic, thermo-elasto-plastic and creep deformations in CANDU fuel. FEAT (Finite Element Analysis for Temperature) is another AECL computer code and is used to assess the thermal integrity of fuel elements. FEAT models the steady-state and transient heat flows in CANDU fuel, under conditions such as flux depression, end flux peaking, temperature-dependent thermal conductivity, and non-uniform time-dependent boundary conditions. Both computer programs are used in design and qualification analyses of CANDU fuel. Formal qualifications (including coding verification and validation) of both codes were performed, in accordance with AECL software quality assurance (SQA) manual and procedures that are consistent with CSA N286.7-99. Validation of FEAST 3.0 shows very good agreement with independent analytical solutions or measurements. Validation of FEAT 4.0 also shows very good agreement with independent WIMS-AECL calculations, analytical solutions, ANSYS calculations and measurement. (author)

Qualification of FEAST 3.0 and FEAT 4.0 computer codes

Energy Technology Data Exchange (ETDEWEB)

Xu, Z.; Lai, L.; Sim, K.-S.; Huang, F.; Wong, B. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

2005-07-01

FEAST (Finite Element Analysis for Stresses) is an AECL computer code used to assess the structural integrity of the CANDU fuel element. FEAST models the thermo-elastic, thermo-elasto-plastic and creep deformations in CANDU fuel. FEAT (Finite Element Analysis for Temperature) is another AECL computer code and is used to assess the thermal integrity of fuel elements. FEAT models the steady-state and transient heat flows in CANDU fuel, under conditions such as flux depression, end flux peaking, temperature-dependent thermal conductivity, and non-uniform time-dependent boundary conditions. Both computer programs are used in design and qualification analyses of CANDU fuel. Formal qualifications (including coding verification and validation) of both codes were performed, in accordance with AECL software quality assurance (SQA) manual and procedures that are consistent with CSA N286.7-99. Validation of FEAST 3.0 shows very good agreement with independent analytical solutions or measurements. Validation of FEAT 4.0 also shows very good agreement with independent WIMS-AECL calculations, analytical solutions, ANSYS calculations and measurement. (author)

Modelling of the spent fuel heat-up in the spent fuel pools using one-dimensional system codes and CFD codes

Energy Technology Data Exchange (ETDEWEB)

Grazevicius, Audrius; Kaliatka, Algirdas [Lithuanian Energy Institute, Kaunas (Lithuania). Lab. of Nuclear Installation Safety

2017-07-15

The main functions of spent fuel pools are to remove the residual heat from spent fuel assemblies and to perform the function of biological shielding. In the case of loss of heat removal from spent fuel pool, the fuel rods and pool water temperatures would increase continuously. After the saturated temperature is reached, due to evaporation of water the pool water level would drop, eventually causing the uncover of spent fuel assemblies, fuel overheating and fuel rods failure. This paper presents an analysis of loss of heat removal accident in spent fuel pool of BWR 4 and a comparison of two different modelling approaches. The one-dimensional system thermal-hydraulic computer code RELAP5 and CFD tool ANSYS Fluent were used for the analysis. The results are similar, but the local effects cannot be simulated using a one-dimensional code. The ANSYS Fluent calculation demonstrated that this three-dimensional treatment allows to avoid the need for many one-dimensional modelling assumptions in the pool modelling and enables to reduce the uncertainties associated with natural circulation flow calculation.

COBRA - 3C/KFKI: a digital computer program for steady and transient thermal-hydraulic analysis of rod bundle nuclear fuel elements

International Nuclear Information System (INIS)

Vigassy, J.; Kovacs, L.M.

1977-11-01

COBRA-3C/KFKI is a digital computer program for the CDC-3300 computer in FORTRAN language. The program is a revised version of the original COBRA-3C code. The code calculates steady-state and transient flow and enthalpy transport in rod-bundle nuclear fuel elements in both boiling and nonboiling conditions. The mathematical model is formulated by dividing the bundle flow area into flow subchannels that are assumed to contain one-dimensional flow and are coupled to each other by turbulent and diversion crossflow mixing. The program neglects sonic velocity propagation but allows for a temporal and spatial acceleration of the diversion crossflow in the transverse momentum equation. A semiexplicit finite-difference scheme is used to perform a boundary-value solution where the boundary conditions are the inlet enthalpy, inlet flow rate and exit pressure. (D.P.)

Modelling of pellet-cladding interaction for PWRs reactors fuel rods

International Nuclear Information System (INIS)

Esteves, A.M.

1991-01-01

The pellet-cladding interaction that can occur in a PWR fuel rod design is modelled with the computer codes FRAPCON-1 and ANSYS. The fuel performance code FRAPCON-1 analyzes the fuel rod irradiation behavior and generates the initial conditions for the localized fuel rod thermal and mechanical modelling in two and three-dimensional finite elements with ANSYS. In the mechanical modelling, a pellet fragment is placed in the fuel rod gap. Two types of fuel rod cladding materials are considered: Zircaloy and austenitic stainless steel. Linear and non-linear material behaviors are allowed. Elastic, plastic and creep behaviors are considered for the cladding materials. The modelling is applied to Angra-II fuel rod design. The results are analyzed and compared. (author)

A computer aided solvent extraction process design in nuclear fuel reprocessing

International Nuclear Information System (INIS)

Britto, S.E.; Purandare, H.D.; Lawande, S.V.

1977-01-01

A rigorous and conceptual design is attempted of the first step of flowsheet formulation for solvent extraction process for fuel reprocessing plant. The design incorporates three cycles of extraction contractors; the first optimised to maximise Pu recovery while the second and third cycles to maximise fission product decontaminations. There are three basic types of extraction steps in these different cycles requiring painstaking design, namely, extraction-scrub, Pu strip-scrub and simple strip. The extraction system to start with is: U nitrate - Pu nitrate - fission product nitrates - nitric acid - tri-butyl phosphate/diluent. With suitable simplifying assumptions and adopting the concept of discrete equilibrium stagewise operation, simple X-Y operating diagrams could be used. The calculations could therefore be done using McCabe Thiele graphical method. The procedure adopted was to consider the macro-component of U to obtain initial optimum flow sheet details and the number of theoretical stages for each contactor and later to incorporate the behaviour of Pu and fission products. A computer program was written to calculate, for different combinations of nitric acid salting strengths, (1) the U concentration profiles along the contractors and (2) the number of stages needed for various different solvent and aqueous phase flow ratios, using experimentally obtained equilibrium data. The method used is indicated and some samples of results obtained for three types of extraction-scrub operation studied are given. These simplified calculations provided the necessary insight into these difficult operations. (auth.)

Hastelloy X fuel element creep relaxation and residual effects

International Nuclear Information System (INIS)

Castle, R.A.

1971-01-01

A worst case, seven element, asymmetric fuel, thermal environment was assumed and a creep relaxation analysis generated. The fuel element clad is .020 inch Hastelloy X. The contact load decreased from 11.6 pounds to 5.87 pounds in 100,000 hours. The residual stresses were then computed for various shutdown times. (U.S.)

Flow analysis of tubular fuel assembly using CFD code

International Nuclear Information System (INIS)

Park, J. H.; Park, C.; Chae, H. T.

2004-01-01

Based on the experiences of HANARO, a new research reactor is under conceptual design preparing for future needs of research reactor. Considering various aspects such as nuclear physics, thermal-hydraulics, mechanical structure and the applicability of HANARO technology, a tubular type fuel has been considered as that of a new research reactor. Tubular type fuel has several circular fuel layers, and each layer consists of 3 curved fuel plates arranged with constant small gap to build up cooling channels. In the thermal-hydraulic point, it is very important to maintain each channel flow velocity be equal as much as possible, because the small gaps between curved thin fuel plates independently forms separate coolant channels, which may cause a thermal-hydraulic problem in certain conditions. In this study, commercial CFD(Computational Fluid Dynamics) code, Fluent, has been used to investigate flow characteristics of tubular type fuel assembly. According to the computation results for the preliminary conceptual design, there is a serious lack of uniformity of average velocity on the each coolant channel. Some changes for initial conceptual design were done to improve the balance of velocity distribution, and analysis was done again, too. The results for the revised design showed that the uniformity of each channel velocity was improved significantly. The influence of outermost channel gap width on the velocity distribution was also examined

Small Volume Fuel Testers Report

Energy Technology Data Exchange (ETDEWEB)

Schoegl, I. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); McNenly, M. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Killingsworth, N. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2016-10-31

Micro-liter fuel ignition testing (μ-FIT) is based on the premise that characteristics FREI (Flames with Repetitive Extinction and Ignition, i.e. cyclically occurring combustion events within heated capillaries), are linked to fuel properties. In early FY16, proof-of-concept measurements with primary reference fuel (PRF) mixtures, i.e. blends of n-heptane and iso-octane, yielded clear evidence for the feasibility of the approach. Our experiments showed that it is critical to accurately link observed flame positions to local temperatures, which provides information on ignition, extinction and flame propagation, all of which are known to be impacted by fuel properties. In FY16, one major hurdle was uncertainty of temperature calibration, which required significant efforts for corrective action that were not included in the original scope of work. Temperature calibrations are obtained by translating a thermocouple within the capillary in absence of a flame. While measurements have good repeatability when accounting for transient and insertion effects, results from nominally identical thermocouples reveal unacceptable uncertainty (up to ±50K), which is attributed to variations in thermocouple placement and manufacturing tolerances. This issue is currently being resolved by switching to non-intrusive optical temperature measurements. Updates are expected to yield uncertainties of less than ±10K, while also eliminating transient and insertion effects. The experimental work was complemented by computational efforts where it was shown that a simplified Lagrangian zero-D model with detailed kinetics yields fuelspecific differentiation of ignition temperatures for simple fuels that are consistent with experiments. Further, a 2D transient model was implemented in OpenFOAM to investigate combustion behavior of simple fuels at elevated pressure. In an upcoming visit to LLNL, more advanced simulations using LLNL’s computational tools (e.g. zero-RK) are planned, which will

FRESCO-II: A computer program for analysis of fission product release from spherical HTGR-fuel elements in irradiation and annealing experiments

International Nuclear Information System (INIS)

Krohn, H.; Finken, R.

1983-06-01

The modular computer code FRESCO has been developed to describe the mechanism of fission product release from a HTGR-Core under accident conditions. By changing some program modules it has been extended to take into account the transport phenomena (i.e. recoil) too, which only occur under reactor operating conditions and during the irradiation experiments. For this report, the release of cesium and strontium from three HTGR-fuel elements has been evaluated and compared with the experimental data. The results show that the measured release can be described by the considered models. (orig.) [de

Development of a computational model applied to a unitary 144 CM{sup 2} proton exchange membrane fuel cell; Desenvolvimento de um modelo numerico computacional aplicado a uma celula a combustivel unitaria de 144 CM{sup 2} tipo PEM

Energy Technology Data Exchange (ETDEWEB)

Robalinho, Eric

2009-07-01

This work presents the development of a numerical computer model and methodology to study and design polymeric exchange membrane fuel cell - PEM. For the validation of experimental results, a sequence of routines, appropriate to fit the data obtained in the laboratory, was described. At the computational implementation it was created a new strategy of coupling two 3-dimensional models to satisfy the requirements of the comprehensive model of the fuel cell, including its various geometries and materials, as well as the various physical and chemical processes simulated. To effective assessment of the real cell analogy with numerical model, numerical studies were carried out. Comparisons with values obtained in the literature, characterization of variables through laboratory experiments and estimates from models already tested in the literature were also performed. Regarding the experimental part, a prototype of a fuel cell unit of 144 cm{sup 2} of geometric area was designed, produced and operated at laboratory with the purpose of validating the numerical computer model proposed, with positive results. The results of simulations for the 2D and 3D geometries proposed are presented in the form of polarization curves, highlighting the catalytic layer model based on the geometry of agglomerates. Parametric and sensitivity studies are presented to illustrate the change in performance of the fuel cell studied. The final model is robust and useful as a tool for design and optimization of PEM type fuel cells in a wide range of operating conditions. (author)