CATHENA 4. A thermalhydraulics network analysis code

International Nuclear Information System (INIS)

Aydemir, N.U.; Hanna, B.N.

2009-01-01

Canadian Algorithm for THErmalhydraulic Network Analysis (CATHENA) is a one-dimensional, non-equilibrium, two-phase, two fluid network analysis code that has been in use for over two decades by various groups in Canada and around the world. The objective of the present paper is to describe the design, application and future development plans for the CATHENA 4 thermalhydraulics network analysis code, which is a modernized version of the present frozen CATHENA 3 code. The new code is designed in modular form, using the Fortran 95 (F95) programming language. The semi-implicit numerical integration scheme of CATHENA 3 is re-written to implement a fully-implicit methodology using Newton's iterative solution scheme suitable for nonlinear equations. The closure relations, as a first step, have been converted from the existing CATHENA 3 implementation to F95 but modularized to achieve ease of maintenance. The paper presents the field equations, followed by a description of the Newton's scheme used. The finite-difference form of the field equations is given, followed by a discussion of convergence criteria. Two applications of CATHENA 4 are presented to demonstrate the temporal and spatial convergence of the new code for problems with known solutions or available experimental data. (author)

CFD Analysis for the Steady State Test of CS28-1 Simulating High Temperature Chemical Reactions in CANDU Fuel Channel

International Nuclear Information System (INIS)

Park, Ju Hwan; Kang, Hyung Seok; Rhee, Bo Wook

2006-05-01

The establishment of safety analysis system and technology for CANDU reactors has been performed at KAERI. As for one of these researches, single CANDU fuel bundle has been simulated by CATHENA for the post-blowdown event to consider the complicated geometry and heat transfer in the fuel channel. In the previous LBLOCA analysis methodology adopted for Wolsong 2, 3, 4 licensing, the fuel channel blowdown phase was analyzed by a CANDU system analysis code CATHENA and the post-blowdown phase of fuel channel was analyzed by CHAN-IIA code. The use of one computer code in consecutive analyses appeared to be desirable for consistency and simplicity in the safety analysis process. However, validation of the high temperature post-blowdown fuel channel model in the CATHENA before being used in the accident analysis is necessary. Experimental data for the 37-element fuel bundle that fueled CANDU-6 has not been performed. The benchmark problems for the 37-element fuel bundle using CFD code will be compared with the test results of the 28-element fuel bundle in the CS28-1 experiment. A full grid model of FES to the calandria tube simulating the test section was generated. The number of the generated mesh in the grid model was 4,324,340 cells. The boundary and heat source conditions, and properties data in the CFD analysis were given according to the test results and reference data. Thermal hydraulic phenomena in the fuel channel were simulated by a compressible flow, a highly turbulent flow, and a convection/conduction/radiation heat transfer. The natural convection flow of CO 2 due to a large temperature difference in the gap between the pressure and the calandria tubes was treated by Boussinesq's buoyancy model. The CFD results showed good agreement with the test results as a whole. The inner/middle/outer FES temperature distributions of the CFD results showed a small overestimated value of about 30 .deg. C at the entrance region, but good agreement at the outlet region. The

Validation of CATHENA MOD-3.5/Rev0 for single-phase water hammer

International Nuclear Information System (INIS)

Beuthe, T.G.

2000-01-01

This paper describes work performed to validate the system thermalhydraulics code CATHENA MOD-3.5c/Rev0 for single-phase water hammer. Simulations were performed and are compared quantitatively against numerical tests and experimental results from the Seven Sisters Water Hammer Facility to demonstrate CATHENA can predict the creation and propagation of pressure waves when valves are opened and closed. Simulations were also performed to show CATHENA can model the behaviour of reflected and transmitted pressure waves at area changes, dead ends, tanks, boundary conditions, and orifices in simple and more complex piping systems. The CATHENA results are shown to calculate pressure and wave propagation speeds to within 0.2% and 0.5% respectively for numerical tests and within 3.3% and 5% for experimental results respectively. These results are used to help validate CATHENA for use in single-phase water hammer analysis. They also provide assurance that the fundamental parameters needed to successfully model more complex forms of water hammer are accounted for in the MOD-3.5c/Rev0 version of CATHENA, and represent the first step in the process to validate the code for use in modelling two-phase water hammer and condensation-induced water hammer. (author)

Documentation of CATHENA input files for the APOLLO computer

International Nuclear Information System (INIS)

1988-06-01

Input files created for the VAX version of the CATHENA two-fluid code have been modified and documented for simulation on the AECB's APOLLO computer system. The input files describe the RD-14 thermalhydraulic loop, the RD-14 steam generator, the RD-12 steam generator blowdown test facility, the Stern Laboratories Cold Water Injection Facility (CWIT), and a CANDU 600 reactor. Sample CATHENA predictions are given and compared with experimental results where applicable. 24 refs

Manometer Behavior Analysis using CATHENA, RELAP and GOTHIC Codes

Energy Technology Data Exchange (ETDEWEB)

Kim, Yang Hoon; Han, Kee Soo; Moon, Bok Ja; Jang, Misuk [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

2014-05-15

In this presentation, simple thermal hydraulic behavior is analyzed using three codes to show the possibility of using alternative codes. We established three models of simple u-tube manometer using three different codes. CATHENA (Canadian Algorithm for Thermal hydraulic Network Analysis), RELAP (Reactor Excursion and Leak Analysis Program), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are used for this analysis. CATHENA and RELAP are widely used codes for the analysis of system behavior of CANDU and PWR. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. In this paper, the internal behavior of u-tube manometer was analyzed using 3 codes, CATHENA, RELAP and GOTHIC. The general transient behavior is similar among 3 codes. However, the behavior simulated using GOTHIC shows some different trend compared with the results from the other 2 codes at the end of the transient. It would be resulted from the use of different physical model in GOTHIC, which is specialized for the multi-phase thermal hydraulic behavior analysis of containment system unlike the other two codes.

Preparation and documentation of a CATHENA input file for Darlington NGS

International Nuclear Information System (INIS)

1989-03-01

A CATHENA input model has been developed and documented for the heat transport system of the Darlington Nuclear Generating Station. CATHENA, an advanced two-fluid thermalhydraulic computer code, has been designed for analysis of postulated loss-of-coolant accidents (LOCA) and upset conditions in the CANDU system. This report describes the Darlington input model (or idealization), and gives representative results for a simulation of a small break at an inlet header

Calculation of the Thermal Radiation Benchmark Problems for a CANDU Fuel Channel Analysis Using the CFX-10 Code

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyoung Tae; Park, Joo Hwan; Rhee, Bo Wook

2006-07-15

To justify the use of a commercial Computational Fluid Dynamics (CFD) code for a CANDU fuel channel analysis, especially for the radiation heat transfer dominant conditions, the CFX-10 code is tested against three benchmark problems which were used for the validation of a radiation heat transfer in the CANDU analysis code, a CATHENA. These three benchmark problems are representative of the CANDU fuel channel configurations from a simple geometry to whole fuel channel geometry. With assumptions of a non-participating medium completely enclosed with the diffuse, gray and opaque surfaces, the solutions of the benchmark problems are obtained by the concept of surface resistance to radiation accounting for the view factors and the emissivities. The view factors are calculated by the program MATRIX version 1.0 avoiding the difficulty of hand calculation for the complex geometries. For the solutions of the benchmark problems, the temperature or the net radiation heat flux boundary conditions are prescribed for each radiating surface to determine the radiation heat transfer rate or the surface temperature, respectively by using the network method. The Discrete Transfer Model (DTM) is used for the CFX-10 radiation model and its calculation results are compared with the solutions of the benchmark problems. The CFX-10 results for the three benchmark problems are in close agreement with these solutions, so it is concluded that the CFX-10 with a DTM radiation model can be applied to the CANDU fuel channel analysis where a surface radiation heat transfer is a dominant mode of the heat transfer.

Calculation of the Thermal Radiation Benchmark Problems for a CANDU Fuel Channel Analysis Using the CFX-10 Code

International Nuclear Information System (INIS)

Kim, Hyoung Tae; Park, Joo Hwan; Rhee, Bo Wook

2006-07-01

To justify the use of a commercial Computational Fluid Dynamics (CFD) code for a CANDU fuel channel analysis, especially for the radiation heat transfer dominant conditions, the CFX-10 code is tested against three benchmark problems which were used for the validation of a radiation heat transfer in the CANDU analysis code, a CATHENA. These three benchmark problems are representative of the CANDU fuel channel configurations from a simple geometry to whole fuel channel geometry. With assumptions of a non-participating medium completely enclosed with the diffuse, gray and opaque surfaces, the solutions of the benchmark problems are obtained by the concept of surface resistance to radiation accounting for the view factors and the emissivities. The view factors are calculated by the program MATRIX version 1.0 avoiding the difficulty of hand calculation for the complex geometries. For the solutions of the benchmark problems, the temperature or the net radiation heat flux boundary conditions are prescribed for each radiating surface to determine the radiation heat transfer rate or the surface temperature, respectively by using the network method. The Discrete Transfer Model (DTM) is used for the CFX-10 radiation model and its calculation results are compared with the solutions of the benchmark problems. The CFX-10 results for the three benchmark problems are in close agreement with these solutions, so it is concluded that the CFX-10 with a DTM radiation model can be applied to the CANDU fuel channel analysis where a surface radiation heat transfer is a dominant mode of the heat transfer

A critical flow model for the Cathena thermalhydraulic code

International Nuclear Information System (INIS)

Popov, N.K.; Hanna, B.N.

1990-01-01

The calculation of critical flow rate, e.g., of choked flow through a break, is required for simulating a loss of coolant transient in a reactor or reactor-like experimental facility. A model was developed to calculate the flow rate through the break for given geometrical parameters near the break and fluid parameters upstream of the break for ordinary water, as well as heavy water, with or without non- condensible gases. This model has been incorporated in the CATHENA, one-dimensional, two-fluid thermalhydraulic code. In the CATHENA code a standard staggered-mesh, finite-difference representation is used to solve the thermalhydraulic equations. This model compares the fluid mixture velocity, calculated using the CATHENA momentum equations, with a critical velocity. When the mixture velocity is smaller than the critical velocity, the flow is assumed to be subcritical, and the model remains passive. When the fluid mixture velocity is higher than the critical velocity, the model sets the fluid mixture velocity equal to the critical velocity. In this paper the critical velocity at a link (momentum cell) is first estimated separately for single-phase liquid, two- phase, or single-phase gas flow condition at the upstream node (mass/energy cell). In all three regimes non-condensible gas can be present in the flow. For single-phase liquid flow, the critical velocity is estimated using a Bernoulli- type of equation, the pressure at the link is estimated by the pressure undershoot method

Development of CATHENA Plant Model for Wolsong 1 (Rev.0)

Energy Technology Data Exchange (ETDEWEB)

Kim, Y.H.; Lee, K.H.; Choi, H. [Korea Electric Power Research Institute, Taejon (Korea)

2002-07-01

This document includes CATHENA model development of plant controls, steam generator, feedwater and steam piping for wolsong 1 trip coverage analysis in the project of ''Development of Assessment Technologies for CANDU Reactor Power and Trip Effectiveness''. (author). 11 refs., 2 figs., 3 tabs.

On the importance of valve modelling, reflected pressures, and wall friction, in CATHENA water hammer simulations

International Nuclear Information System (INIS)

Beuthe, T.G.

1998-01-01

The results of code and modelling developments outlined in this paper show that CATHENA can be used to accurately model the behaviour of valve slam generated water hammer if sufficient care and detail are used to model the characteristics of the valve. It also shows that CATHENA can accurately predict the reflection and transmission of travelling water pressure waves at expansions, contractions, and dead ends. Finally, although CATHENA is capable of accurately predicting the critical phenomena observed in water hammer, the inter-peak timing of the pressure excursions is not well predicted when significant bulk flows occur. The use of an unsteady wall friction factor to correct for this discrepancy has been examined, but the implementation of relationships suggested in the literature provided too much damping. A good match between experimental and simulation data can be achieved, but it is suggested that the default implementation of such a relationship take place only after an investigation of further potential loss terms can be completed. (author)

Fuel assembly, channel box of fuel assembly, fuel spacer of fuel assembly and method of manufacturing channel box

International Nuclear Information System (INIS)

Chaki, Masao; Kanazawa, Toru; Orii, Akihito; Nagayoshi, Takuji; Nishida, Koji; Kawasaki, Terufumi.

1997-01-01

In a fuel assembly of a BWR type reactor, fuel rods disposed at corners of side walls of a channel box or in the periphery of the side walls are partially removed, and recessed portions are formed on the side walls of the channel box from which the fuel rods are removed. Spaces closed at the sides are formed in the inner side of the corner portions. Openings are formed for communicating the closed space with the outside of the channel box. Then, the channel area of the outer side of the channel box is increased, through which much water flows to increase the amount of water in the reactor core thereby promoting the moderation of neutrons and providing thermal neutrons suitable to nuclear fission. The degree of freedom for distribution of the spaces in the reactor core is increased to improve neutron economy thereby enabling to utilize reactor fuels effectively. (N.H.)

Development and validation of the shutdown cooling system CATHENA model for Gentilly-2

International Nuclear Information System (INIS)

Lecuyer, H.; Hasnaoui, C.; Sabourin, G.; Chapados, S.

2008-01-01

A CATHENA representation of the Gentilly-2 Shutdown Cooling system has been developed for Hydro-Quebec. The model includes the SDCS circuit piping, valves, pumps and heat exchangers. The model is integrated in the G2 CATHENA overall plant model and coupled with the plant control software simulator TROLG2 to allow the simulation of various plant operational modes using the SDCS. Results have been obtained for normal cooling of the primary heat transport system following a planned shut down (transition from full power to shutdown) and for two special SDCS configurations that were used on September 14 and 15, 2006 at Gentilly-2. The results show close match with values measured at Gentilly-2 during either steady or transient states. (author)

Development and validation of the shutdown cooling system CATHENA model for Gentilly-2

Energy Technology Data Exchange (ETDEWEB)

Lecuyer, H.; Hasnaoui, C. [Nucleonex Inc., Westmount, Quebec (Canada); Sabourin, G. [Atomic Energy of Canada Limited, Montreal, Quebec (Canada); Chapados, S. [Hydro-Quebec, Unite Analyse et Fiabilite, Montreal, Quebec (Canada)

2008-07-01

A CATHENA representation of the Gentilly-2 Shutdown Cooling system has been developed for Hydro-Quebec. The model includes the SDCS circuit piping, valves, pumps and heat exchangers. The model is integrated in the G2 CATHENA overall plant model and coupled with the plant control software simulator TROLG2 to allow the simulation of various plant operational modes using the SDCS. Results have been obtained for normal cooling of the primary heat transport system following a planned shut down (transition from full power to shutdown) and for two special SDCS configurations that were used on September 14 and 15, 2006 at Gentilly-2. The results show close match with values measured at Gentilly-2 during either steady or transient states. (author)

Consolidation equipment for irradiated nuclear fuel channels

International Nuclear Information System (INIS)

Taguchi, M.; Komatsu, Y.; Ose, T.

1989-01-01

The authors have developed and put into use a new type of mechanical consolidation equipment for irradiated nuclear fuel channels. This includes round-slice cutting of the top 100mm of the fuel channel with a guillotine cutter, and press cutting of the two corners of the remaining length of the fuel channel. Four guillotine blades work in combination with receiving blades arranged inside the fuel channel to cut the top 100mm, including the clips and spacers, of the fuel channel into a round slice. A press assembled in the consolidation equipment then presses the slice to achieve volume reduction. The press cutting operation uses two press cutting blades arranged inside the fuel channel and the receiving blades outside the fuel channel. The remaining length of fuel channel is cut off into L-shaped pieces by press cutting. This consolidation equipment is highly efficient because the round-slice cutting, pressing, and press cutting are all achieved by one unit

Controlled beta-quench treatment of fuel channels

International Nuclear Information System (INIS)

Moeckel, Andreas; Cremer, Ingo; Kratzer, Anton; Walter, Dirk; Perkins, Richard A.

2007-01-01

The trend towards higher fuel assembly discharge burnups poses new challenges for fuel channels in terms of their dimensional behavior and corrosion resistance. Beta-quenching of fuel channels has been applied by the nuclear industry to improve the dimensional stability of this component. This led AREVA NP to develop a new technique for beta quenching of fuel channels that combines the effect of beta-quenching with the optimization of the microstructure in order to improve the dimensional behavior of fuel channels by randomizing the crystallographic texture, while maintaining the excellent corrosion behavior of the fuel channels by providing intermetallic phase particles of optimum average size. The first fuel channels with these optimized material properties have been placed in the core of a German boiling water reactor (BWR) nuclear power plant in spring of 2004. Some more channels will follow in 2007 to broaden in-pile experience and to receive irradiation feedback from two other nuclear power plants. (authors)

Comparison of the CATHENA model of Gentilly-2 end shield cooling system predictions to station data

Energy Technology Data Exchange (ETDEWEB)

Zagre, G.; Sabourin, G. [Candu Energy Inc., Montreal, Quebec (Canada); Chapados, S. [Hydro-Quebec, Montreal, Quebec (Canada)

2012-07-01

As part of the Gentilly-2 Refurbishment Project, Hydro-Quebec has elected to perform the End Shield Cooling Safety Analysis. A CATHENA model of Gentilly-2 End Shield Cooling System was developed for this purpose. This model includes new elements compared to other CANDU6 End Shield Cooling models such as a detailed heat exchanger and control logic model. In order to test the model robustness and accuracy, the model predictions were compared with plant measurements.This paper summarizes this comparison between the model predictions and the station measurements. It is shown that the CATHENA model is flexible and accurate enough to predict station measurements for critical parameters, and the detailed heat exchanger model allows reproducing station transients. (author)

Conversion of CATHENA from the VAX to APOLLO computer

International Nuclear Information System (INIS)

Kawa, D.M.

1986-12-01

The thermal-hydraulic computer code developed by Atomic Energy of Canada Limited, CATHENA, has been converted from the VAX computer to operate on a low-cost, high-performance APOLLO 3000 work station. Full functionality of the program has been retained. Three verification cases have been analyzed which indicate accuracy to the sixth decimal place as printed in output listings. The APOLLO 3000 computing times for the 3 cases are approximately 15% faster than achieved on the Wardrop VAX 11 750

Water channel reactor fuels and fuel channels: Design, performance, research and development. Proceedings of a technical committee meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-01-01

The International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended holding a Technical Committee Meeting on Water Channel Reactor Fuel including into this category fuels and pressure tubes/fuel channels for Atucha-I and II, BWR, CANDU, FUGEN and RBMK reactors. The IWGFPT considered that even if the characteristics of Atucha, CANDUs, BWRs, FUGEN and RBMKs differ considerably, there are also common features. These features include materials aspects, as well as core, fuel assembly and fuel rod design, and some safety issues. There is also some similarity in fuel power history and operating conditions (Atucha-I and II, FUGEN and RBMK). Experts from 11 countries participated at the meeting and presented papers on technology, performance, safety and design, and materials aspects of fuels and pressure tubes/fuel channels for the above types of water channel reactors. Refs, figs, tabs.

Water channel reactor fuels and fuel channels: Design, performance, research and development. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1998-01-01

The International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended holding a Technical Committee Meeting on Water Channel Reactor Fuel including into this category fuels and pressure tubes/fuel channels for Atucha-I and II, BWR, CANDU, FUGEN and RBMK reactors. The IWGFPT considered that even if the characteristics of Atucha, CANDUs, BWRs, FUGEN and RBMKs differ considerably, there are also common features. These features include materials aspects, as well as core, fuel assembly and fuel rod design, and some safety issues. There is also some similarity in fuel power history and operating conditions (Atucha-I and II, FUGEN and RBMK). Experts from 11 countries participated at the meeting and presented papers on technology, performance, safety and design, and materials aspects of fuels and pressure tubes/fuel channels for the above types of water channel reactors

Subchannel analysis code development for CANDU fuel channel

International Nuclear Information System (INIS)

Park, J. H.; Suk, H. C.; Jun, J. S.; Oh, D. J.; Hwang, D. H.; Yoo, Y. J.

1998-07-01

Since there are several subchannel codes such as COBRA and TORC codes for a PWR fuel channel but not for a CANDU fuel channel in our country, the subchannel analysis code for a CANDU fuel channel was developed for the prediction of flow conditions on the subchannels, for the accurate assessment of the thermal margin, the effect of appendages, and radial/axial power profile of fuel bundles on flow conditions and CHF and so on. In order to develop the subchannel analysis code for a CANDU fuel channel, subchannel analysis methodology and its applicability/pertinence for a fuel channel were reviewed from the CANDU fuel channel point of view. Several thermalhydraulic and numerical models for the subchannel analysis on a CANDU fuel channel were developed. The experimental data of the CANDU fuel channel were collected, analyzed and used for validation of a subchannel analysis code developed in this work. (author). 11 refs., 3 tabs., 50 figs

Assesment of strength and integrity of fuel channels

International Nuclear Information System (INIS)

2000-01-01

Detailed analysis to base strength and integrity of fuel channels was necessary for the licensing process. Description of tasks performed in this direction in 1999 is presented: fuel channel independent strength calculations, assessment of present fuel channels state, analysis of dynamic processes during partial group distribution header rupture, structural integrity analysis of fuel channels located next to broke channel

Application of the SALOME platform to the loose coupling of the CATHENA and ELOCA codes

International Nuclear Information System (INIS)

Zhuchkova, A.

2012-01-01

Use of coupled codes for the safety analysis of nuclear power plants is highly desirable, as it permits multi-disciplinary studies of complex reactor behaviors and, in particular, accident simulations. The present work demonstrates the potential of the SALOME platform as an interface for creating integrated, multi-disciplinary simulations of reactor scenarios. For this purpose two codes currently in use within the Canadian nuclear industry, CATHENA and ELOCA, were coupled by means of SALOME. The coupled codes were used to model the Power Burst Facility (PBF)-CANDU Test, which was to test the thermal-mechanical behavior of PHWR (pressurized heavy water reactor) fuel during a simulated Large Loss-Of-Coolant Accident (LLOCA). The results of the SALOME-coupled simulations are compared with a previous analysis in which the two codes were coupled using a package of scripts. (author)

On the domestic fuel channel for BWR

International Nuclear Information System (INIS)

Fukada, Hiroshi

1979-01-01

Kobe Steel Ltd. started the domestic manufacture of fuel channel boxes for BWRs in 1967, and entered the actual production stage four years after that. Since 1976, the mass production system was adopted with the increase of the demand. The requirements about the surface contamination and the dimensional accuracy over whole length are very strict in the fuel channel boxes, moreover, special consideration must be given so as to prevent the deformation in use. The unique working methods such as electron beam welding, high temperature press forming and so on are employed in Kobe Steel Ltd. to satisfy such strict requirements, therefore the quality of the produced fuel channel boxes is superior to imported ones. At present, the fuel channel boxes domestically made by Kobe Steel Ltd. are used for almost all BWRs in Japan. The functions of fuel channel boxes are to flow boiling coolant uniformly upward, to guide control rods, and to increase the rigidity of fuel assembly. The fuel channel boxes are the square tubes of zircaloy 4 of 134.06 mm inside width, 2.03 mm thickness, and 4118 or 4239 mm length. The progress of the development and the features of the fuel channel boxes and the manufacturing processes are described. Zircaloy plates are formed into channels, and two channels are electron beam-welded after the edge preparation, to make a box. Ultrasonic examination and stress relief treatment are applied, and clips and spacers are welded. (Kako, I.)

Measuring vibrations in fuel channels CNE

International Nuclear Information System (INIS)

Martín Ghiselli, A.; Fiori, J.; Sacchi, M.; Villabrille, G.

2013-01-01

This paper present a description of implementation and execution of vibration measurements made at the request of NUCLEOELECTRICA ARGENTINA S.A. on the ends of the reactor fuel channels of Embalse Nuclear Power Plant to explore possible differences between the dynamic behavior of empty fuel channel and with full charge of fuel elements inside. (author)

CATHENA analysis of CANDU 6 steam generators for steam main break at a remote location

Energy Technology Data Exchange (ETDEWEB)

Liu, T. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

2009-07-01

CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) is a nonequilibrium, two-phase, two fluid network analysis code that has been in use for over two decades by various groups in Canada and around the world. It is the primary Thermalhydraulics network analysis tool used by Atomic Energy Canada Ltd. (AECL) in the design, safety and licensing analysis of power and research reactors as well as test facilities. In the thermalhydraulic model, the liquid and vapor phases may have different pressures, velocities, and temperatures. The objective of the present paper is to present the detailed modeling of a CANDU 6 Steam Generator (SG) using the transient, thermalhydraulics network code CATHENA. The model represents the secondary side, primary side and the main steam system including the main steam line up to the assumed break location. The present model is designed such that the transient pressure drops across Tube Support Plates (TSP) could be extracted. The resistances of degraded/fouled TSPs were modeled by using the increased/reduced flow area of the TSPs. CATHENA then calculates the flow resistance in two-phase flow based on the area contraction/expansion at the TSPs. Three sets of simulations were performed; one with the degraded steam generator data provided by the utility users, and the other two with waterlanced (cleaned with high pressure water jet) TSPs. One run assumed the flow area increased by 25 percent, the other run assumed the flow area increased by 50 percent. on the hot side of the SG. No significant changes in the break discharge rates were observed between the runs. However, the steam generator downcomer flow for the waterlanced case did not reverse during the blowdown as was calculated for the degraded case. As expected, the pressure drop across the TSPs was decreased in the waterlanced cases comparing with degraded cases. The CATHENA simulation provides estimates of the velocity, density, and quality in the tube bundle as well as

Controlled beta-quenching of fuel channels using inert gas

Energy Technology Data Exchange (ETDEWEB)

Moeckel, Andreas; Cremer, Ingo; Kratzer, Anton; Walter, Dirk [AREVA NP (Germany)

2008-07-01

The trend towards higher fuel assembly discharge burnups poses new challenges for fuel channels in terms of their dimensional behavior and corrosion resistance. This led AREVA NP to develop a new technique for beta quenching of fuel channels that combines the effect of beta-quenching with the optimization of the microstructure. The first set of fuel channels with these optimized material properties have been placed in the core of a German boiling water reactor (BWR) nuclear power plant in spring of 2004. Some more channels have been sited in the core of a Scandinavian BWR in fall of 2007 to broaden the in-pile experience with these channels. Dimensional stability is the major requirement that is applied to fuel channels. High corrosion resistance and low hydrogen pickup are certainly required as well. However, corrosion and hydrogen pickup are usually not life limiting factors due to the large wall thickness of the material. Since thick layers of oxide may spall off extensively at high burnup and cause increase of the dose rate for the personnel, high corrosion resistance of fuel channels is mandatory. The fuel channels which surround BWR fuel assemblies are exposed to neutron irradiation as well as to loads induced by the reactor coolant flowing through them. These service conditions induce material growth and creep which cause permanent changes in the dimensions of the channels. Especially, fuel channel bow is of certain interest as increased channel bow may lead to some friction with control blades. Fuel channel bow is mainly induced by fluence gradients. However, there may be additional influences such as oxidation and hydrogen uptake to cause increased channel bow. The effect of hydrogen is currently discussed in the nuclear community to explain the unexpected high fuel channel bow that has been observed in some nuclear power plants. (orig.)

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)

An equipment for the dimensional characterization of irradiated fuel channels

International Nuclear Information System (INIS)

Cederquist, H.

1985-01-01

The reuse of irradiated fuel channels in BWRs is highly beneficial. However, one prerequisite for reuse of a fuel channel is the detailed knowledge of its dimensions, which are affected by irradiation and pressure drop during operation. Therefore an equipment for fast and accurate dimensional measurement of irradiated fuel channels has been developed. The measurements are carried out when the fuel assembly is supported in the same manner as in the reactor core. The equipment utilizes stationary ultrasonic transducers that measure the fuel channel at a number of predetermined axial levels. Measurement data are fed into a computer which calculates the requested dimensional characteristics such as transversal flatness, bow, twist, side perpendicularity etc. Data are automatically printed for subsequent evaluation. Measurements can be performed both when the fuel channel is placed on a fuel bundle and on an empty fuel channel

Detailed CATHENA Model of the Wolsong 1 Pressure and Inventory Control System

Energy Technology Data Exchange (ETDEWEB)

Cha, K.H. [Korea Electric Power Research Institute, Taejon (Korea)

2002-07-01

The Detailed CATHENA model of Wolsong 1 is development to be able to simulate a theramal hydraulic behavior of heat transport system(HTS) Pressure and Inventory Control System(PNIC) at any power operation condition and during transient events such as mall LOCA(small loss of coolant inventory and small breaks in the primary system piping) and non-LOCA(loss of reactivity regulation, loss of flow, loss if Class IV power, loss of PNIC). (author). 12 refs., 7 figs., 6 tabs.

Complete Flow Blockage of a Fuel Channel for Research Reactor

International Nuclear Information System (INIS)

Lee, Byeonghee; Park, Suki

2015-01-01

The CHF correlation suitable for narrow rectangular channels are implemented in RELAP5/MOD3.3 code for the analyses, and the behavior of fuel temperatures and MCHFR(minimum critical heat flux ratio) are compared between the original and modified codes. The complete flow blockage of fuel channel for research reactor is analyzed using original and modified RELAP5/MOD3.3 and the results are compared each other. The Sudo-Kaminaga CHF correlation is implemented into RELAP5/MOD3.3 for analyzing the behavior of fuel adjacent to the blocked channel. A flow blockage of fuel channels can be postulated by a foreign object blocking cooling channels of fuels. Since a research reactor with plate type fuel has isolated fuel channels, a complete flow blockage of one fuel channel can cause a failure of adjacent fuel plates by the loss of cooling capability. Although research reactor systems are designed to prevent foreign materials from entering into the core, partial flow blockage accidents and following fuel failures are reported in some old research reactors. In this report, an analysis of complete flow blockage accident is presented for a 15MW pool-type research reactor with plate type fuels. The fuel surface experience different heat transfer regime in the results from original and modified RELAP5/MOD3.3. By the discrepancy in heat transfer mode of two cases, a fuel melting is expected by the modified RELAP5/MOD3.3, whereas the fuel integrity is ensured by the original code

Modeling two-phase flow in PEM fuel cell channels

Energy Technology Data Exchange (ETDEWEB)

Wang, Yun; Basu, Suman; Wang, Chao-Yang [Electrochemical Engine Center (ECEC), and Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park, PA 16802 (United States)

2008-05-01

This paper is concerned with the simultaneous flow of liquid water and gaseous reactants in mini-channels of a proton exchange membrane (PEM) fuel cell. Envisaging the mini-channels as structured and ordered porous media, we develop a continuum model of two-phase channel flow based on two-phase Darcy's law and the M{sup 2} formalism, which allow estimate of the parameters key to fuel cell operation such as overall pressure drop and liquid saturation profiles along the axial flow direction. Analytical solutions of liquid water saturation and species concentrations along the channel are derived to explore the dependences of these physical variables vital to cell performance on operating parameters such as flow stoichiometric ratio and relative humility. The two-phase channel model is further implemented for three-dimensional numerical simulations of two-phase, multi-component transport in a single fuel-cell channel. Three issues critical to optimizing channel design and mitigating channel flooding in PEM fuel cells are fully discussed: liquid water buildup towards the fuel cell outlet, saturation spike in the vicinity of flow cross-sectional heterogeneity, and two-phase pressure drop. Both the two-phase model and analytical solutions presented in this paper may be applicable to more general two-phase flow phenomena through mini- and micro-channels. (author)

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)

Zirconium ignition in exposed fuel channel

Energy Technology Data Exchange (ETDEWEB)

Elias, E., E-mail: merezra@technion.ac.il; Hasan, D.; Nekhamkin, Y.

2015-05-15

Highlights: • We demonstrate the idea of runaway zirconium–steam reactions in severe accidents in today's LWRs. • We predict the thermal-hydraulics conditions relevant to cladding oxidation in an exposed fuel channel of a partially uncovered core. • The Semenov theory of metal combustion is extended to define a criterion for runaway oxidation reaction in fuel cladding. - Abstract: A theoretical model based on simultaneous solution of the heat and mass transfer equations is developed for predicting the rate of thermo-chemical reaction between zirconium cladding and a hot steam environment. Ignition conditions relevant to cladding oxidation in an exposed fuel channel of a partially uncovered core are predicted based on the theory of metal combustion. A range of decay power, convective heat transfer coefficients, and initial temperatures leading to uncontrolled runaway cladding oxidation is identified. The model could be readily integrated as part of a fuel channel analysis code for predicting possible outcomes of different accident mitigation procedures in light water nuclear reactors under LOCA conditions.

CANDU-PHW fuel channel replacement experience

International Nuclear Information System (INIS)

Dunn, J.T.; Kakaria, B.K.

1982-09-01

One of the main characteristics of the CANDU pressurized heavy water reactor is the use of pressure tubes rather than one large pressure vessel to contain the fuel and coolant. This provides an inherent design capability to permit their replacement in an expeditious manner, without seriously affecting the high capacity factors of the reactor units. Of th eight Ontario Hydro commercial nuclear generating units, the lifetime performance places seven of them (including two that have had some of their fuel channels replaced), in the top ten positions in the world's large nuclear-electric unit performance ranking. Pressure tube cracks in the rolled joint region have resulted in 70 fuel channels being replaced in three reactor units, the latest being at the Bruce Nuclear Generating Station 'A', Unit 2 in February 1982. The rolled joint design and rolling procedures have been modified to eliminate this problem on CANDU units subsequent to Bruce 'A'. This paper describes the CANDU pressure tube performance history and expectations, and the tooling and procedures used to carry out the fuel channel replacement

Fuel Cooling in Absence of Forced Flow at Shutdown Condition with PHTS Partially Drained

Energy Technology Data Exchange (ETDEWEB)

Parasca, L.; Pecheanu, D.L., E-mail: laurentiu.parasca@cne.ro, E-mail: doru.pecheanu@cne.ro [Cernavoda Nuclear Power Plant, Cernavoda (Romania)

2014-09-15

During the plant outage for maintenance on primary side (e.g. for the main Heat Transport System pumps maintenance, the Steam Generators inspection), there are situations which require the primary heat transport system (HTS) drainage to a certain level for opening the circuit. The primary fuel heat sink for this configuration is provided by the shutdown cooling system (SDCS). In case of losing the forced cooling (e.g. due to the loss of SDCS, design basis earthquake-DBE), flow conditions in the reactor core may become stagnant. Inside the fuel channels, natural circulation phenomena known as Intermittent Buoyancy Induced Flow (IBIF) will initiate, providing an alternate heat sink mechanism for the fuel. However, this heat sink is effective only for a limited period of time (recall time). The recall time is defined as the elapsed time until the water temperature in the HTS headers exceeds a certain limit. Until then, compensatory measures need to be taken (e.g. by re-establishing the forced flow or initiate Emergency Core Cooling system injection) to preclude fuel failures. The present paper briefly presents the results of an analysis performed to demonstrate that fuel temperature remains within acceptable limits during IBIF transient. One of the objectives of this analysis was to determine the earliest moment since the reactor shut down when maintenance activities on the HTS can be started such that IBIF is effective in case of losing the forced circulation. The resulting peak fuel sheath and pressure tube temperatures due to fuel heat up shall be within the acceptable limits to preclude fuel defect or fuel channel defects.Thermalhydraulic circuit conditions were obtained using a CATHENA model for the primary side of HTS (drained to a certain level), an ECC system model and a system model for SDCS. A single channel model was developed in GOTHIC code for the fuel assessment analysis. (author)

Fuel assembly and fuel channel box

International Nuclear Information System (INIS)

Sakuma, Toraki; Hirakawa, Hiromasa; Ishizaki, Hideaki; Nakajima, Junjiro; Aizawa, Yasuhiro.

1992-01-01

A fuel channel box has a square cylindrical shape and, in the transversal cross sectional shape, the wall thickness of a corner portion is greater than that of a central portion of the side wall except for an upper portion thereof. The upper portion of the channel box includes a region to be in contact with an upper lattice plate and a region to attach a channel spacer. Then, the wall thickness of these regions is uniform in the transversal cross section and they have the same wall thickness with that of the corner portion which has the increased wall thickness. With such a constitution, the upper portion of the channel box receives a counter force applied from the upper lattice plate upon occurrence of earthquakes and moderate it to reduce local stresses and deformation. Further, a similar region with increased wall thickness is disposed also to the lower portion of the channel box, thereby enabling to suppress the amount of coolants leaked from a portion between the lower portion and a lower tie plate, and improve the mechanical integrity of the channel box. (I.N.)

In-pile behavior of controlled beta-quenched fuel channels

Energy Technology Data Exchange (ETDEWEB)

Moeckel, Andreas; Pflaum, Wolfgang; Cremer, Ingo [AREVA NP GmbH, Erlangen (Germany); Zbib, Ali A. [AREVA NP Inc., Richland, WA (United States)

2011-07-01

Dimensional stability during in-reactor service is the major requirement that is put on fuel channels to provide good moderation and power distribution, and to guarantee unrestricted movement of the control blades during operation. High corrosion resistance and low hydrogen pick-up are required as well. The latter are usually not considered to be life limiting, but may contribute to channel deformation since increased oxide layers due to shadow corrosion on the control blade sides of a channel result in differential oxide thickness and differential volume expansion due to hydride formation. This would be in addition to the well known effects of irradiation induced channel deformation, especially channel growth and bow. In order to meet the trend toward increased fuel assembly discharge burnup levels and the industry wide need for improved dimensional stability of fuel channels, AREVA NP has developed the Controlled Beta-Quenching of fuel channels. The process combines the positive effect of randomization of the crystallographic texture by beta-quenching with the optimization of the microstructure for good corrosion resistance by providing intermetallic phase particles in the optimum size range. The Controlled Beta-Quenching is a continuous heat treatment operation. Its key features are the two-step induction heating to uniformly reach the target temperature, the tight control of the quench rate by cooling the fuel channel from the outer surface using a controlled argon mass flow for quenching, and the protection of the inner surface from oxidation by providing an argon atmosphere. Due to the utilization of argon, the surfaces of the channels remain metal bright after beta-quenching. All in all, the Controlled Beta-Quenching provides an overall 'clean' and environment friendly operation without the need of additional surface conditioning. The first set of beta-quenched fuel channels, exhibiting these optimized material properties, were inserted in the core

Operating performance and reliability of CANDU PHWR fuel channels in Canada

International Nuclear Information System (INIS)

Strachan, B.; Brown, D.R.

1983-03-01

CANDU nuclear plants use many small-diameter high-pressure fuel channels. Good operating performance from the CANDU fuel channels has made a major contribution to the world-leading operating record of the CANDU nuclear power plants. As of 1982 December 31, there were 7,480 fuel channels installed in 18 CANDU reactors over 500 MW(e) in size. Eight of these reactors have been declared in-service and have accumulated 24,000 fuel channel-years of operation. The only significant operating problems with fuel channels have been the occurrence of leaking cracks in 70 fuel channels and a larger amount of axial creep on the early reactors than was originally provided for in the design. Both of these problems have been corrected on all CANDU reactors built since the Bruce GS 'A' station and the newer reactors should exhibit even better performance

Fuel handling alternatives to prepare for large scale fuel channel replacement

International Nuclear Information System (INIS)

Martire, S.; Sandu, I.

2007-01-01

It is desirable to reduce the duration of defuelling the reactor in preparation for retube, as the cost of replacement power is $750K/day. Three fast defuelling concepts are presented. With the Through Flow Defuelling method, the fuel string is hydraulically pushed into the downstream Fuelling Machine (FM) by flow passing through the fuel channel. The Long Stroke C Ram method replaces the FM C Ram with a longer one capable of pushing all fuel bundles into the receiving FM. Defuelling Hardware uses enhanced design of ram extensions that interconnect mechanically to extend the Ram stroke to push fuel bundles into the receiving FM. This paper will present descriptions of each defuelling concept to prepare for Large Scale Fuel Channel Replacement. Advantages and disadvantages of each concept will be discussed and a recommendation will be made for future implementation. (author)

Removal of a seized fuel channel from the KANUPP reactor

International Nuclear Information System (INIS)

Butt, W.M.; Gunn, R.J.

1995-01-01

In support of the Safe operation of KANUPP program, AECL was commissioned in early 1992 to assist the Karachi Nuclear Power Plant in the design and supply of equipment and procedures for removal of a seized fuel channel from the KANUPP CANDU reactor. In addition AECL was also asked to supply technical site support to assist the KANUPP station staff during the removal of the G-12 channel. The design of a fuel channel removal system presented an interesting challenge. The fuel channel design was unique to KANUPP with no history of previous channel removal, consequently nearly all tools and equipment had to be specially designed. In addition, the seized end fitting posed a special problem requiring the development several contingency tools and techniques. This paper is an account of the design and development of the removal system and the site experiences during the actual fuel channel removal. After the channel had been removed, it was confirmed that a corrosion seizure between the end fitting sleeve bearings was inhibiting normal channel elongation. (author)

Fuel channel design improvements for large CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Villamagna, A; Price, E G; Field, G J [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1996-12-31

From the initial designs used in NPD and Douglas point reactors, the CANDU fuel channel and its components have undergone considerable development. Two major designs have evolved: the Pickering/CANDU 6 design which has 12 fuel bundles in the core and where the new fuel is inserted into the inlet end, and the Bruce/Darlington design which has 13 bundles in the channel and where new fuel is inserted into the outlet end. In the development of a single unit CANDU reactor of the size of a Bruce or Darlington unit which would use a Darlington design calandria, the decision has been made to use the CANDU 6 fuel channel rather than the Darlington design. The CANDU 6 channel has provided excellent performance and will not encounter the degree of maintenance required for the Bruce/Darlington design. The channel design in turn influences the fuelling machine/fuel handling concepts required. The changes to the CANDU 6 fuel channel design to incorporate it in the large unit are small. In fact, the changes that are proposed relate to the desire to increase margins between pressure tube properties and design conditions or ameliorate the consequences of postulated accident conditions, rather than necessary adaptation to the larger unit. Better properties have been achieved in the pressure tube material resulting from alloy development program over the past 10 years. Pressure tubes can now he made with very low hydrogen concentrations so that the hydrogen picked up as deuterium will not exceed the terminal solid solubility for the in-core region in 30 years. The improvements in metal chemistry allow the production of high toughness tubes that retain a high level of toughness during service. A small increase in wall thickness will reduce the dimensional changes without significantly affecting burnup. Changes to increase safety margins from postulated accidents are concentrated on containing the consequences of pressure tube damage. The changes are concentrated on the calandria tube

Improving the service life and performance of CANDU fuel channels

International Nuclear Information System (INIS)

Coleman, C.E.; Cheadle, B.A.; Causey, A.R.; Doubt, G.L.; Fong, R.W.L.; Venkatapathi, S.

1996-03-01

The development objective for CANDU fuel channels is to produce a design that can operate for 40 years at 90% capacity. Steady progress toward this objective is being made. The factors that determine the life of a CANDU fuel channel are reviewed and the processes necessary to achieve the objectives are identified. Performance of future fuel channels will be enhanced by reduced operating costs and increased safety margins to postulated accident conditions compared with those for current channels. The approaches to these issues are discussed briefly in this report. (author)

Finite element analysis of BWR fuel channel buckling during a seismic event

International Nuclear Information System (INIS)

Kinoshita, Mika; Iwamoto, Yuji; Ledford, Kevin; Cantonwine, Paul

2014-01-01

This paper documents the predicted response of three BWR fuel channel designs in bending using a typical moment profile for GNF fuel designs. The bending performance of the fuel channel is predicted using ANSYS, a finite element modeling tool. Specifically, linear and non-linear buckling analyses were performed to determine the onset of elastic buckling, which causes a wavy structure on the compression face in bending that might also increase channel – control blade friction, and to determine to onset of channel collapse, which causes permanent deformation and would inhibit control rod insertion. The three channel designs considered in this paper are the 0.080 inch uniform channel, the 0.100 inch uniform channel and the 0.120 inch uniform channel at the beginning of fuel life (BOL) and at the end of fuel life (EOL). (author)

Improving the service life and performance of CANDU fuel channels

International Nuclear Information System (INIS)

Causey, A.R.; Cheadle, B.A.; Coleman, C.E.; Price, E.G.

1996-02-01

The development objective for CANDU fuel channels is to produce a design that can operate for 40 years at 90% capacity. Steady progress toward this objective is being made. The factors that determine the life of a CANDU fuel channel are reviewed and the processes necessary to achieve the objectives identified. Performance of future fuel channels will be enhanced by reduced operating costs, increased safety margins to postulated accident conditions, and reduced retubing costs compared with those for current channels. The approaches to these issues are discussed briefly in this report. (author). 14 refs., 1 tab., 8 figs

Inspection of Candu Nuclear Reactor Fuel Channels

International Nuclear Information System (INIS)

Baron, J.; Jarvis, G.N.; Dolbey, M.P.; Hayter, D.M.

1986-01-01

The Channel Inspection and Gauging Apparatus of Reactors (CIGAR) is a fully atomated, remotely operated inspection system designed to perform multi-channel, multi-task inspection of CANDU reactor fuel channels. Ultrasonic techniques are used for flaw detection, (with a sensitivity capable of detecting a 0.075 mm deep notch with a signal to noise ratio of 10 dB) and pressure tube wall thickness and diameter measurements. Eddy currrent systems are used to detect the presence of spacers between the coaxial pressure tube and calandria tube, as well as to measure their relative spacing. A servo-accelerometer is used to estimate the sag of the fuel channels. This advanced inspection system was commissioned and declared in service in September 1985. The paper describes the inspection systems themselves and discussed the results achieved to-date. (author) [pt

Performance monitoring of zircaloy-4 square fuel channels at TAPS-1 and 2

International Nuclear Information System (INIS)

Akhtar, J.; Ramu, A.; Anilkumar, K.R.; Sharma, B.L.; Bhattacharjee, S.; Ramamurty, U.; Srivastava, S.P.; Prasad, P.N.; Anantharaman, K.

2006-01-01

Tarapur Atomic Power Station is a twin unit Boiling Water Reactors. The initial rated capacity of each unit was 210 MWe. Subsequently due to Secondary Steam Generator tube leak problem, the units were de-rated to 160 MWe in the year 1984-85. The station has completed 36 years of successful commercial operation. TAPS reactor fuel channels are made of Zircaloy-4, material. These are used along with 6x6 array nuclear fuel assemblies. The fuel channels need to be discharged once it reaches an optimum exposure limit and based on the surveillance programme, which monitors the channels performance. NFC has indigenously developed fuel channels for TAPS and these are at various stages of exposure in both the reactor cores. The performance review of these channels was carried out by the experts from TAPS-Site, NPCIL-ED and RED, BARC. The two major factors, which affect fuel channels performance, are (a) Bulge and (b) Bow. The phenomenon of longitudinal bow occurs due to the neutron flux gradient across the channels faces. Studies made on this subject by General Electric (GE) indicated that this channel deflection occurs at a slow rate. Therefore, fuel channels surveillance programme is essential to check the irradiated fuel channels performance in order to replace the fuel channels once it reaches the optimum exposure limit. To estimate the useful life of irradiated fuel channels, channel deflection/bulge measurement inspection system and methodology was developed jointly by TAPS and Centre for Design and manufacture (CDM), BARC. This system was successfully deployed at TAPS. This paper briefly describes the developmental efforts made by Nuclear Fuel Complex (NFC), Hyderabad, NPCIL-Fuel Group, Engg.Directorate, RED/BARC, CDM/BARC. (author)

Modeling approach for annular-fuel elements using the ASSERT-PV subchannel code

International Nuclear Information System (INIS)

Dominguez, A.N.; Rao, Y.

2012-01-01

The internally and externally cooled annular fuel (hereafter called annular fuel) is under consideration for a new high burn-up fuel bundle design in Atomic Energy of Canada Limited (AECL) for its current, and its Generation IV reactor. An assessment of different options to model a bundle fuelled with annular fuel elements is presented. Two options are discussed: 1) Modify the subchannel code ASSERT-PV to handle multiple types of elements in the same bundle, and 2) coupling ASSERT-PV with an external application. Based on this assessment, the selected option is to couple ASSERT-PV with the thermalhydraulic system code CATHENA. (author)

Fuel temperature analysis method for channel-blockage accident in HTTR

International Nuclear Information System (INIS)

Maruyama, So; Fujimoto, Nozomu; Sudo, Yukio; Kiso, Yoshihiro; Hayakawa, Hitoshi

1994-01-01

During operation of the High Temperature Engineering Test Reactor (HTTR), coolability must be maintained without core damage under all postulated accident conditions. Channel blockage of a fuel element was selected as one of the design-basis accidents in the safety evaluation of the reactor. The maximum fuel temperature for such a scenario has been evaluated in the safety analysis and is compared to the core damage limits.For the design of the HTTR, an in-core thermal and hydraulic analysis code ppercase[flownet/trump] was developed. This code calculates fuel temperature distribution, not only for a channel blockage accident but also for transient conditions. The validation of ppercase[flownet/trump] code was made by comparison of the analytical results with the results of thermal and hydraulic tests by the Helium Engineering Demonstration Loop (HENDEL) multi-channel test rig (T 1-M ), which simulated one fuel column in the core. The analytical results agreed well with the experiments in which the HTTR operating conditions were simulated.The maximum fuel temperature during a channel blockage accident is 1653 C. Therefore, it is confirmed that the integrity of the core is maintained during a channel blockage accident. ((orig.))

Fuel channel closure and adapter

International Nuclear Information System (INIS)

Cashen, W.S.

1985-01-01

This invention provides a mechanical closure/actuating ram combination particularly suited for use in sealing the ends of the pressure tubes when a CANDU-type reactor is refueled. It provides a cluster that may be inserted into a fuel channel end fitting to provide at least partial closing off of a pressure tube while permitting the disengagement of the fueling machine and its withdrawal from the closure for other purposes. The invention also provides a ram/closure combination wherein the application of loading force to a deformable sealing disk is regulated by a massive load bar component forming part of the fueling machine and being therefore accessible for maintenance or replacement

Detailed channel thermal-hydraulic calculation of nuclear reactor fuel assemblies

International Nuclear Information System (INIS)

Zhukov, A.V.; Sorokin, A.P.; Ushakov, P.A.; Yur'ev, Yu.S.

1981-01-01

The system of equations of mass balance, quantity of motion and energy used in calculation of nuclear reactor fuel assemblies is obtained. The equation system is obtained on the base of integral equations of hydrodynamics interaction in assemblies of smooth fuel elements and fuel elements with wire packing. The calculation results of coolant heating distributions by the fast reactor assembly channels are presented. The analysis of the results obtained shows that interchannel exchange essentially uniforms the coolant heating distribution in the peripheral range of the assembly but it does not remove non-uniformity caused by power distribution non-uniformity in the cross section. Geometry of the peripheral assembly range plays an essential role in the heating distribution. Change of the calculation gap between the peripheral fuel elements and assembly shells can result either in superheating or in subcooling in the peripheral channels relatively to joint internal channels of the assembly. Heat supply to the coolant passing through interassembly gaps decreases temperature in the assembly periphery and results in the increase of temperature non-uniformity by the perimeter of peripheral fuel elements. It is concluded that the applied method of the channel-by-channel calculation is ef-- fective in thermal-physical calculation of nuclear reactor fuel assemblies and it permits to solve a wide range of problems [ru

Manufacturing method of fuel assembly and channel box for the fuel assembly

International Nuclear Information System (INIS)

Fujieda, Tadashi; Inagaki, Masatoshi; Takase, Iwao; Nishino, Yoshitaka; Yamashita, Jun-ichi; Yamanaka, Akihiro; Ito, Ken-ichi; Nakajima, Junjiro; Seto, Takehiro.

1998-01-01

An MOX fuel assembly to be used for a BWR type reactor comprises a channel box, a great number of fuel rod bundles and a water rod. BP members incorporated with a burnable neutron absorbing poison (BP) are buried in the vicinity of corners of four sides of the channel box in the longitudinal direction. The channel box is formed by fitting the BP members in concaves formed in the longitudinal direction of zircaloy plates, laminating other zircaloy plates and welding the seams. Then, hot rolling, cold rolling and annealing are conducted to form them into a single plate. Integrated two single plates after bending treatment are abutted and welded, and heat-treatment is applied to complete the channel box. With such a constitution, since the BP member is not brought into contact with reactor water directly, crevice corrosion or galvanic corrosion can be prevented. (I.N.)

Molten Fuel Mass Assessment for Channel Flow Blockage Event in CANDU6

International Nuclear Information System (INIS)

Lee, Kwang Ho; Kim, Yong Bae; Choi, Hoon; Park, Dong Hwan

2011-01-01

In CANDU6, a fuel channel flow blockage causes a sudden reduction of flow through the blocked channel. Depending on the severity of the blockage, the reduced flow through the channel can result in severe heat up of the fuel, hence possibly leading to pressure tube and calandria tube failure. If the calandria tube does not fail the fuel and sheath would continue to heat up, and ultimately melting could occur. Eventually, molten material runs down onto the pressure tube. Even a thin layer of molten material in contact with the pressure tube causes the pressure tube and calandreia tube to heat up rapidly. The thermal transient is so rapid that failure temperatures are reached quickly. After channel failure, the contents of the channel, consisting of superheated coolant, fission products and possibly overheated of molten fuel, are rapidly discharged into the moderator. Fuel discharged into the moderator is quenched and cooled. The rapid discharge of hot fuel and coolant into the calandria causes the moderator pressure and temperature to increase, which may cause damage to some in-core components. Thus, the assessment results of molten fuel mass are inputs to the in-core damage analysis. In this paper, the analysis methodology and results of molten fuel mass assessment for the channel flow blockage event are presented

Aspects regarding the lifetime of a fuel channel in a CANDU nuclear power plant

International Nuclear Information System (INIS)

Calinescu, A.

1998-01-01

The paper presents the analysis of factors influencing upon the time life of a fuel channel of CANDU reactors built in Romania. Fuel channels are made of Zr-2.5%Nb alloy. Means and methodology to detect cracking of fuel channels are described, as well as improvements to increase life time of Cernavoda NPP fuel channels and national programme in this area. (author)

Improving the service life and performance of CANDU fuel channels

International Nuclear Information System (INIS)

Causey, A.R.; Cheadle, B.A.; Coleman, C.E.; Price, E.G.

1997-01-01

The development objective for CANDU fuel channels is to produce a design that can operate for 40 years at 90% capacity. Steady progress toward this objective is being made. The factors that determine the life of the channel are reviewed and the processes necessary to achieve the objectives identified. Performance of future fuel channels will be enhanced by reduced operating costs, increased safety margins to postulated accident conditions, and reduced retubing costs compared to current channels. The approaches to these issues are discussed briefly in the paper. (author)

Establishment of Safety Analysis System and Technology for CANDU Reactors

International Nuclear Information System (INIS)

Min, Byung Joo; Kim, W. Y.; Kim, H. T.; Rhee, B. W.; Yoon, C.; Kang, H. S.; Yoo, K. J.

2005-03-01

To improve the CANDU design/operation safety analysis codes and the CANDU safety analysis methodology, the following works have been done. From the development of the lattice codes (WIMS/CANDU), the lattice model simulates the real core lattice geometry and the effect of the pressure tube creep to the core lattice parameter has been evaluated. From the development of the 3-dimensional thermal-hydraulic analysis model of the moderator behavior (CFX4-CAMO), validation of the model against STERN Lab experiment has been executed. The butterfly-shaped grid structure and the 3-dimensional flow resistance model for porous media were developed and applied to the moderator analysis for Wolsong units 2/3/4. The single fuel channel analysis codes for blowdown and post-blowdown were unified by CATHENA. The 3-dimensional fuel channel analysis model (CFX-CACH) has been developed for validation of CATHENA fuel channel analysis model. The interlinking analysis system (CANVAS) of the thermal-hydraulic safety analysis codes for the primary heat transport system and containment system has been executed. The database system of core physics and thermal-hydraulics experimental data for safety analysis has been established on the URL: http://CANTHIS.kaeri.re.kr. For documentation and Standardization of the general safety analysis procedure, the general safety analysis procedure is developed and applied to a large break LOCA. The present research results can be utilized for establishment of the independent safety analysis technology and acquisition of the optimal safety analysis technology

Method for using a channel box for fuel assemblies in a reactor

International Nuclear Information System (INIS)

Ito, Ken-ichi; Shinpo, Katsutoshi; Watahiki, Minoru.

1975-01-01

Object: To extend a service life of a channel box used in a light water nuclear reactor. Structure: A channel box mounted in the outer periphery of a fuel bundle is removed from the fuel bundle after use for a predetermined period of time, and the removed channel box is re-mounted on the fuel bundle with opposite ends of the box reversed in position for further use. By this arrangement, structural deformation of the channel box may be minimized to extend the service life of the box. (Kamimura, M.)

Analysis of LOCA/LOECC with a non-stop CATHENA simulation

International Nuclear Information System (INIS)

Sabourin, G.; Huynh, H.M.

1997-01-01

This paper documents a new approach which simulates without interruption the blowdown and the post-blowdown portions of a LOCA/LOECC. The blowdown portion is simulated first with the pressures, enthalpies, and void fractions of the headers as boundary conditions. The transient inlet header flowrates are written to a file. The blowdown portion is then simulated again with the inlet header flowrates as boundary conditions. At the end of the blowdown, the flowrates are gradually changed to obtain the desired constant gas flowrate of the post-blowdown portion. This new approach was applied with CATHENA MOD3.5a Rev. 0 for a 20% reactor inlet header break coincident with a total loss of emergency core cooling injection. In summary, this paper shows a successful new approach where the blowdown and the post-blowdown portions of a large LOCA coincident with a total loss of emergency core cooling are simulated continuously. (author)

Simulation of a turbine trip transient at Embalse NPP with full-circuit CATHENA model

Energy Technology Data Exchange (ETDEWEB)

Rabiti, A., E-mail: arabiti@na-sa.com.ar [Nucleoelectrica Argentina S.A., Embalse Nuclear Power Plant, Engineering Management Branch, Embalse (Argentina); Parrondo, A., E-mail: aparrondo@na-sa.com.ar [Nucleoelectrica Argentina S.A., Engineering Management, Buenos Aires (Argentina); Serrano, P., E-mail: pserrano@na-sa.com.ar [Nucleoelectrica Argentina S.A., Licensing Coordination Branch, Atucha II Project Branch (Unidad de Gestion), Buenos Aires (Argentina); Sablayrolles, A.; Damiani, H., E-mail: asablayrolles@na-sa.com.ar, E-mail: hdamiani@na-sa.com.ar [Nucleoelectrica Argentina S.A., Embalse Nuclear Power Plant, Embalse Life Extension Project Management, Embalse (Argentina)

2015-07-01

Embalse NPP is carrying on a Periodic Safety Review to deal with its life extension. This review includes tasks like Deterministic Analysis review for the Final Safety Analysis Report. In 2011, NA-SA (Nucleoelectrica Argentina S.A.) issued a first CATHENA full-circuit model representing the current plant. This model is used in this work. The simulation presented here corresponds to a turbine trip that occurred at Embalse NPP. Consistency between the simulation and the real event is demonstrated. Furthermore, NASA is currently performing Safety Analysis with a new model developed jointly with AECL and Candu Energy which includes post refurbishment changes and other improvements. (author)

Fuel channel in-service inspection programs program design for maximum cost effectiveness

International Nuclear Information System (INIS)

Van den Brekel, N.C.

1995-01-01

Inspection is an integral part of fuel channel life management strategy. Inspection data is used to assess the state of reactor core integrity and provide the information necessary to optimize long term maintenance programs. This paper will provide an overview of the structured approach to developing fuel channel inspection programs within OHN. The inspection programs are designed to balance the resources utilized (cost, outage time, and dose expenditure) with the benefits provided by the inspection data obtained (improved knowledge of component status, degradation mechanisms and rates, etc..). The CANDU community has yet to have a fuel channel operate for a full 30 year design life. Since research programs can not fully simulate reactor operating conditions, inspections become an essential feature of the life management strategy as the components age. Inspection programs often include activities designed to develop predictive capability for long term fuel channel behaviour and provide early warning of changes in behaviour. It should be noted that although this paper addresses the design of fuel channel inspection programs, the basic principles presented can be applied to the design of inspection programs for any major power plant component or system. (author)

Analysis of systematic error deviation of water temperature measurement at the fuel channel outlet of the reactor Maria

International Nuclear Information System (INIS)

Bykowski, W.

2000-01-01

The reactor Maria has two primary cooling circuits; fuel channels cooling circuit and reactor pool cooling circuit. Fuel elements are placed inside the fuel channels which are parallely linked in parallel, between the collectors. In the course of reactor operation the following measurements are performed: continuous measurement of water temperature at the fuel channels inlet, continuous measurement of water temperature at the outlet of each fuel channel and continuous measurement of water flow rate through each fuel channel. Based on those thermal-hydraulic parameters the instantaneous thermal power generated in each fuel channel is determined and by use of that value the thermal balance and the degree of fuel burnup is assessed. The work contains an analysis concerning estimate of the systematic error of temperature measurement at outlet of each fuel channel and so the erroneous assessment of thermal power extracted in each fuel channel and the burnup degree for the individual fuel element. The results of measurements of separate factors of deviations for the fuel channels are enclosed. (author)

Numerical study of the thermo-hydraulic behavior for the Candu type fuel channel

International Nuclear Information System (INIS)

Lazaro, Pavel Gabriel; Balas Ghizdeanu, Elena Nineta

2008-01-01

Candu type reactors use fuel channel in a horizontal lattice. The fuel bundles are positioned in two Zircaloy tubes: the pressure tube surrounded by calandria tube. Inside the pressure tube the coolant heavy water flows. The coolant reaches high temperatures and pressures. Due to irregular neutron spatial distribution, the fuel channel stress differs from one channel to other. In one improbable event of severe accident, the fuel channel behaves differently according to its normal function history. Over the years, there have been many research projects trying to analyze thermal hydraulic performance of the design and to add some operational improvements in order to achieve an efficient thermal hydraulic distribution. This paper discusses the thermo hydraulic behavior (influence of the temperature and velocity distribution) of the most solicited channel, simulated with Fluent 6.X. Code. Moreover it will be commented the results obtained using different models and mesh applied. (authors)

Technology channel fuel cells; Reseau technologique piles a combustible

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-07-01

This document presents the PACo channel, its research and development program and the calendar of the first year. The PACo channel aims at stimulate the technology innovation in the domain of the fuel cells and organize collaborations between enterprises and research laboratories. (A.L.B.)

Research on nondestructive examination methods for CANDU fuel channel inspection

International Nuclear Information System (INIS)

Soare, M.; Petriu, F.; Toma, V.; Revenco, V.; Calinescu, A.; Ciocan, R.; Iordache, C.; Popescu, L.; Mihalache, M.; Murgescu, C.

1995-01-01

The requirements of the 1994 edition of CAN/CSA-N285.4 Periodic Inspection Standard, which address all known and postulated degradation mechanisms and introduce material surveillance demands, involve a growing need for improved nondestructive examination (NDE) methods and technologies. In order to have a proper technical support in its decisions concerning fuel channel inspections at Cernavoda NPP, the Romanian Power Authority (RENEL) initiated a Research Program regarding the nondestructive characterization of the fuel channels structural integrity. The paper presents the most significant results obtained on this Research Program: the ENDUS experimental system for Laboratory simulation of the fuel channel inspection, ultrasonic Rayleigh-Lamb waves technique for pressure tubes examination, phase analysis technique for near-surface flaws, influence of the metallurgical state of the pressure tube material on the eddy current defectoscopic signals, characterization of plastic deformation and fracture of zirconium alloys by acoustic emission. (author)

Methodology, status and plans for development and assessment of TUF and CATHENA codes

Energy Technology Data Exchange (ETDEWEB)

Luxat, J.C.; Liu, W.S.; Leung, R.K. [Ontario Hydro, Toronto (Canada)] [and others

1997-07-01

An overview is presented of the Canadian two-fluid computer codes TUF and CATHENA with specific focus on the constraints imposed during development of these codes and the areas of application for which they are intended. Additionally a process for systematic assessment of these codes is described which is part of a broader, industry based initiative for validation of computer codes used in all major disciplines of safety analysis. This is intended to provide both the licensee and the regulator in Canada with an objective basis for assessing the adequacy of codes for use in specific applications. Although focused specifically on CANDU reactors, Canadian experience in developing advanced two-fluid codes to meet wide-ranging application needs while maintaining past investment in plant modelling provides a useful contribution to international efforts in this area.

Methodology, status and plans for development and assessment of TUF and CATHENA codes

International Nuclear Information System (INIS)

Luxat, J.C.; Liu, W.S.; Leung, R.K.

1997-01-01

An overview is presented of the Canadian two-fluid computer codes TUF and CATHENA with specific focus on the constraints imposed during development of these codes and the areas of application for which they are intended. Additionally a process for systematic assessment of these codes is described which is part of a broader, industry based initiative for validation of computer codes used in all major disciplines of safety analysis. This is intended to provide both the licensee and the regulator in Canada with an objective basis for assessing the adequacy of codes for use in specific applications. Although focused specifically on CANDU reactors, Canadian experience in developing advanced two-fluid codes to meet wide-ranging application needs while maintaining past investment in plant modelling provides a useful contribution to international efforts in this area

CATHENA Analysis Of Candu Advanced Passive Moderator Concept In Normal Operation Condition

International Nuclear Information System (INIS)

Alfa, Sudjatmi K

2001-01-01

In the CANDU - advanced passive moderator (APM) concept, the positive void reactivity is eliminated by reducing the density of the moderator. The simple model for the CANDU APM concept consists of the calandria, heat exchanger, pump, and a stabilizing tank, along with connecting piping. The calandria is divided into two parts, one part simulates the down area, while the other simulates up flow area. To demonstrate the thermalhydraulic behavior of the APM concept, Canadian algorithm for thermalhydraulic network analysis (CATHENA) code is used. The simulation for a pressure boundary condition of 300, 330 and 360 kPa and for water coolant mass flow rate boundary conditions of 2000 and 3000 kg/s respectively have been studied. Preliminary results show that there is boiling in the core, with vapor condensing in the heat exchanger. It is important to note, that the solution had not reached steady state when the boiling occurred

Water droplet accumulation and motion in PEM (Proton Exchange Membrane) fuel cell mini-channels

International Nuclear Information System (INIS)

Carton, J.G.; Lawlor, V.; Olabi, A.G.; Hochenauer, C.; Zauner, G.

2012-01-01

Effective water management is one of the key strategies for improving low temperature PEM (Proton Exchange Membrane) fuel cell performance and durability. Phenomena such as membrane dehydration, catalyst layer flooding, mass transport and fluid flow regimes can be affected by the interaction, distribution and movement of water in flow plate channels. In this paper a literature review is completed in relation to PEM fuel cell water flooding. It is clear that droplet formation, movement and interaction with the GDL (Gas Diffusion Layer) have been studied extensively. However slug formation and droplet accumulation in the flow channels has not been analysed in detail. In this study, a CFD (Computational Fluid Dynamic) model and VOF (Volume of Fluid) method is used to simulate water droplet movement and slug formation in PEM fuel cell mini-channels. In addition, water slug visualisation is recorded in ex situ PEM fuel cell mini-channels. Observation and simulation results are discussed with relation to slug formation and the implications to PEM fuel cell performance. -- Highlights: ► Excess water in mini-channels from the collision and coalescence of droplets can directly form slugs in PEM fuel cells. ► Slugs can form at low flow rates so increasing the flow rate can reduce the size and frequency of slugs. ► One channel of a double serpentine mini-channel may become blocked due to the redistribution of airflow and pressure caused by slug formation. ► Correct GDL and mini-channel surface coatings are essential to reduce slug formation and stagnation. ► Having geometry changes (bends and steps) in the flow fields can disrupt slug movement and avoid channel blockages.

Estimation of Aging Effects on LOHS for CANDU-6

Energy Technology Data Exchange (ETDEWEB)

Yoon, Yong Ki; Moon, Bok Ja; Kim, Seoung Rae [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

2014-05-15

To evaluate the Wolsong Unit 1's capacity to respond to large-scale natural disaster exceeding design, the loss of heat sink(LOHS) accident accompanied by loss of all electric power is simulated as a beyond design basis accident. This analysis is considered the aging effects of plant as the consequences of LOHS accident. Various components of primary heat transport system(PHTS) get aged and some of the important aging effects of CANDU reactor are pressure tube(PT) diametral creep, steam generator(SG) U-tube fouling, increased feeder roughness, and feeder orifice degradation. These effects result in higher inlet header temperatures, reduced flows in some fuel channels, and higher void fraction in fuel channel outlets. Fresh and aged models are established for the analysis where fresh model is the circuit model simulating the conditions at retubing and aged model corresponds to the model reflecting the aged condition at 11 EFPY after retubing. CATHENA computer code[1] is used for the analysis of the system behavior under LOHS condition. The LOHS accident is analyzed for fresh and aged models using CATHENA thermal hydraulic computer code. The decay heat removal is one of the most important factors for mitigation of this accident. The major aging effect on decay heat removal is the reduction of heat transfer efficiency by steam generator. Thus, the channel failure time cannot be conservatively estimated if aged model is applied for the analysis of this accident.

Analysis of IFR driver fuel hot channel factors

International Nuclear Information System (INIS)

Ku, J.Y.; Chang, L.K.; Mohr, D.

1994-01-01

Thermal-hydraulic uncertainty factors for Integral Fast Reactor (IFR) driver fuels have been determined based primarily on the database obtained from the predecessor fuels used in the IFR prototype, Experimental Breeder Reactor II. The uncertainty factors were applied to the channel factors (HCFs) analyses to obtain separate overall HCFs for fuel and cladding for steady-state analyses. A ''semistatistical horizontal method'' was used in the HCFs analyses. The uncertainty factor of the fuel thermal conductivity dominates the effects considered in the HCFs analysis; the uncertainty in fuel thermal conductivity will be reduced as more data are obtained to expand the currently limited database for the IFR ternary metal fuel (U-20Pu-10Zr). A set of uncertainty factors to be used for transient analyses has also been derived

Analysis of IFR driver fuel hot channel factors

International Nuclear Information System (INIS)

Ku, J.Y.; Chang, L.K.; Mohr, D.

2004-01-01

Thermal-hydraulic uncertainty factors for Integral Fast Reactor (IFR) driver fuels have been determined based primarily on the database obtained from the predecessor fuels used in the IFR prototype. Experimental Breeder Reactor II. The uncertainty factors were applied to the hot channel factors (HCFs) analyses to obtain separate overall HCFs for fuel and cladding for steady-state analyses. A 'semistatistical horizontal method' was used in the HCFs analyses. The uncertainty factor of the fuel thermal conductivity dominates the effects considered in the HCFs analysis; the uncertainty in fuel thermal conductivity will be reduced as more data are obtained to expand the currently limited database for the IFR ternary metal fuel (U-20Pu-10Zr). A set of uncertainty factors to be used for transient analyses has also been derived. (author)

Advances in fuel channel technology for CANDU reactors

International Nuclear Information System (INIS)

Cheadle, B.A.; Coleman, C.E.

1994-05-01

The components of the CANDU fuel channels are being developed to have service lives of over 30 years with large margins of safety. Information from research programs and the examination of components removed from reactors has enable improvements to be made to pressure tubes, spacers, calandria tubes and end fittings. Improvements have also been made to the channel design to facilitate planned retubing. (author). 22 refs., 5 tabs., 31 figs

Impact of Zr + 2.5% Nb alloy corrosion upon operability of RBMK-1000 fuel channels

International Nuclear Information System (INIS)

Kovyrshin, V.; Zaritsky, N.

1999-01-01

The basic components of RBMK-1000 core (fuel channels, bimetal adapters, claddings of fuel elements, etc.) are of zirconium alloys. Their corrosion is one of factors influencing upon fuel channels operability. Dynamics of channel tubes nodular corrosion development is presented by the results of in-reactor investigation at ChNPP. Radiation-induced mechanism of corrosion damage of tubes surface in contact with coolant was formulated and substantiated by data of post-reactor studies. Within the certain time period of operation corrosion of zirconium alloy of lower bimetal adapter along with removal from there of corrosion products are predominant within the whole process of reactor elements corrosion. The experimental and calculating method was proposed and substantiated to predict time duration up to loss of fuel channels leak tightness. The approaches were generalized to control state of fuel channels material to assess their operability under operation of RBMK-1000 reactors. (author)

Mechanistic study of fuel freezing, channel plugging, and continued coolability during fast reactor overpower excursions

International Nuclear Information System (INIS)

Wong, K.W.; Catton, I.; Kastenberg, W.E.

1977-07-01

A mechanistic model is presented which describes events following fuel pin failure which may lead to in-channel fuel plate-out. The thermal and hydraulic effects of the plate-out fuel are also evaluated. Given the amount and particle size of the fuel injected into the coolant channel during fuel pin failure, and the initial conditions of the interaction zone, the physical states of the fuel particles and the coolant in the interaction zone can be determined. The trajectories of the fuel particles in the coolant channel are determined by assuming a slip factor between the local tangential velocities of the coolant and the fuel particles. The time and distance after which a fuel particle hits a wire wrap are then determined and the impact stresses induced in the thin solid fuel crust can be evaluated

Experimental fuel channel for samples irradiation at the RB reactor

International Nuclear Information System (INIS)

Pesic, M.; Markovic, H.; Sokcic-Kostic, M.; Miric, I.; Prokic, M.; Strugar, P.

1984-12-01

An 80% enriched UO 2 fuel channel at the RB nuclear reactor in the 'Boris Kidric' Institute of Nuclear Sciences is modified for samples irradiation by fast neutrons. Maximum sample diameter is 25 mm and length up to 1000 mm. Characteristics of neutron and gamma radiation fields of this new experimental channel are investigated. In the centre of the channel, the main contribution to the total neutron absorbed dose, i.e. 0.29 Gy/Wh of reactor operation, is due to the fast neutron spectrum component. Only 0.05 Gy and 0.07 Gy in the total neutron absorbed dose are due to intermediate and thermal neutrons, respectively. At the same time the gamma absorbed dose is 0.35 Gy. The developed experimental fuel channel, EFC, has wide possibilities for utilization, from fast neutron spectrum studies, electronic component irradiations, dosemeters testing, up to cross-section measurements. (author)

Modified fuel channel for sample irradiation at the RB reactor

International Nuclear Information System (INIS)

Pesic, M.; Markovic, H.; Sokcic, M.; Miric, I.; Prokic, M.; Strugar, P.

1983-01-01

Fuel channel of 80% enriched UO 2 at RB reactor in Boris Kidric Institute of nuclear sciences is modified for sample irradiation in the fast neutron field. Maximum sample diameter is 25 mm and length up to 100 mm. Characteristics of neutron as well as gamma radiation fields of this new experimental channel are investigated. In the center of channel, the main contribution to the total neutron absorbed dose i.e. 0.29 Gy per 1 Wh of reactor operation, is due to the fast neutron spectrum component. Only 0.05 Gy and 0.07 Gy in the total neutron absorbed dose are due to epithermal and thermal neutrons respectively. At the same time gamma absorption dose is 0.35 Gy. The development of experimental fuel channel GRK has wide possibility for utilization, from electronic components fast neutron studies, dosimeters testing, to cross section measurements for fast neutron energies. (author)

Replacement of the Pumps for Fuel Channel Cooling Circuit of the Maria Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Krzysztoszek, G.; Mieleszczenko, W.; Moldysz, A. [National Centre for Nuclear Research, Otwock–Świerk (Poland)

2014-08-15

The high flux Maria research reactor is operated by the National Centre for Nuclear Research in Świerk. It is a pool type reactor with pressurized fuel channels located in the beryllium matrix. According to the Global Threat Reduction Initiative programme our goal is to convert the Maria reactor from HEU to LEU fuel. Hydraulic losses in the new LEU fuel produced by CERCA are about 30% higher than the existing HEU fuel of type MR-6. For the MR-6 fuel were installed four two speed pumps. These pumps performed the function of the main circulations pumps during reactor operation with residual pumping power provided by emergency pumps. In the new system four main pumps will be used for circulating coolant while the reactor is operation with three auxiliary pumps for decay heat removal after reactor shutdown, meaning that the conversion of Maria research reactor will be possible after increasing flow in the primary cooling circuit of the fuel channels. The technical design of replacement of the pumps in the primary fuel channel cooling circuit was finished in April 2011 and accepted by the Safety Committee. After delivery of the new pumps we are planning to upgrade the primary fuel channel cooling circuit during October–November 2012. (author)

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.

Pressure data for various flow channels in proton exchange membrane (PEM) fuel cell

International Nuclear Information System (INIS)

Cho, Son Ah; Lee, Pil Hyong; Han, Sang Seok; Hwang, Sang Soon

2008-01-01

Micro flow channels in flow plates of fuel cells have become much narrower and longer to improve reactant flow distribution leading to increase of pumping power. Therefore it is very important to minimize the pressure drops in the flow channel because increased pumping power reduces overall efficiency. We investigated pressure drops in a micro flow channel at the anode and cathode compared to pressure losses for cold flow in straight, bended and serpentine channels. The results show that friction factors for cold flow channels could be used for parallel and bended flow channel designs for fuel cells. Pressure drop in the serpentine flow channel is the lowest among all flow channels due to bypass flow across the gas diffusion layer under reactive flow condition, although its pressure drop is highest for a cold flow condition. So the effect of bypass flow for serpentine flow channels should be considered when designing flow channels

Flow field optimization for proton exchange membrane fuel cells with varying channel heights and widths

International Nuclear Information System (INIS)

Wang Xiaodong; Huang Yuxian; Cheng, C.-H.; Jang, J.-Y.; Lee, D.-J.; Yan, W.-M.; Su Ay

2009-01-01

The optimal cathode flow field design of a single serpentine proton exchange membrane fuel cell is obtained by adopting a combined optimization procedure including a simplified conjugate-gradient method (SCGM) and a completely three-dimensional, two-phase, non-isothermal fuel cell model. The cell output power density P cell is the objective function to be maximized with channel heights, H 1 -H 5 , and channel widths, W 2 -W 5 as search variables. The optimal design has tapered channels 1, 3 and 4, and diverging channels 2 and 5, producing 22.51% increment compared with the basic design with all heights and widths setting as 1 mm. Reduced channel heights of channels 2-4 significantly enhance sub-rib convection to effectively transport oxygen to and liquid water out of diffusion layer. The final diverging channel prevents significant leakage of fuel to outlet via sub-rib convection from channel 4. Near-optimal design without huge loss in cell performance but is easily manufactured is discussed.

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)

Simulation of transient heat transfer in MACSTOR/KN-400 module storing irradiated CANDU fuel

Energy Technology Data Exchange (ETDEWEB)

Sabourin, G. [Atomic Energy of Canada Limited, Montreal, Quebec (Canada); Lee, K.-H.; Yoon, J.-H.; Choi, B.-I.; Lee, H.-Y.; Song, M.-J. [KHNP, Nuclear Environment Technology Inst., Taejon (Korea, Republic of)

2004-07-01

Korea Hydro and Nuclear Power (KHNP), in collaboration with Atomic Energy of Canada Limited (AECL), are developing a new module for the dry storage of spent fuel from the four CANDU 6 nuclear reactors at the Wolsong site in South Korea, the MACSTOR/KN-400. The simulation of transient conditions for AECL's spent fuel dry storage systems, presented in this paper, has not been performed before and is considered a major achievement of the present work. In a fist step, CATHENA was compared to MACSTOR-200 temperature measurements and the accuracy of the results were very good. In a second step, CATHENA was applied to the MACSTOR/KN-400. Four cases were performed for the MACSTOR/KN-400: Off-normal cases in summer and winter and reduced air flow cases in summer and winter. The maximum local concrete temperatures were predicted to be 63{sup o}C for the off-normal case and 65{sup o}C in the reduced air flow case. The maximum temperature gradients in the concrete are predicted to be 28{sup o}C for the off-normal case and 30{sup o}C in the reduced air flow case, incorporating a 3{sup o}C uncertainty. This paper shows that the maximum temperature for the module is expected to meet the temperature limitations of appropriate standards. (author)

Flow channel shape optimum design for hydroformed metal bipolar plate in PEM fuel cell

Energy Technology Data Exchange (ETDEWEB)

Peng, Linfa; Lai, Xinmin; Liu, Dong' an; Hu, Peng [State Key Laboratory of Mechanical System and Vibration, Shanghai Jiao Tong University, Shanghai 200240 (China); Ni, Jun [Department of Mechanical Engineering and Applied Mechanics, University of Michigan, Ann Arbor, MI 48109 (United States)

2008-03-15

Bipolar plate is one of the most important and costliest components of polymer electrolyte membrane (PEM) fuel cells. Micro-hydroforming is a promising process to reduce the manufacturing cost of PEM fuel cell bipolar plates made of metal sheets. As for hydroformed bipolar plates, the main defect is the rupture because of the thinning of metal sheet during the forming process. The flow channel section decides whether high quality hydroformed bipolar plates can be successively achieved or not. Meanwhile, it is also the key factor that is related with the reaction efficiency of the fuel cell stacks. In order to obtain the optimum flow channel section design prior the experimental campaign, some key geometric dimensions (channel depth, channel width, rib width and transition radius) of flow channel section, which are related with both reaction efficiency and formability, are extracted and parameterized as the design variables. By design of experiments (DOE) methods and an adoptive simulated annealing (ASA) optimization method, an optimization model of flow channel section design for hydroformed metal bipolar plate is proposed. Optimization results show that the optimum dimension values for channel depth, channel width, rib width and transition radius are 0.5, 1.0, 1. 6 and 0.5 mm, respectively with the highest reaction efficiency (79%) and the acceptable formability (1.0). Consequently, their use would lead to improved fuel cell efficiency for low cost hydroformed metal bipolar plates. (author)

Upgrade of Dhruva fuel channel flow instrumentation

International Nuclear Information System (INIS)

Gadgil, Kaustubh; Awale, P.K.; Sengupta, C.; Sumanth, P.; Roy, Kallol

2014-01-01

Dhruva, a 100 MW Heavy Water moderated and cooled, vertical tank-type Research Reactor, using metallic natural Uranium fuel has flow instrumentation for all the 144 fuel channels, consisting of venturi and triplicate DP gauges for each fuel channel. These gauges provide contacts for generation of reactor trip on low flow through fuel channel. These DP gauges were facing numerous generic and ageing related failures over the years and was also difficult to maintain owing to obsolescence. While considering an upgrade for these DP gauges, it was also planned to replace the existing Coolant Low Flow Trip (CLFT) system with a computer based Reactor Trip Logic System (RTLS). Being a retrofit job, the existing panels for mounting the gauges, cable layout, impulse tubing layout, etc. were retained, thereby simplifying the site execution, reducing reactor down time and also reducing person-milli-Sievert consumption. A customized Electronic DP Indicating Switch (EDPIS) was conceptualized for achieving these objectives. Such a design, utilizing a standard DP transmitter with customized electronic circuitry, was developed, evaluated and finalized after a series of factory trials, field trials and prototyping. The instrument design included contact input for existing CLFT system and also provision for 4-20 mA current output for the proposed computer based RTLS. The display and form factor of the instrument remained identical to older one and ensures familiarity of O and M personnel. Since EDPIS is classified as Safety Class IA, stringent type tests, hardware FMEA and V and V of the micro-controller software were carried out as per the requirements laid down by relevant standards for qualification of these instruments. Being a customized instrument, the manufacturing process was closely monitored and was followed by stringent QA plan and acceptance tests. A total of 396 gauges were replaced in a phased manner during scheduled fuelling outages and thereby did not affect reactor

Analysis of molten fuel behavior in coolant channel during severe accidents in KALIMER

International Nuclear Information System (INIS)

Suk, Soo Dong; Lee, Yong Bum; Hahn, Do Hee

2004-11-01

Preliminary safety analyses of the KALIMER-600 design have shown that the design has inherent safety characteristics and is capable of accommodating double fault initiators such as ATWS events without boiling coolant or melting fuel. For the future design of liquid metal reactor, however, the evaluation of the safety performance and the determination of containment requirements may require consideration of tripe-fault accident sequences of extremely low probability of occurrence that leads to fuel melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will required as a design requirement for the future design of LMR. For sodium-cooled core designs with metallic fuel, one of the major phenomenological modeling uncertainties to be resolved is the potential for freezing and plugging of molten metallic fuel in above- and below-core structures and possibly in inter-subassembly spaces. In this study, scoping analyses were carried out to evaluate the penetration depths in the coolant channels by molten fuel mixture during the unprotected loss-of-flow accidents in the core of the KALIMER-600. It is assumed in the analyses that a solid fuel crust would start to form upon contact with the coolant channel structure temperature of which is below the fuel solidus. The analysis results predict that the coolant channels would be plugged by the freezing molten fuel in the inlet lower shield as well as in the outlet, fission-gas-plenum region for the KALIMER-600 design

On the channel box for the fuel bundle of BWR

International Nuclear Information System (INIS)

Yokoyama, Hiroomi; Yamamoto, Takeo

1976-01-01

Channel boxes play the important roles of making coolant flow uniform and protecting fuel rods as the component of fuel assemblies for BWRs. About ten years ago, the domestic production of channel boxes was first investigated, and now, the original technology has been developed, and the channel boxes sufficiently satisfying the required quality can be produced. The actual experience by being charged in reactors has also been accumulated. At present, the supply capacity is almost sufficient to meet the domestic demand, and the future increase of demand can be dealt with promptly. The channel boxes are made of Zircaloy-4 plates which are favorable in view of neutron absorption, and are the boxes with 138 mm hollow square section, 2 mm thickness, and 4240 mm length. Two channels were welded together and made into a box. In order to eliminate the residual stress caused during the manufacture, high temperature heating with an electric furnace was adopted. The measurement of dimensions and the inspection of appearance of the channel boxes after irradiation proved that they were rather superior to imported ones. The production processes, the system for the quality guarantee, and the quality control in the Kobe Steel Ltd. are explained. The test and inspection are carried out at the time of accepting outside products, before starting the production, after the completion of longitudinal welding and after the completion of production. (Kako, I.)

Fuel cell plates with skewed process channels for uniform distribution of stack compression load

Science.gov (United States)

Granata, Jr., Samuel J.; Woodle, Boyd M.

1989-01-01

An electrochemical fuel cell includes an anode electrode, a cathode electrode, an electrolyte matrix sandwiched between electrodes, and a pair of plates above and below the electrodes. The plate above the electrodes has a lower surface with a first group of process gas flow channels formed thereon and the plate below the electrodes has an upper surface with a second group of process gas flow channels formed thereon. The channels of each group extend generally parallel to one another. The improvement comprises the process gas flow channels on the lower surface of the plate above the anode electrode and the process gas flow channels on the upper surface of the plate below the cathode electrode being skewed in opposite directions such that contact areas of the surfaces of the plates through the electrodes are formed in crisscross arrangements. Also, the plates have at least one groove in areas of the surfaces thereof where the channels are absent for holding process gas and increasing electrochemical activity of the fuel cell. The groove in each plate surface intersects with the process channels therein. Also, the opposite surfaces of a bipolar plate for a fuel cell contain first and second arrangements of process gas flow channels in the respective surfaces which are skewed the same amount in opposite directions relative to the longitudinal centerline of the plate.

Sub-channel analysis of a HPLWR fuel assembly with STAR-CD

International Nuclear Information System (INIS)

Himmel, Steffen R.; Class, Andreas G.; Schulenberg, Thomas; Laurien, Eckart

2008-01-01

Hofmeister et. al. developed a first design proposal for a HPLWR fuel assembly, consisting of a square 7 by 7 fuel pin arrangement within an assembly box and a water box in the centre, replacing 9 fuel rods. Instead of conventional grid spacers, wire wraps are considered due to good coolant mixing and low pressure drop in either flow direction. Within the present work, a novel approach describing the coolant heat up in the sub-channels of such an assembly has been investigated: the commercial software package STAR-CD has been used as a sub-channel code to investigate the thermal-hydraulic performance of such an HPLWR fuel assembly. The aim of the work is to demonstrate that a widely accepted commercial Computational Fluid Dynamics (CFD) code can be used for full rod bundle analysis by applying minor modifications to it. In steady of writing a dedicated code system with numerical solver routines and post-processing tools for sub-channel analyses, the user benefits from the optimized Graphical User Interface (GUI) already provided in STAR-CD. Moreover, a smooth transition to full three-dimensional modeling of the fluid flow inside rod bundles will be possible with the same code system, if considered to be necessary, just by refining the spatial discretization. Steady-state and transient flow regimes can be studied for design as well as reactor safety analysis. As the STAR-CD code uses the Finite Volume Method (FVM) for spatial discretization, the conservation equations for mass, momentum and energy were modified via user-subroutines to obtain the equations known from the usual sub-channel approach. The method will be explained in detail and results will be discussed. (author)

Channel geometric scales effect on performance and optimization for serpentine proton exchange membrane fuel cell (PEMFC)

Science.gov (United States)

Youcef, Kerkoub; Ahmed, Benzaoui; Ziari, Yasmina; Fadila, Haddad

2017-02-01

A three dimensional computational fluid dynamics model is proposed in this paper to investigate the effect of flow field design and dimensions of bipolar plates on performance of serpentine proton exchange membrane fuel cell (PEMFC). A complete fuel cell of 25 cm2 with 25 channels have been used. The aim of the work is to investigate the effect of flow channels and ribs scales on overall performance of PEM fuel cell. Therefore, geometric aspect ratio parameter defined as (width of flow channel/width of rib) is used. Influences of the ribs and openings current collector scales have been studied and analyzed in order to find the optimum ratio between them to enhance the production of courant density of PEM fuel cell. Six kind of serpentine designs have been used in this paper included different aspect ratio varying from 0.25 to 2.33 while the active surface area and number of channels are keeping constant. Aspect ratio 0.25 corresponding of (0.4 mm channel width/ 1.6mm ribs width), and Aspect ratio2.33 corresponding of (0.6 mm channel width/ 1.4mm ribs width. The results show that the best flow field designs (giving the maximum density of current) are which there dimensions of channels width is minimal and ribs width is maximal (Γ≈0.25). Also decreasing width of channels enhance the pressure drop inside the PEM fuel cell, this causes an increase of gazes velocity and enhance convection process, therefore more power generation.

Control of helium activity in the fuel reactor channels; Kontrola aktivnosti heliuma u tehnoloskim kanalima

Energy Technology Data Exchange (ETDEWEB)

Vidmar, M; Milosevic, M; Hadzic, S [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Yugoslavia)

1961-02-15

The objective of this task was to study the possibility of detecting a damaged fuel channel, and to introduce automated procedure for continuous control of reactor channels during operation. The existing control systems at the RA reactor (permanent control of heavy water and helium activity, radiation monitoring of heavy water and helium system, measurements of fire damp gas percent) are not sufficient for fast detection of fuel element failures. Since a 'hot' fuel channel cannot be removed from the core because it should be cooled in the core by heavy water circulation, it is not possible to prevent contamination of heavy water by fission products. It is concluded that it is not indispensable to detect the failed fuel element promptly, i.e. that tome is not a critical issue.

The Analysis of the Effect of Coolant Channel Width on Fuel Loading of the RSG-GAS Core

International Nuclear Information System (INIS)

Surbakti; Tukiran

2004-01-01

The RGS-GAS using uranium silicide fuel, plate type and 250 g U of loading is planned to increase the fuel loading to 300 g U even to 400 g U. The silicide fuel has advantages when increase the fuel loading in the same volume. Because of that case, it is necessary to analyze the effect of coolant channel width on fuel loading of the RSG-GAS core. Analyzing the effect the work which done is to generate cell and core calculation using WIMSD/4 and Batan-2DIFF codes. The WIMSD/4 code is used to generate cross section of core material and Batan-2DIFF is used to calculate the effective multiplication factor. The model that used in this calculation there are three kind of fuel loading namely, 250 g U, 300 g U and 400 g U. The coolant channel width is simulated from 1.75 mm to 2.55 mm. From that fuel loadings, it is analyzed which coolant channel width gave the best effective multiplication factor. From result of analysis showed that the best effective multiplication factor is on the coolant channel width of 2.55 mm for third of fuel loadings. (author)

New concepts, requirements and methods concerning the periodic inspection of the CANDU fuel channels

International Nuclear Information System (INIS)

Denis, J.R.

1995-01-01

Periodic inspection of fuel channels is essential for a proper assessment of the structural integrity of these vital components of the reactor. The development of wet channel technologies for non-destructive examination (NDE) of pressure tubes and the high technical performance and reliability of the CIGAR equipment have led, in less than 1 0 years, to the accumulation of a very significant volume of data concerning the flaw mechanisms and structural behaviour of the CANDU fuel channels. On this basis, a new form of the CAN/CSA-N285.4 Standard for Periodic Inspection of CANDU Nuclear Power Plant components was elaborated, introducing new concepts and requirements, in accord with the powerful NDE methods now available. This paper presents these concepts and requirements, and discusses the NDE methods, presently used or under development, to satisfy these requirements. Specific features regarding the fuel channel inspections of Cernavoda NGS Unit 1 are also discussed. (author)

RBMK fuel channel integrity. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

International Nuclear Information System (INIS)

1999-01-01

The fuel channel integrity in the RBMK NPPs is an issue of high safety concern. To date, three single fuel channel ruptures have occurred. Fuel channel rupture results in release of radioactivity to the reactor cavity and may lead to a release of radioactivity to the environment if the confinement safety system does not function properly. A multiple fuel channel rupture exceeding the venting capacity of the reactor cavity overpressure protection system poses a major impact on the plant safety. Further, due to incorrect prediction at the design stage the gas gap between the fuel channel pressure tube and the graphite blocks closes after approximately 17 years of plant operation. There is no safety justification available for the continued plant operation in this condition and the reactors are being retubed to avoid operation in this out of design condition, which may have negative impact on the fuel channel integrity. The loss of the mechanical integrity of fuel channel pressure tubes is a major safety concern for RBMK reactors since it may lead to overpressurization of the reactor cavity and consequently develop into a severe accident. In this report, information on the main design features of the RBMK reactor related to the fuel channel integrity is given. Further, detailed information on the fuel channel pressure tube and the graphite blocks with respect to their design, manufacture, in-service inspection, operating experience, ageing behaviour including degradation mechanisms is discussed in detail. The behaviour of the system fuel channel-graphite core including the corrective actions developed and implemented is discussed. Both normal operating conditions and accident conditions are addressed, considering also the gas gap closure process and its impact. The report also covers the fuel channel ducts. It is concluded in the report that for RBMK-1000 reactors and the adopted retubing strategy, limited local gas gap closure occurs at the time of pressure tube

Influences of bipolar plate channel blockages on PEM fuel cell performances

International Nuclear Information System (INIS)

Heidary, Hadi; Kermani, Mohammad J.; Dabir, Bahram

2016-01-01

Highlights: • Effect of partial- or full-blockage of PEMFC flow channels is numerically studied. • The anode blockage does not show any positive effects on cell performance. • Full blockages, despite higher pressure drop, better enhance net electrical power. • Additions of blocks more than five do not improve the cell performance. • Full blockage of cathode channels with five blocks enhances the net power by 30%. - Abstract: In this paper, the effect of partial- or full-block placement along the flow channels of PEM fuel cells is numerically studied. Blockage in the channel of flow-field diverts the flow into the gas diffusion layer (GDL) and enhances the mass transport from the channel core part to the catalyst layer, which in turn improves the cell performance. By partial blockage, only a part of the channel flow is shut off. While in full blockage, in which the flow channel cross sections are fully blocked, the only avenue left for the continuation of the gas is to travel over the blocks via the porous zone (GDL). In this study, a 3D numerical model consisting of a 9-layer PEM fuel cell is performed. A wide spectrum of numerical studies is performed to study the influences of the number of blocks, blocks height, and anode/cathode-side flow channel blockage. The results show that the case of full blockage enhances the net electrical power more than that of the partial blockage, in spite of higher pressure drop. Performed studies show that full blockage of the cathode-side flow channels with five blocks along the 5 cm channel enhances the net power by 30%. The present work provides helpful guidelines to bipolar plate manufacturers.

High-temperature thermal-chemical analysis of nuclear fuel channels

Energy Technology Data Exchange (ETDEWEB)

Nekhamkin, Y; Rosenband, V; Hasan, D; Elias, E; Wacholder, E; Gany, A [Technion-Israel Inst. of Tech., Haifa (Israel)

1996-12-01

In a severe accident situation, e.g., a postulated loss of coolant accident with a coincident loss of emergency core cooling (LOCA/LOECC), the core may become partially uncovered and steam may become the only coolant available. The thermodynamic conditions in the core, in this case, depend on ability of the steam to effectively remove the fuel decay heat and the heat generated by the exothermic steam/Zircaloy reaction., Therefore, it is important to understand the high-temperature behavior of an oxidizing fuel channel. The main objective of this work is to develop a methodology for calculating the clad temperature and rate of oxidation of a partially covered fuel pin. A criterion is derived to define the importance of the chemical reaction in the overall heat balance. The main parameters affecting the fuel thermal behavior are outlined (authors).

Fabrication of metallic channel-containing UO2 fuels

International Nuclear Information System (INIS)

Yang, Jae Ho; Song, Kun Woo; Kim, Keon Sik; Jung, Youn Ho

2004-01-01

The uranium dioxide is widely used as a fuel material in the nuclear industry, owing to many advantages. But it has a disadvantage of having the lowest thermal conductivity of all kinds of nuclear fuels; metal, carbide, nitride. It is well known that the thermal conductivity of UO 2 fuel is enhanced by making, so called, the CERMET (ceramic-metal) composite which consists of both continuous body of highly thermal-conducting metal and UO 2 islands. The CERMET fuel fabrication technique needs metal phase of at least 30%, mostly more than 50%, of the volume of the pellet in order to keep the metal phase interconnected. This high volume fraction of metal requires such a high enrichment of U that the parasitic effect of metal should be compensated. Therefore, it is attractive to develop an innovative composite fuel that can form continuous metal phase with a small amount of metal. In this investigation, a feasibility study was made on how to make such an innovative fuel. Candidate metals (W, Mo, Cr) were selected, and fabrication process was conceptually designed from thermodynamic calculations. We have experimentally found that a metal phase envelops perfectly UO 2 grains, forming continuous channel throughout the pellet, and improving the thermal conductivity of pellet

A Sub-channel Analysis of a VHTR Fuel Block with Tin Gap-Filler

International Nuclear Information System (INIS)

Cho, Chung Ho; Kim, Yong Hee; Yi, Yong Sun; Kim, Hong Pyo

2005-01-01

In the Nuclear Hydrogen Development and Demonstration (NHDD) project, two types of VHTRs (Very High Temperature Reactors), prismatic or pebble bed, are under investigation as the nuclear heat source for hydrogen production. In general, the targeted coolant outlet temperature of VHTR is 950∼1000 .deg. C and the maximum allowable fuel temperature is 1250 .deg. C during the normal operation. In the case of the prismatic reactor (PMR), conventional fuel designs result in a small margin in the maximum fuel temperature. This is one of the biggest demerits of the prismatic type In this paper, a technique of lowering the maximum fuel temperature is suggested. The PMR fuel assembly is comprised of many coolant holes and fuel channels. Cylindrical fuel compacts are stacked inside the fuel channel. Consequently, there should be a physical gap between the fuel compact and graphite block, which is filled with the He gas in the conventional design. The heat transfer coefficient of the He gap is very poor, and this increases the fuel temperature substantially. In the proposed design measure, the gap is filled with a liquid metal, tin (Sn) that has a very high thermal conductivity. The effects of tin in the gap with gap distance variation in the viewpoint of thermal hydraulics are quantitatively discussed. Also, the effects of the variations of the axial power distribution are discussed

Modeling of fuel bundle vibration and the associated fretting wear in a CANDU fuel channel

International Nuclear Information System (INIS)

Mohany, A.; Hassan, M.

2011-01-01

In this paper a numerical model is developed to predict the vibration response of a CANDU® fuel bundle and the associated fretting wear in the surrounding pressure tube. One excitation mechanism is considered in this model; turbulence-induced excitation caused by coolant flow inside the fuel channel. The numerical model can be easily adapted to include the effects of seismic events, fuel bundle impact during refuelling and start-up of the reactor, and the acoustic pressure pulsations caused by the primary heat transport (PHT) pumps. The simulation is performed for a typical CANDU fuel bundle with 37 fuel elements. The clearances between the buttons of the inner fuel elements, and between the bearing pads of the outer fuel elements and the pressure tube were measured from an actual fuel bundle. Some variability among the measured clearance values was observed. Therefore, probability density functions of the measured clearance values were established and the simulation was performed for the probabilistic distribution of the clearance values. The contact between the fuel bundle and the pressure tube is modeled using pseudo-force contact method. The proposed modelling technique can be used in future CANDU reactors to avoid fuel and pressure tube fretting damage due to the aforementioned excitation mechanisms. (author)

An experimental investigation of heat transfer from a reactor fuel channel to surrounding water

International Nuclear Information System (INIS)

Gillespie, G.E.

An important feature of the CANDU-PHW reactor is that each fuel channel is surrounded by cool heavy-water moderator that can act as a sink for heat generated in the fuel if other means of heat removal were to fail. During postulated loss-of-coolant accidents there are two scenarios in which the primary cooling system may not prevent fuel-channel overheating. These situations arise when: (1) for a particular break size and location, called the critical break, the coolant flow through a portion of the reactor core stagnates before the emergency coolant injection system restores circulation, or, (2) the emergency coolant injection system fails to operate. In either case, the heat generated in the fuel is transferred mainly by radiation to the pressure tube and calandria tube, and then by boiling heat transfer to the moderator. This paper describes a simple one-dimensional model developed to analyse the thermal behaviour of a fuel channel when the internal pressure is high. Also described is a series of experiments in which the pressure-tube segment is pressurized and heated at a constant rate until it contacts a surrounding calandria-tube segment. Predictions of the one-dimensional model are compared with the experimental results

Gothic simulation of single-channel fuel heatup following a loss of forced flow

Energy Technology Data Exchange (ETDEWEB)

Chen, X-Q; Tahir, A. [NSS, Dept. of Thermal Hydraulics Analysis, Toronto, Ontario (Canada); Parlatan, Y. [Ontario Power Generation, NSATD, Pickering, Ontario (Canada); Kwee, M. [Bruce Power, NSASD, Toronto, Ontario (Canada)

2011-07-01

GOTHIC v7.2 was used to develop a computer model for the simulation of 28- and 37-element fuel heat-up at a loss of forced flow. The model has accounted for the non-uniformity of both axial and radial power distributions along the fuel channel for a typical CANDU reactor. In addition, the model has also accounted for the fuel rods, end-fittings, feeders and headers. Experimental test conditions for both 28- and 37-element bundles at either low or high powers were used for model validation. GOTHIC predictions of the rod and/or pressure-tube temperatures at a variety of test locations were compared with the corresponding experimental measurements. It is found that the numerical results agree well with the experimental measurements for most of the test locations. Results have also shown that the channel venting time is sensitive to the initial temperature distribution in the feeders and headers. An imposed temperature asymmetry at the beginning will cause the channel flow to vent earlier. (author)

Liquid Water Transport in the Reactant Channels of Proton Exchange Membrane Fuel Cells

Science.gov (United States)

Banerjee, Rupak

Water management has been identified as a critical issue in the development of PEM fuel cells for automotive applications. Water is present inside the PEM fuel cell in three phases, i.e. liquid phase, vapor phase and mist phase. Liquid water in the reactant channels causes flooding of the cell and blocks the transport of reactants to the reaction sites at the catalyst layer. Understanding the behavior of liquid water in the reactant channels would allow us to devise improved strategies for removing liquid water from the reactant channels. In situ fuel cell tests have been performed to identify and diagnose operating conditions which result in the flooding of the fuel cell. A relationship has been identified between the liquid water present in the reactant channels and the cell performance. A novel diagnostic technique has been established which utilizes the pressure drop multiplier in the reactant channels to predict the flooding of the cell or the drying-out of the membrane. An ex-situ study has been undertaken to quantify the liquid water present in the reactant channels. A new parameter, the Area Coverage Ratio (ACR), has been defined to identify the interfacial area of the reactant channel which is blocked for reactant transport by the presence of liquid water. A parametric study has been conducted to study the effect of changing temperature and the inlet relative humidity on the ACR. The ACR decreases with increase in current density as the gas flow rates increase, removing water more efficiently. With increase in temperature, the ACR decreases rapidly, such that by 60°C, there is no significant ACR to be reported. Inlet relative humidity of the gases does change the saturation of the gases in the channel, but did not show any significant effect on the ACR. Automotive powertrains, which is the target for this work, are continuously faced with transient changes. Water management under transient operating conditions is significantly more challenging and has not

A three-dimensional model of PEM fuel cells with serpentine flow channels

International Nuclear Information System (INIS)

Nguyen, P.T.; Berning, T.; Bang, M.; Djilali, N.

2003-01-01

A three-dimensional computational model of PEM fuel cell with serpentine flow field channels is presented in this paper. This model presents a comprehensive account for all important transport phenomena in fuel cell such as heat transfer, mass transfer, electrode kinetics, and potential fields in the membrane and gas diffusion layers. A new approach of solving for the potential losses across the cell was also developed in this model. The dependency of local current density on oxygen concentration and activation overpotential is fully addressed in this model. The computational domain consists of serpentine gas flow channels, porous gas diffusion layers, catalyst layers, and a membrane. Results obtained from this model are in good agreement with experimental results. (author)

Thermosyphoning analysis with the CATHENA model of the blanket and first wall cooling loop for the SEAFP reactor design

International Nuclear Information System (INIS)

Ross, W.E.

1994-02-01

This report documents the thermosyphoning analysis which was performed with the CATHENA network model of one of the blanket and first wall cooling loops of the SEAFP reactor design. This thermosyphoning analysis includes four simulations, each with a slightly different model feature or assumption. These simulations are performed to assess the primary heat transport system behaviour for a complete loss of electrical power event (total loss of flow) and to estimate the rate and extent of heat-up of the incore components. For each event, a description of some of the important aspects of the transient thermalhydraulic behaviour including coolant temperatures, circuit and sector flows, circuit pressure, pressurizer level and outflow, and first wall and blanket temperatures is provided. (author). 4 refs., 2 tabs., 32 figs

Eddy current and ultrasonic fuel channel inspection at Karachi Nuclear Power Plant

International Nuclear Information System (INIS)

Mayo, W.R.; Alam, M.M.

1997-01-01

In November of 1993 and in-service inspection was performed on eight fuel channels in the Karachi Nuclear Power Plant (KANUPP) reactor. The workscope included ultrasonic and eddy current volumetric examinations, and eddy current measurement of pressure-to calandria tube gap. This paper briefly discusses the planning strategy of the ultrasonic and eddy current examinations, and describes the equipment developed to meet the requirements, followed by details of the actual channel inspection campaign. The presented nondestructive examinations assisted in determining fitness for service of KANUPP reactor channels in general, and confirmed that the problems associated with channel G12 were not generic in nature. (author)

Modeling and simulation of PEM fuel cell's flow channels using CFD techniques

International Nuclear Information System (INIS)

Cunha, Edgar F.; Andrade, Alexandre B.; Robalinho, Eric; Bejarano, Martha L.M.; Linardi, Marcelo; Cekinski, Efraim

2007-01-01

Fuel cells are one of the most important devices to obtain electrical energy from hydrogen. The Proton Exchange Membrane Fuel Cell (PEMFC) consists of two important parts: the Membrane Electrode Assembly (MEA), where the reactions occur, and the flow field plates. The plates have many functions in a fuel cell: distribute reactant gases (hydrogen and air or oxygen), conduct electrical current, remove heat and water from the electrodes and make the cell robust. The cost of the bipolar plates corresponds up to 45% of the total stack costs. The Computational Fluid Dynamic (CFD) is a very useful tool to simulate hydrogen and oxygen gases flow channels, to reduce the costs of bipolar plates production and to optimize mass transport. Two types of flow channels were studied. The first type was a commercial plate by ELECTROCELL and the other was entirely projected at Programa de Celula a Combustivel (IPEN/CNEN-SP) and the experimental data were compared with modelling results. Optimum values for each set of variables were obtained and the models verification was carried out in order to show the feasibility of this technique to improve fuel cell efficiency. (author)

A modal method for transient thermal analysis of CANDU fuel channel

International Nuclear Information System (INIS)

Park, J-W.; Muzumdar, A.J.

1996-01-01

The classical modal expansion technique has been applied to predict transient fuel and coolant temperatures under on-power conditions in a CANDU fuel channel. The temperature profile across the fuel pellet is assumed to be parabolic and fuel and coolant temperatures are expanded with Fourier series. The coefficient derivatives are written in state space form and solved by the Runge-Kutta method of fifth order. To validate the present model, the calculated fuel temperatures for several sample cases were compared with HOTSPOT-II, which employs a more rigorous finite-difference model. The agreement was found to be reasonable for the operational transients simulated. The advantage of the modal method is the fast computation speed for application to the real-time system such as the CANDU simulator which is being currently developed at the Institute for Advanced Engineering (IAE). (author)

RBMK fuel channel blockage analysis by MCNP5, DRAGON and RELAP5-3D codes

International Nuclear Information System (INIS)

Parisi, C.; D'Auria, F.

2007-01-01

The aim of this work was to perform precise criticality analyses by Monte-Carlo code MCNP5 for a Fuel Channel (FC) flow blockage accident, considering as calculation domain a single FC and a 3x3 lattice of RBMK cells. Boundary conditions for MCNP5 input were derived by a previous transient calculation by state-of-the-art codes HELIOS/RELAP5-3D. In a preliminary phase, suitable MCNP5 models of a single cell and of a small lattice of RBMK cells were set-up; criticality analyses were performed at reference conditions for 2.0% and 2.4% enriched fuel. These analyses were compared with results obtained by University of Pisa (UNIPI) using deterministic transport code DRAGON and with results obtained by NIKIET Institute using MCNP4C. Then, the changes of the main physical parameters (e.g. fuel and water/steam temperature, water density, graphite temperature) at different time intervals of the FC blockage transient were evaluated by a RELAP5-3D calculation. This information was used to set up further MCNP5 inputs. Criticality analyses were performed for different systems (single channel and lattice) at those transient' states, obtaining global criticality versus transient time. Finally the weight of each parameter's change (fuel overheating and channel voiding) on global criticality was assessed. The results showed that reactivity of a blocked FC is always negative; nevertheless, when considering the effect of neighboring channels, the global reactivity trend reverts, becoming slightly positive or not changing at all, depending in inverse relation to the fuel enrichment. (author)

Performance Characteristics of a PEM Fuel Cell with Parallel Flow Channels at Different Cathode Relative Humidity Levels

Directory of Open Access Journals (Sweden)

Sang Soon Hwang

2009-11-01

Full Text Available In fuel cells flow configuration and operating conditions such as cell temperature, humidity at each electrode and stoichiometric number are very crucial for improving performance. Too many flow channels could enhance the performance but result in high parasite loss. Therefore a trade-off between pressure drop and efficiency of a fuel cell should be considered for optimum design. This work focused on numerical simulation of the effects of operating conditions, especially cathode humidity, with simple micro parallel flow channels. It is known that the humidity at the cathode flow channel becomes very important for enhancing the ion conductivity of polymer membrane because fully humidified condition was normally set at anode. To investigate the effect of humidity on the performance of a fuel cell, in this study humidification was set to 100% at the anode flow channel and was changed by 0–100% at the cathode flow channel. Results showed that the maximum power density could be obtained under 60% humidified condition at the cathode where oxygen concentration was moderately high while maintaining high ion conductivity at a membrane.

In-situ optical profilometry of CANDU fuel channels

Energy Technology Data Exchange (ETDEWEB)

Jarvis, G.N.; Cornblum, E.O.; Grabish, M.G. [Ontario Hydro Nuclear, Nuclear Technology Services, Pickering, ON (Canada); Mader, D.L.; Kuurstra, J.C.; McNabb, S.C. [Ontario Hydro Technologies, Toronto, ON (Canada)

1996-12-31

This paper describes the application of optical profilometry techniques, to measure directly the depth and root-radius of open inside-surface flaws, within a flooded reactor pressure tube. The use of optical profilometry in one form or another is not new. Systems typically make use of scanned laser beams to perform tasks such as weld bead profiling and contour mapping of various solid objects. The specific application of such techniques within the harsh, confined environmental conditions described herein is thought to be somewhat unique and has provided some technical challenges. A brief outline of the fuel channel geometry will be given together with the basics of the profilometry technique employed. This will be followed by a detailed description of the in-channel tooling, key features of the data collection and analysis software and operational experience. (UK).

In-situ optical prolifometry of CANDU fuel channels

Energy Technology Data Exchange (ETDEWEB)

Jarvis, G.N.; Cornblum, E.O.; Grabish, M.G. [Ontario Hydro, Pickering, ON (Canada). Pickering Generating Station; Mader, D.L.; Kuurstra, J.C.; McNabb, S.C. [Ontario Hydro, Toronto, ON (Canada)

1996-12-31

This paper will describe the application of optical profilometry techniques, to measure directly the depth and root-radius of open inside-surface flaws, within a flooded reactor pressure tube. The use of optical profilometry in one form or another is not new. Systems typically make use of scanned laser beams to perform tasks such as weld bead profiling and contour mapping of various solid objects. The specific application of such techniques within the harsh, confined environmental conditions described herein is thought to be somewhat unique and has provided some technical challenges. A brief outline of the fuel channel geometry will be given together with the basics of the profilometry technique employed. This will be followed by a detailed description of the in-channel tooling, key features of the data collection and analysis software and operational experience. (Author).

THERMOSS: a thermohydraulic model of flow stagnation in a horizontal fuel channel

International Nuclear Information System (INIS)

Gulshani, P.; Caplan, M.Z.; Spinks, N.J.

1984-01-01

Following a postulated inlet-side small break in the CANDU reactor, emergency coolant is injected to refull the horizontal fuel channels and remove the decay heat. As part of the accident analysis, the effects of loss of forced circulation during the accident are predicted. A break size exists for which, at the end of pump rundown, the break force balances the natural circulation force and the channel flow is reduced to near zero. The subcooled, stagnant channel condition is referred to as the standing-start condition. Subsequently, the channel coolant boils and stratifies. Eventually the steam flow from the channel heats up the endfitting to the saturation temperature and reaches the vertical feeder. The resulting buoyancy-induced flow then refills the channel. One dimensional, two-fluid conservation equations are solved in closed form to predict the duration of stagnation. In this calculation the channel water level is an important intermediate variable because it determines the amount of steam production

CFD modelling of cooling channel geometry of PEM fuel cell for ...

African Journals Online (AJOL)

In this study, a numerical investigation was carried out to deter mine the impact of cooling channel geometry in combination with temperature dependent operating parameters on thermal management and overall performance of a PEM fuel cell system. The evaluation is performed using a computational fluid dynamics ...

Effect of flow field with converging and diverging channels on proton exchange membrane fuel cell performance

International Nuclear Information System (INIS)

Zehtabiyan-Rezaie, Navid; Arefian, Amir; Kermani, Mohammad J.; Noughabi, Amir Karimi; Abdollahzadeh, M.

2017-01-01

Highlights: • Effect of converging and diverging channels on fuel cell performance. • Over rib flow is observed from converging channels to neighbors. • Proposed flow field enriches oxygen level and current density in catalyst layer. • Net output power is enhanced more than 16% in new flow field. - Abstract: In this study, a novel bipolar flow field design is proposed. This new design consists of placed sequentially converging and diverging channels. Numerical simulation of cathode side is used to investigate the effects of converging and diverging channels on the performance of proton exchange membrane fuel cells. Two models of constant and variable sink/source terms were implemented to consider species consumption and production. The distribution of oxygen mole fraction in gas diffusion and catalyst layers as a result of transverse over rib velocity is monitored. The results indicate that the converging channels feed two diverging neighbors. This phenomenon is a result of the over rib velocity which is caused by the pressure difference between the neighboring channels. The polarization curves show that by applying an angle of 0.3° to the channels, the net electrical output power increases by 16% compared to the base case.

Development of a flow restrictor for CANDU fuel channels

International Nuclear Information System (INIS)

Schroeter, F.; Antonaccio, C.; Masciotra, H.; Klink, A.

2013-01-01

Due to the creep and neutron growth phenomena experienced by the components inside the reactor during operation of CNE it is expected that both the fuel channels and the Liquid Injection lances increase their permanent deformation. One of the deformation types that these two components experiment is the SAG, which is what happens with any beam supported on their extremes to which a load is applied, except for this case that due to creep and neutronic effect growth, part of the deformation is not elastic and increases with time To solve or avoid this condition, two solutions exist, one is to replace the pressure tubes, forcing the calandria tube to recover to a near to original position or to design a device that permits defueling of the channel without modifying the pressure drop and in this way not to affect the distribution of coolant in the core. In some channels it was decided to replace the pressure tube and in others it was decided to defuel them proposing a design for a flow restrictor. (author

TRAN.1 - a code for transient analysis of temperature distribution in a nuclear fuel channel

International Nuclear Information System (INIS)

Bukhari, K.M.

1990-09-01

A computer program has been written in FORTRAN that solves the time dependent energy conservation equations in a nuclear fuel channel. As output from the program we obtained the temperature distribution in the fuel, cladding and coolant as a function of space and time. The stability criteria have also been developed. A set of finite difference equations for the steady state temperature distribution have also been incorporated in this program. A number of simplifications have been made in this version of the program. Thus at present, TRAN.1 uses constant thermodynamics properties and heat transfer coefficient at fuel cladding gap, has absence of phase change and pressure loss in the coolant, and there is no change in properties due to changes in burnup etc. These effects are now in the process of being included in the program. The current version of program should therefore be taken as a fuel channel, and this report should be considered as a status report on this program. (orig./A.B.)

SHOSPA-MOD, Hot Spot Factors for Fuel and Clad, Hot Channel Factors

International Nuclear Information System (INIS)

Amendola, A.

1982-01-01

1 - Nature of the physical problem solved: SHOSPA evaluates the hot spot factors for fuel and cladding as well as the hot channel factor as a function of the confidence level. Moreover, it evaluates the probability on n hot subassemblies. The code has been developed with emphasis on sodium cooled fast reactors, but it is applicable to any type of reactors constituted of bundled fuel rods with single phase coolant. An option for plotting is available in this version. 2 - Restrictions on the complexity of the problem: This code is applicable to any type of reactors constituted of fuel rods with single phase coolant

Application of Sub-cooled Boiling Model to Thermal-hydraulic Analysis Inside a CANDU-6 Fuel Channel

International Nuclear Information System (INIS)

Kim, Man Woong; Lee, Sang Kyu; Kim, Hyun Koon; Yoo, Kun Joong; Kang, Hyoung Chul; Yoo, Seong Yeon

2007-01-01

Forced convection nucleate boiling is encountered in heat exchangers during normal and non-nominal modes of operation in pressurized water or boiling water reactors (PWRs or BWRs). If the wall temperature of the piping is higher than the saturation temperature of the nearby liquid, nucleate boiling occurs. In this regime, bubbles are formed at the wall. Their growth is promoted by the wall superheat (the difference between the wall and saturation temperatures), and they depart from the wall as a result of gravitational and liquid inertia forces. If the bulk liquid is subcooled, condensation at the bubble-liquid interface takes place and the bubble may collapse. This convection nucleate boiling is called as a sub-cooled nucleate boiling. As for the fuel channel of a CANDU 6 reactor, forced convection nucleate boiling models for flows along fuel elements enclosed inside typical CANDU-6 fuel channel has encountered difficulties due to the modeling of local effects along the horizontal channel. Therefore, the subcooled nucleate boiling has been modeled through temperature driven boiling heat and mass transfer, using a model developed at Rensselaer Polytechnic Institute. The objectives of this study are: (i) to investigate a proposed sub-cooled boiling model developed at Rensselaer Polytechnic Institute and (ii) to apply against a experiment and (iii) to predict local distributions of flow fields for the actual fuel channel geometries of CANDU-6 reactors. The numerical implementation is conducted using by the FLUENT 6.2 CFD computer code

Modeling and simulation of PEM fuel cell's flow channels using CFD techniques

Energy Technology Data Exchange (ETDEWEB)

Cunha, Edgar F.; Andrade, Alexandre B.; Robalinho, Eric; Bejarano, Martha L.M.; Linardi, Marcelo [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mails: efcunha@ipen.br; abodart@ipen.br; eric@ipen.br; mmora@ipen.br; mlinardi@ipen.br; Cekinski, Efraim [Instituto de Pesquisas Tecnologicas (IPT-SP), Sao Paulo, SP (Brazil)]. E-mail: cekinski@ipt.br

2007-07-01

Fuel cells are one of the most important devices to obtain electrical energy from hydrogen. The Proton Exchange Membrane Fuel Cell (PEMFC) consists of two important parts: the Membrane Electrode Assembly (MEA), where the reactions occur, and the flow field plates. The plates have many functions in a fuel cell: distribute reactant gases (hydrogen and air or oxygen), conduct electrical current, remove heat and water from the electrodes and make the cell robust. The cost of the bipolar plates corresponds up to 45% of the total stack costs. The Computational Fluid Dynamic (CFD) is a very useful tool to simulate hydrogen and oxygen gases flow channels, to reduce the costs of bipolar plates production and to optimize mass transport. Two types of flow channels were studied. The first type was a commercial plate by ELECTROCELL and the other was entirely projected at Programa de Celula a Combustivel (IPEN/CNEN-SP) and the experimental data were compared with modelling results. Optimum values for each set of variables were obtained and the models verification was carried out in order to show the feasibility of this technique to improve fuel cell efficiency. (author)

Analysis of fuel pin mechanics in case of flow blockage of a single RBMK channel

International Nuclear Information System (INIS)

Pierro, F.; Moretti, F.; Mazzini, D.; D'Auria, F.

2005-01-01

The evaluation of the consequences of the pressure tube rupture due to accidental overheating is one of the key elements for addressing an RBMK safety analysis, since it causes the lost of design boundaries against the fission products release. Several events are expected to take place: thermal hydraulic crisis (energy unbalance), fuel overheating, fuel rod damage, pressure tube overheating, pressure tube failure and graphite stack damage, Hydrogen and fission products release. The present work deals with the research activity carried out at ''Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione'' (DIMNP) of the University of Pisa aimed at assessing numerical models for safety analysis of the RBMK-1000. The attention is focused on the modelling of (1) a single fuel channel and its surrounding graphite column for evaluating the transient conditions enabling the different damaging phenomena, (2) a single fuel rod for investigating fuel pin behaviour, (3) the ruptured fuel channel for figuring the magnitude of the hydrodynamic loads acting on fuel rods. Different codes were employed to cover the competences for the investigation of each field; in particular, RELAP5 code for thermal-hydraulics, FRAPCON-3 and FRAPTRAN1-2 codes for fuel pin mechanics, FLUENT-6 for fluid dynamics. The paper discusses the numerical models, the analysis capabilities of numerical models in comparison with available data about the Leningrad NPP 1992 accident. Furthermore, the possibility to draw a failure map identifying the range of the cladding safety respect to the transient condition is outlined. (author)

Integrated aerosol and thermalhydraulics modelling for CANDU safety analysis

International Nuclear Information System (INIS)

McDonald, B.H.; Hanna, B.N.

1990-08-01

Analysis of postulated accidents in CANDU reactors that could result in severe fuel damage requires the ability to model the formation of aerosols containing fission product materials and the transport of these aerosols from the fuel, through containment, to any leak to the atmosphere. Best-estimate calculations require intimate coupling and simultaneous solution of all the equations describing the entire range of physical and chemical phenomena involved. The prototype CATHENA/PACE-3D has been developed for integrated calculation of thermalhydraulic and aerosol events in a CANDU reactor during postulated accidents. Examples demonstrate the ability of CATHENA/PACE-3D to produce realistic flow and circulation patterns and reasonable accuracy in solution of two simple fluid-flow test cases for which analytical solutions exist

Evaluation of fuel performance for fresh and aged CANDU reactor

Energy Technology Data Exchange (ETDEWEB)

Jung, Jong Yeob; Bae, Jun Ho; Park, Joo Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-10-15

Like all other industrial plants, nuclear power plants also undergo degradations, so called ageing, with their operation time. Accordingly, in the recent safety analysis for a refurbished Wolsong 1 NPP, various ageing effects were incorporated into the hydraulic models of a number of the components in the primary heat transport system for conservatism. The ageing data of thermal-hydraulic components for 11 EFPY of Wolsong 1 were derived by using NUCIRC code based on the site operation data and they were modified to the appropriate input data for CATHENA code which is a thermal hydraulic code for a postulated accident analysis. This paper deals with the ageing effect of the PHTS (primary heat transport system) of CANDU reactor on the fuel performance during the normal operation. Initial conditions for fuel performance analysis were derived from the thermal-hydraulic analysis for both fresh and aged core models. Here, fresh core means a core state just right after the refurbishment and the aged core is 11 EFPY state after the refurbishment of Wolsong 1. The fuel performance was analyzed by using ELESTRES code for both fresh and aged core state and the results were compared in order to verify the ageing effect of CANDU HTS on the fuel performance.

Application of Shuttle Remote Manipulator System technology to the replacement of fuel channels in the Pickering CANDU reactor

International Nuclear Information System (INIS)

Stratton, D.; Butt, C.

1982-04-01

Spar Aerospace Limited of Toronto was the prime contractor to the National Research Council of Canada for the design and development of the Shuttle Remote Manipulator (SRMS). Spar is presently under contract to Ontario Hydro to design and build a Remote Manipulation Control System to replace the fuel channels in the Pickering A Nuclear Generating Station. The equipment may be used to replace the fuel channels in six other early generation CANDU reactors

CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR. PART 6 - PRESENTATION OF THE DECOMMISSIONING DEVICE

Directory of Open Access Journals (Sweden)

Gabi ROSCA FARTAT

2015-05-01

Full Text Available The objective of this paper is to present a possible solution for the designing of a device for the decommissioning of the horizontal fuel channels in the CANDU 6 nuclear reactor. The decommissioning activities are dismantling, demolition, controlled removal of equipment, components, conventional or hazardous waste (radioactive, toxic in compliance with the international basic safety standards on radiation protection. One as the most important operation in the final phase of the nuclear reactor dismantling is the decommissioning of fuel channels. For the fuel channels decommissioning should be taken into account the detailed description of the fuel channel and its components, the installation documents history, adequate radiological criteria for decommissioning guidance, safety and environmental impact assessment, including radiological and non-radiological analysis of the risks that can occur for workers, public and environment, the description of the proposed program for decommissioning the fuel channel and its components, the description of the quality assurance program and of the monitoring program, the equipments and methods used to verify the compliance with the decommissioning criteria, the planning of performing the final radiological assessment at the end of the fuel channel decommissioning. These will include also, a description of the proposed radiation protection procedures to be used during decommissioning. The dismantling of the fuel channel is performed by one device which shall provide radiation protection during the stages of decommissioning, ensuring radiation protection of the workers. The device shall be designed according to the radiation protection procedures. The decommissioning device assembly of the fuel channel components is composed of the device itself and moving platform support for coupling of the selected channel to be dismantled. The fuel channel decommissioning device is an autonomous device designed for

Integral-fuel blocks

International Nuclear Information System (INIS)

Cunningham, C.; Simpkin, S.D.

1975-01-01

A prismatic moderator block is described which has fuel-containing channels and coolant channels disposed parallel to each other and to edge faces of the block. The coolant channels are arranged in rows on an equilateral triangular lattice pattern and the fuel-containing channels are disposed in a regular lattice pattern with one fuel-containing channel between and equidistant from each of the coolant channels in each group of three mutually adjacent coolant channels. The edge faces of the block are parallel to the rows of coolant channels and the channels nearest to each edge face are disposed in two rows parallel thereto, with one of the rows containing only coolant channels and the other row containing only fuel-containing channels. (Official Gazette)

THERMOSS: A thermohydraulic model of flow stagnation in a horizontal fuel channel

International Nuclear Information System (INIS)

Gulshani, P.; Caplan, M.Z.; Spinks, N.J.

1984-01-01

A model, called THERMOSS, is developed to compute the duration of stagnation in a CANDU reactor fuel channel with subcooled, stagnant initial conditions. The model solves, in closed form, the one dimensional, two-fluid conservation equations. In the computation of the duration of stagnation, the channel water level is an important intermediate variable because it determines the amount of steam production. A feature of the model is that water level is determined by a momentum balance between frictional pressure drop in the steam phase and hydrostatic head in the liquid phase. This is in contrast to an ealier model in which the level was determined from mass balance considerations. A satisfactory agreement between the predicted and experimentally observed channel water level and duration of stagnation is obtained. (orig.)

Operation of Atucha I nuclear power plant with 25 cooling channels without fuel elements

International Nuclear Information System (INIS)

Perez, R.A.; Sidelnik, J.I.; Salom, G.F.

1987-01-01

In view of the need of removing the irradiation probes from the reactor of Atucha I nuclear power plant, a study about the consequences of operating with 25 channels without their respective fuel elements was performed. This condition was simulated by means of the code PUMA symmetry I and the consequences were analyzed. From the study resulted a program of stepped power reduction of the nuclear plant that would take place during the process of channel emptying. (Author)

The problem of gas gap between graphite - fuel channel reduction impact at Ignalina NPP

International Nuclear Information System (INIS)

1999-01-01

Safety analysis of Ignalina NPP operation in the case when gap closure between graphite - fuel channel occur was performed. The main results of this analysis as well as data of gap measurements during the year 1996 - 1998 are provided

Extending Pickering Fuel Channel Life to 247,000 EFPH (2020)

International Nuclear Information System (INIS)

Malek, I.

2013-01-01

This presentation is about extending the life of PNGS Pressure tubes from 210,000 EFPH (Effective Full Power Hours) to 247,000 EFPH. The l license renewal is in 2013 (R & D defined in 2009). The unknowns are the extent of degradation in some key areas. There are regulatory requirements for fuel channel extended life. It is necessary to integrate inspection capability and do-ability in outages. This would result in safe operation and demonstrated margin to extend life with operation to 2020.

An update on CANDU fuel channel performance

International Nuclear Information System (INIS)

Price, E.G.; Holt, R.A; Wong, H.; Gautheir, P.; Ellis, P.J.; Slade, J.

1998-01-01

Fuel channel components, particularly pressure tubes, are monitored to establish changes in their condition expected from their service environment. Such monitoring takes the form of the in-service NDE, in-service sampling and periodic tube removal for destructive examination. The pressure tubes in early CANDU 6 stations have now reached mid-life. The expected trends in behaviour of the various properties of the pressure tubes are being confirmed. Diametral deformation potentially limits the life of some pressure tubes, and fuel developments, such as the CANFLEX fuel bundle, will provide additional CHF margins. Corrosion and the associated deuterium pick-up are monitored to assess if any change in rates have occurred. Both have historically continued on an essentially linear trend with time. Recent data indicate that the rates of pick-up could be gradually increasing. Changes in mechanical properties, particularly toughness, have reached saturation levels. In the early reactors, the toughness is acceptable, assuring leak-before-break. The pressure tubes in later reactors have very low concentrations of chlorine and phosphorous and the level of hydrogen are also low, and these tubes are predicted to retain relativity high fracture toughness in-service. Debris fretting wear has occurred in CANDU reactors. Laboratory testing and experience show that no cracks have initiated from debris frets to date. Further debris fretting will have to be avoided by care in maintenance procedures, and existing defects must be monitored in the future as hydrogen concentration levels increase. Spacer location and repositioning (SLAR) has been necessary in earlier reactors with loose spacers. Procedures and equipment for SLAR have been developed and successfully applied. The efficiency of spacer relocation operations has improved with experience. (authors)

Flow behaviour in a CANDU horizontal fuel channel from stagnant subcooled initial conditions

International Nuclear Information System (INIS)

Caplan, M.Z.; Gulshani, P.; Holmes, R.W.; Wright, A.C.D.

1984-01-01

The flow behaviour in a CANDU primary system with horizontal fuel channels is described following a small inlet header break. With the primary pumps running, emergency coolant injection is in the forward direction so that the channel outlet feeders remain warmer than the inlet thereby promoting forward natural circulation. However, the break force opposes the forward driving force. Should the primary pumps run down after the circuit has refilled, there is a break size for which the natural circulation force is balanced by the break force and channels could, theoretically, stagnate. Result of visualization and of full-size channel tests on channel flow behaviour from an initially stagnant channel condition are discussed. After a channel stagnation, the decay power heats the coolant to saturation. Steam is then formed and the coolant stratifies. The steam expands into the subcooled water in the end fitting in a chugging type of flow regime due to steam condensation. After the end fitting reaches the saturation temperature, steam is able to penetrate into the vertical feeder thereby initiating a large buoyancy induced flow which refills the channel. The duration of stagnation is shown to be sensitive to small asymmetries in the initial conditions. A small initial flow can significantly shorten the occurrence and/or duration of boiling as has been confirmed by reactor experience. (author)

Preliminary analysis for u tube degradation in CANDU steam generator using CATHENA

Energy Technology Data Exchange (ETDEWEB)

Shin, So Eun; Lee, Jeong Hun; Park, Tong Kyu; Hwang, Su Hyun [FNC Technology Co., Seoul (Korea, Republic of); Jung, Jong Yeo [KAERI, Daejeon (Korea, Republic of)

2012-10-15

The interest in plant safety and integrity has been increasing due to long term operation of nuclear power plants (NPPs) and lots of efforts have been devoted to developing the degradation evaluation model for all the Structure, System, and Components (SSCs) of NPPs in these days. The efforts, however, were mainly concentrated on pressurized light water reactors (PWRs) in domestic. In contrast, the study for the aging degradation of counterparts of CANDU (CANada Deuterium Uranium) reactors has been rarely performed, even though Wolsong unit 1 (WS1), that is a CANDU 6 NPP in Korea, has been operating for almost 30 years. Therefore, the assessment of the aging degradation is required and the proper and exact evaluation model for the aging degradation of SCCs of CANDU, especially WS1, is urgently needed. In this study, the aging degradation of steam generators (SGs) in WS1 was mainly discussed. Based on cases of the aging degradation of SGs in overseas CANDU reactors, the major potential aging mechanisms of SGs were estimated since there has been no case of accident due to degradation in CANDU NPPs in Korea . Some core parameters which are indicators of the degree of degradation were calculated by CATHENA (Canadian algorithm for thermal hydraulic network analysis). In the result of comparing two calculation cases; core parameters for only aged SGs in fresh plant and those for all the aged component, it can be concluded that aging of SGs is a main component in the degradation assessment of CANDU NPPs, and keeping the integrity of steam generator (SG) tubes is important to guarantee the safety of the NPPs.

Large scale replacement of fuel channels in the Pickering CANDU reactor using a man-in-the-loop remote control system

International Nuclear Information System (INIS)

Stratton, D.

1991-01-01

Spar Aerospace Limited of Toronto is presently under contract to Ontario Hydro to design a Remote Manipulation and Control System (RMCS) to be used during the large scale replacement of the fuel channels in the Pickering A Nuclear Generating Station. The system is designed to support the replacement of all 390 fuel channels in each of the four reactors at the Pickering A station in a safe manner that minimizes worker radiation exposure and unit outage time

Assessments of long term mechanical behavior of CANDU fuel channel by means of PFEM analysis

International Nuclear Information System (INIS)

Florea, S.; Pavelescu, M.

2005-01-01

Structural analysis with finite elements method is today a usual way to evaluate and predict the behavior of structural assemblies exposed to severe conditions, in order to ensure their safety end reliability. CANDU 600 fuel channel is an example in which long time irradiation with implicit consequences on material properties evolution interfere with the corrosion and thermal aggression. A high degree of uncertainty in the evolution of the material's properties must be considered. These are the reasons for developing, in association with deterministic evaluations with computer codes, the probabilistic and statistical methods, in order to predict the structural components response. In INR (Institute of Nuclear Research) in the past years, a code for thermo-mechanical analysis of the fuel channel from CANDU 600 nuclear power plant of Cernavoda was developed using finite element methods (FEM). The CANTUP code evaluate the stress and strain state of the mechanical assembly of fuel channel, considered to be divided into pressure tube, calandria tube and four spacers, supposed to be equidistantly distributed along the pressure tube. The main achievement obtained with this code was the prediction of the long-term behavior of the sag of the pressure tube, by analysis in which the creep phenomenon and the contact between the spacers and calandria tube were considered. This reason has sustained the attempt to estimate the possibility to use this code in order to perform probabilistic evaluations. (authors)

Assessments of long term mechanical behavior of CANDU fuel channel by means of PFEM analysis

International Nuclear Information System (INIS)

Florea, S.; Pavelescu, M.

2005-01-01

Full text: Structural analysis with finite elements method is today a usual way to evaluate and predict the behavior of structural assemblies exposed to severe conditions, in order to ensure their safety end reliability. CANDU 600 fuel channel is an example in which long time irradiation with implicit consequences on material properties evolution interfere with the corrosion and thermal aggression. A high degree of uncertainty in the evolution of the material's properties must be considered. These are the reasons for developing, in association with deterministic evaluations with computer codes, the probabilistic and statistical methods, in order to predict the structural components response. In INR (Institute of Nuclear Research) in the past years, a code for thermo-mechanical analysis of the fuel channel from CANDU 600 nuclear power plant of Cernavoda was developed using finite element methods (FEM). The CANTUP code evaluate the stress and strain state of the mechanical assembly of fuel channel, considered to be divided into pressure tube, calandria tube and four spacers, supposed to be equidistantly distributed along the pressure tube. The main achievement obtained with this code was the prediction of the long-term behavior of the sag of the pressure tube, by analysis in which the creep phenomenon and the contact between the spacers and calandria tube were considered. This reason has sustained the attempt to estimate the possibility to use this code in order to perform probabilistic evaluations. (authors)

Dynamic characteristics of hydrocarbon fuel within the channel at supercritical and pyrolysis condition

Science.gov (United States)

Yu, Bin; Zhou, Weixing; Qin, Jiang; Bao, Wen

2017-12-01

Regenerative cooling with fuel as the coolant is used in the scramjet engine. In order to grasp the dynamic characteristics of engine fuel supply processes, this article studies the dynamic characteristics of hydrocarbon fuel within the channel. A one-dimensional dynamic model was proved, the thermal energy storage effect, fuel volume effect and chemical dynamic effect have been considered in the model, the ordinary differential equations were solved using a 4th order Runge-Kutta method. The precision of the model was validated by three groups of experimental data. The effects of input signal, working condition, tube size on the dynamic characteristics of pressure, flow rate, temperature have been simulated. It is found that cracking reaction increased the compressibility of the fuel pyrolysis mixture and lead to longer responding time of outlet flow. The responding time of outlet flow can reach 3s when tube is 5m long which will greatly influence the control performance of the engine thrust system. Meanwhile, when the inlet flow rate appears the step change, the inlet pressure leads to overshoot, the overshoot can reach as much as 100%, such highly transient impulse will result in detrimental effect on fuel pump.

Enhance performance of micro direct methanol fuel cell by in situ CO2 removal using novel anode flow field with superhydrophobic degassing channels

Science.gov (United States)

Liang, Junsheng; Luo, Ying; Zheng, Sheng; Wang, Dazhi

2017-05-01

Capillary blocking caused by CO2 bubbles in anode flow field (AFF) is one of the bottlenecks for performance improvement of a micro direct methanol fuel cell (μDMFC). In this work, we present a novel AFF structure with nested layout of hydrophilic fuel channels and superhydrophobic degassing channels which can remove most of CO2 from AFF before it is released to the fuel channels. The new AFFs are fabricated on Ti substrates by using micro photochemical etching combined with anodization and fluorination treatments. Performance of the μDMFCs with and without superhydrophobic degassing channels in their AFF is comparatively studied. Results show that the superhydrophobic degassing channels can significantly speed up the exhaust of CO2 from the AFF. CO2 clogging is not observed in the new AFFs even when their comparison AFFs have been seriously blocked by CO2 slugs under the same operating conditions. 55% and 60% of total CO2 produced in μDMFCs with N-serpentine and N-spiral AFF can be respectively removed by the superhydrophobic degassing channels. The power densities of the μDMFCs equipped with new serpentine and spiral AFFs are respectively improved by 30% and 90% compared with those using conventional AFFs. This means that the new AFFs developed in this work can effectively prevent CO2-induced capillary blocking in the fuel channels, and finally significantly improve the performance of the μDMFCs.

Dimension measuring method for channel box

International Nuclear Information System (INIS)

Jo, Hiroto.

1995-01-01

The device of the present invention concerns detection of a channel box for spent fuel assemblies of a BWR type reactor, which measures a cross sectional shape and dimension of the channel box to check deformation amount such as expansion. That is, a customary fuel exchanger and a dimension measuring device are used. The lower end of the channel box is measured by a distance sensor of the dimension measuring device when it is aligned with a position of the distance sensor. The channel box is lowered at the same time while detecting axial position data of the fuel exchanger. The position of the channel box in an axial direction is detected based on axial position data of the fuel exchanger. The lower end of the channel box can accurately be recognized by the detection of both of them. Subsequent deformation measurement for the channel box at accurate axial positions is enabled. In addition, since the axial position data of the fuel exchanger per se are detected, an axial profile of the channel box can be measured even if a lifting speed of the channel box is varied on every region. (I.S.)

On Line Neutron Flux Mapping in Fuel Coolant Channels of a Research Reactor

International Nuclear Information System (INIS)

Barbot, Loic; Domergue, Christophe; Villard, Jean-Francois; Destouches, Christophe; Braoudakis, George; Wassink, David; Sinclair, Bradley; Osborn, John-C.; Wu, Huayou; Blandin, C.; Thevenin, Mathieu; Corre, Gwenole; Normand, Stephane

2013-06-01

This work deals with the on-line neutron flux mapping of the OPAL research reactor. A specific irradiation device has been set up to investigate fuel coolant channels using subminiature fission chambers to get thermal neutron flux profiles. Experimental results are compared to first neutronic calculations and show good agreement (C/E ∼0.97). (authors)

Numerical evaluation of various gas and coolant channel designs for high performance liquid-cooled proton exchange membrane fuel cell stacks

International Nuclear Information System (INIS)

Sasmito, Agus P.; Kurnia, Jundika C.; Mujumdar, Arun S.

2012-01-01

A careful design of gas and coolant channel is essential to ensure high performance and durability of proton exchange membrane (PEM) fuel cell stack. The channel design should allow for good thermal, water and gas management whilst keeping low pressure drop. This study evaluates numerically the performance of various gas and coolant channel designs simultaneously, e.g. parallel, serpentine, oblique-fins, coiled, parallel-serpentine and a novel hybrid parallel-serpentine-oblique-fins designs. The stack performance and local distributions of key parameters are investigated with regards to the thermal, water and gas management. The results indicate that the novel hybrid channel design yields the best performance as it constitutes to a lower pumping power and good thermal, water and gas management as compared to conventional channels. Advantages and limitation of the designs are discussed in the light of present numerical results. Finally, potential application and further improvement of the design are highlighted. -- Highlights: ► We evaluate various gas and coolant channel designs in liquid-cooled PEM fuel cell stack. ► The model considers coupled electrochemistry, channel design and cooling effect simultaneously. ► We propose a novel hybrid channel design. ► The novel hybrid channel design yields the best thermal, water and gas management which is beneficial for long term durability. ► The novel hybrid channel design exhibits the best performance.

BWR fuel assembly having fuel rod spacers axially positioned by exterior springs

International Nuclear Information System (INIS)

Taleyarkhan, R.P.

1988-01-01

In a fuel assembly having spaced fuel rods, an outer hollow tubular flow channel surrounding the fuel rods so as to direct flow of coolant/moderator fluid there-along, and at least one spacer being disposed along the channel and about the fuel rods so as to maintain them in side-by-side spaced relationship, an arrangement for disposing the spacer in a desired axial position along the fuel rods is described comprising: yieldably resilient springs disposed between an interior side of the outer channel and an exterior side of the spacer. The springs have an inherent spring bias directed away from the exterior sides of the spacers and toward the interior side of the channel such that by contact with the channel and spacer the springs assume states in which they are deflected away from the channel interior side so as to exert sufficient compressive contacting force thereon to maintain the spacer substantially stationary in the desired axial position along the fuel rods

Analysis of loss of electrical power with the CATHENA model of the blanket and first wall cooling loop for the SEAFP reactor design

International Nuclear Information System (INIS)

Ross, W.E.

1994-08-01

This report documents the thermosyphoning analysis which was performed with the CATHENA network model of one of the blanket and first wall cooling loops of the SEAFP reactor design. This thermosyphoning analysis is similar to that reported in CFFTP-G--9355, Volume 4 except that a much larger decay power transient is used. Also, the pressurizer heaters are turned off following the loss of electrical power. This analysis is performed to assess the primary heat transport system behaviour for a complete loss of electrical power event (total loss of flow) and to estimate the rate of heatup of the in-core components. A description of the important aspects of the transient thermalhydraulic behaviour including coolant temperatures, circuit and sector flows, circuit pressure, pressurizer level and steam bleed flow, and first wall and blanket temperatures are provided. (author). 8 refs., 2 tabs., 26 figs

Thermal-hydraulic behavior of physical quantities at critical velocities in a nuclear research reactor core channel using plate type fuel

Directory of Open Access Journals (Sweden)

Sidi Ali Kamel

2012-01-01

Full Text Available The thermal-hydraulic study presented here relates to a channel of a nuclear reactor core. This channel is defined as being the space between two fuel plates where a coolant fluid flows. The flow velocity of this coolant should not generate vibrations in fuel plates. The aim of this study is to know the distribution of the temperature in the fuel plates, in the cladding and in the coolant fluid at the critical velocities of Miller, of Wambsganss, and of Cekirge and Ural. The velocity expressions given by these authors are function of the geometry of the fuel plate, the mechanical characteristics of the fuel plate’s material and the thermal characteristics of the coolant fluid. The thermal-hydraulic study is made under steady-state; the equation set-up of the thermal problem is made according to El Wakil and to Delhaye. Once the equation set-up is validated, the three critical velocities are calculated and then used in the calculations of the different temperature profiles. The average heat flux and the critical heat flux are evaluated for each critical velocity and their ratio reported. The recommended critical velocity to be used in nuclear channel calculations is that of Wambsganss. The mathematical model used is more precise and all the physical quantities, when using this critical velocity, stay in safe margins.

Nuclear reactor coolant channels

International Nuclear Information System (INIS)

Macbeth, R.V.

1978-01-01

Reference is made to coolant channels for pressurised water and boiling water reactors and the arrangement described aims to improve heat transfer between the fuel rods and the coolant. Baffle means extending axially within the channel are provided and disposed relative to the fuel rods so as to restrict flow oscillations occurring within the coolant from being propagated transversely to the axis of the channel. (UK)

AECL experience in fuel channel inspection

Energy Technology Data Exchange (ETDEWEB)

Van Drunen, G. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Gunn, R. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Mayo, W.R. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Scott, D.A. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

1999-06-01

Inspection of CANDU fuel channels (FC) is performed to ensure safe and economic reactor operation. CANDU reactor FCs have features that make them a unique non-destructive testing (NDT) challenge. The thin, 4 mm pressure-tube wall means flaws down to about 0.1 mm deep must be reliably detected and characterized. This is one to two orders of magnitude smaller than is usually considered of significant concern for steel piping and pressure vessels. A second unique feature is that inspection sensors must operate in the reactor core--often within 20 cm of highly radioactive fuel. Work on inspection of CANDU reactor FCs at AECL dates back over three decades. In that time, AECL staff have provided equipment and conducted or supervised in-service inspections in about 250 FCs, in addition to over 8000 pre-service FCs. These inspections took place at every existing CANDU reactor except those in India and Romania. Early FC inspections focussed on measurement of changes in dimensions (gauging) resulting from exposure to a combination of neutrons, stress and elevated temperature. Expansion of inspection activities to include volumetric inspection (for flaws) started in the mid-1970s with the discovery of delayed hydride cracking in Pickering 3 and 4 rolled joints. Recognition of other types of flaw mechanisms in the 1980s led to further expansion in both pre-service and in-service inspections. These growing requirements, to meet regulatory as well as economic needs, led to the development of a wide spectrum of inspection technology that now includes tests for hydrogen concentration, structural integrity of core components, flaws, and dimensional change. This paper reviews current CANDU reactor FC inspection requirements. The equipment and techniques developed to satisfy these requirements are also described. The paper concludes with a discussion of work in progress in AECL aimed at providing state-of-the-art FC inspection services. (author)

Measurement of the heavy water level in the fuel channels of the RA reactor - Annex 11

International Nuclear Information System (INIS)

Nikolic, M.

1964-01-01

The objective of measuring the heavy water level in the reactor channels was to verify experimentally the possibilities of reactor cooling with parallel operation of heavy water pumps od 1500 rotations/min at nominal power of 6.5 MW. Measurements were done in 2 periphery and 2 central fuel channels with pumps speed 1500, 1800 and 3000 rotations/min by a contact probe with electric resistance measuring device. precision of the measurement was ±1 cm

Numerical study on channel size effect for proton exchange membrane fuel cell with serpentine flow field

International Nuclear Information System (INIS)

Wang Xiaodong; Yan Weimon; Duan Yuanyuan; Weng Fangbor; Jung Guobin; Lee Chiyuan

2010-01-01

This work numerically investigates the effect of the channel size on the cell performance of proton exchange membrane (PEM) fuel cells with serpentine flow fields using a three-dimensional, two-phase model. The local current densities in the PEM, oxygen mass flow rates and liquid water concentrations at the interface of the cathode gas diffusion layer and catalyst layer were analyzed to understand the channel size effect. The predictions show that smaller channel sizes enhance liquid water removal and increase oxygen transport to the porous layers, which improve cell performance. Additionally, smaller channel sizes also provide more uniform current density distributions in the cell. However, as the channel size decreases, the total pressure drops across the cell increases, which leads to more pump work. With taking into account the pressure losses, the optimal cell performance occurs for a cell with a flow channel cross-sectional area of 0.535 x 0.535 mm 2 .

Channel follower leakage restrictor

International Nuclear Information System (INIS)

Williamson, H.E.; Smith, B.A.

1977-01-01

An improved means is provided to control coolant leakage between the flow channel and the lower tie plate of a nuclear fuel assembly. The means includes an opening in the lower tie plate and a movable element adjacent thereto. The coolant pressure within the tie plate biases the movable means toward the inner surface of the surrounding flow channel to compensate for any movement of the flow channel away from the lower tie plate to thereby control the leakage of coolant flow from the fuel assemblies to the spaces among the fuel assemblies of the core. 9 figures

Channel box

International Nuclear Information System (INIS)

Tanabe, Akira.

1993-01-01

In a channel box of a BWR type reactor, protruding pads are disposed in axial position on the lateral side of a channel box opposing to a control rod and facing the outer side portion of the control rod in a reactor core loaded state. In the initial loading stage of fuel assemblies, channel fasteners and spacer pads are abutted against each other in the upper portion between the channel boxes sandwiching the control rod therebetween. Further, in the lower portion, a gap as a channel for the movement of the control rod is ensured by the support of fuel support metals. If the channel box is bent toward the control rod along with reactor operation, the pads are abutted against each other to always ensure the gap through which the control rod can move easily. Further, when the pads are brought into contact with each other, the bending deformation of the channel box is corrected by urging to each other. Thus, the control rod can always be moved smoothly to attain reactor safety operation. (N.H.)

The performance analysis of direct methanol fuel cells with different hydrophobic anode channels

Science.gov (United States)

Yeh, Hung-Chun; Yang, Ruey-Jen; Luo, Win-Jet; Jiang, Jia-You; Kuan, Yean-Der; Lin, Xin-Quan

In order to enhance the performance of the direct methanol fuel cell (DMFC), the product of CO 2 bubble has to be efficiently removed from the anode channel during the electrochemical reaction. In this study, the materials of Polymethyl Methacrylate (PMMA) with hydrophilic property and polydimethylsiloxane (PDMS) with hydrophobic property are used to form the anode cannel. The channel is fabricated through a microelectromechanical system (MEMS) manufacture process of the DMFCs. In addition, some particles with high hydrophobic properties are added into the PDMS materials in order to further reduce the hydro-resistance in the anode channel. The performance of the DMFCs is investigated under the influence of operation conditions, including operation temperature, flow rate, and methanol concentration. It is found that the performance of the DMFC, which is made of PDMS with high hydrophobic particles, can be greatly enhanced and the hydrophobic property of the particles can be unaffected by different operation conditions.

Computerized representation of experimental data on burnout in tubes, annular channels and fuel bundles

International Nuclear Information System (INIS)

Katan, I.B.; Sal'nikova, O.V.; Vinogradov, V.N.

1983-01-01

Realization of TEFOR formate for presentation in data bases of bibliographic information obtained when studying heat exchange crisis in channels of the most widely spread types (tubes, annular channels, fuel bundles) has been described. The use of the unified formate, providing a possibility to completely describe the information from the initial source, results in standardization of data base formation in different sections of thermal physics and hydrodynamics of NPPs, permits to develop the general apparatus of bank control in the form of packet of applied programs and to use unified techniques, algorithms and programs during calculations with the use of data of the banks

A discussion of hyperbolicity in CATHENA 4. Virtual mass and phase-to-interface pressure differences

International Nuclear Information System (INIS)

Aydemir, Nusret U.

2012-01-01

It is well known that the one-dimensional equations of motion for two-phase flow are non-hyperbolic. Non-hyperbolicity can lead to numerical instabilities, destroying the solution. However, researchers in the last few decades were able to show that inclusion of virtual mass and/or phase-to-interface pressure differences in the momentum equations successfully render the equations of motion hyperbolic. In the present paper, the effect of including virtual mass and phase-to-interface pressure terms in the momentum equations on the hyperbolicity of the two-phase model in the CATHENA 4 code is discussed. The study is motivated by the fact that the inclusion of either model has been shown in the open literature to lead to a hyperbolic system separately. However, no known study exists that examine hyperbolicity in the presence of both these terms in the momentum equations. In this work, both terms are considered in the model equations simultaneously and their implications on the hyperbolicity of the two-phase model are discussed. Specifically, it is shown that in the case of mixed flow, there is a distinct region of non-hyperbolicity that developers need to be aware of when their equations include both the virtual mass and the phase-to-interface terms. Selecting the coefficients of phase-to-interface pressure difference terms properly ensures that the equations are hyperbolic for a wide range of conditions. (orig.)

Sub-channel analysis of LBE-cooled fuel assemblies of accelerator driven systems

International Nuclear Information System (INIS)

Cheng, X.; Hwang, D.H.

2005-01-01

In the frame of the European PDS-XADS project, two concepts of the sub-critical reactor core cooled by liquid lead-bismuth eutectic (LBE) were proposed. In this paper, the local thermal-hydraulic behavior of both LBE-cooled fuel assemblies was analyzed. For this purpose, the sub-channel analysis code MATRA was selected, and modification was made for its applications to XADS conditions. Compared to the small core concept, the large core concept has a much lower temperatures of coolant, cladding and fuel pins. This enables a short-term realization of the core design using available technologies. The high power density of the small core results in high local temperatures of coolant, cladding and fuel. Both coolant velocity and cladding temperature are such that special attention has to be paid to avoid corrosion and erosion damage of cladding materials. A parametric study shows that under the parameters considered, mixing coefficient has the biggest effect on the coolant temperature distribution, whereas the cladding temperature is strongly affected by the selection of heat transfer correlations. (author)

Pressure drop variation as a function of axial and radial power distribution in CANDU fuel channel with standard and CANFLEX 43 bundles

International Nuclear Information System (INIS)

Catana, Alexandru; Department of Energy Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel

2007-01-01

CANDU 600 nuclear reactors are usually fuelled with STANDARD (STD), 37 rods fuel bundles. Natural uranium (NU) dioxide (UO 2 ), is used as fuel composition. A new fuel bundle geometry called CANFLEX (CFX) with 43 rods is proposed and some new fuel composition are considered. Flexibility is the key word for the attempt to use some different fuel geometries and compositions for CANDU 600 nuclear reactors as well as for innovative ACR-700/1000 nuclear reactors. The fuel bundle considered in this paper is CFX-RU-0.90 that encodes the CANFLEX geometry, recycled dioxide uranium (RU) with 0.90% enrichment. The goal of this proposal is ambitious: a higher average discharge burn-up up to 14000 MWd/tU and, for the same amount of generated electric power, reduction in nuclear fuel fabrication, reduction of spent nuclear fuel radioactive waste and reduction of refueling operational work by using fewer bundles. An improved sub-channel approach for thermal-hydraulic analysis is used in this paper to compute some flow parameters, mainly the pressure drop along the CANDU 600 fuel channel when STD or CFX-RU-0.90 fuel bundles. Also an intermediate CFX-NU fuel bundle are used, for gradual comparison. For CFX-RU- 0.90 four fuel bundle shift refueling scheme is used instead of eight, that will determine different axial power distributions. At the same time radial power distribution is affected by the geometry and by the fuel composition of fuel bundle type used. Some other thermal-hydraulic flow parameters will be influenced, too. One of the most important parameter is pressure drop (PD) along the fuel channel because of its importance in drag force evaluation. We start with an axial power distribution, which is characteristic for a refueling scheme of eight or four fuel bundles on a shift. Comparative results are presented between STD37, CFX-NU CFX-RU-0.90 fuel bundles in a CANDU nuclear reactor operating conditions. Neutron flux distribution analysis shows that four bundle shift

Numerical simulations of heat transfer in an annular fuel channel with three-dimensional spacer ribs set up periodically under a fully developed turbulent flow

International Nuclear Information System (INIS)

Takase, Kazuyuki; Akino, Norio

1996-06-01

Thermal-hydraulic characteristics of an annular fuel channel with spacer ribs for high temperature gas-cooled reactors were analyzed numerically by three-dimensional heat transfer computations under a fully developed turbulent flow. The two-equations κ-ε turbulence model was applied to the present turbulent analysis. In particular, the κ-ε turbulence model constants and the turbulent Prandtl number were improved from the previous standard values proposed by Jones and Launder in order to obtain heat transfer predictions with higher accuracy. Consequently, heat transfer coefficients and friction factors in the spacer-ribbed fuel channel were predicted with sufficient accuracy in the range of Reynolds number exceeding 3000. It was clarified quantitatively from the present study that main mechanism for the heat transfer augmentation in the spacer-ribbed fuel channel was combined effects of the turbulence promoter effect by the spacer ribs and the velocity acceleration effect by a reduction in the channel cross-section. (author)

Uncertainty analysis for hot channel

International Nuclear Information System (INIS)

Panka, I.; Kereszturi, A.

2006-01-01

The fulfillment of the safety analysis acceptance criteria is usually evaluated by separate hot channel calculations using the results of neutronic or/and thermo hydraulic system calculations. In case of an ATWS event (inadvertent withdrawal of control assembly), according to the analysis, a number of fuel rods are experiencing DNB for a longer time and must be regarded as failed. Their number must be determined for a further evaluation of the radiological consequences. In the deterministic approach, the global power history must be multiplied by different hot channel factors (kx) taking into account the radial power peaking factors for each fuel pin. If DNB occurs it is necessary to perform a few number of hot channel calculations to determine the limiting kx leading just to DNB and fuel failure (the conservative DNBR limit is 1.33). Knowing the pin power distribution from the core design calculation, the number of failed fuel pins can be calculated. The above procedure can be performed by conservative assumptions (e.g. conservative input parameters in the hot channel calculations), as well. In case of hot channel uncertainty analysis, the relevant input parameters (k x, mass flow, inlet temperature of the coolant, pin average burnup, initial gap size, selection of power history influencing the gap conductance value) of hot channel calculations and the DNBR limit are varied considering the respective uncertainties. An uncertainty analysis methodology was elaborated combining the response surface method with the one sided tolerance limit method of Wilks. The results of deterministic and uncertainty hot channel calculations are compared regarding to the number of failed fuel rods, max. temperature of the clad surface and max. temperature of the fuel (Authors)

Nuclear fuel assembly

International Nuclear Information System (INIS)

Borrman, B.; Nylund, O.

1984-01-01

A fuel assembly with a fuel channel which surrounds a plurality of fuel rods and which is divided, by means of a stiffening device of cruciform cross-section and four wings, into four sub-channels each of which comprises a bundle of fuel rods. Each fuel channel side has a plurality of stamped, inwardly-directed projections, arranged vertically one after the other, aid projections being welded to one and the same stiffening wing. Each one of the wall portions located between the projections defines, together with two adjacently positioned projections and a portion of the stiffening wing, a communiation opening between two bundles located on on one side each of the stiffening wing. (Author)

Monte Carlo simulation of irradiation of MTR fuel plates in the BR2 reactor using a full-scale 3-d model with inclined channels

International Nuclear Information System (INIS)

Kuzminov, V. V; Koonen, E.; Ponsard, B.

2002-01-01

A three-dimensional full-scale Monte Carlo model of the BR2 reactor has been developed for simulation of irradiation conditions of materials and fuel loaded in various irradiation devices. This new reactor model includes a detailed geometrical description of the inclined reactor channels, the irradiation devices loaded in these channels including the materials to be tested/loaded in these devices, the burn-up of the BR2 fuel elements and the poisoning of the beryllium matrix. Recently a benchmark irradiation of new irradiation device for testing and qualification of MTR fuel plates has been performed. For this purpose the detailed irradiation conditions of fuel plates had to be predetermined. Monte Carlo calculations of neutron fluxes and heat load distributions in irradiated MTR fuel plates were performed taking into account the contents of all loaded experimental devices in the reactor channels. A comparison of the calculated and measured values of neutron fluxes and of heat loads in the BR2 reactor is presented in this paper. The comparison is part of the validation process of the new reactor model. It also serves to establish the capability to conduct a fuel plate irradiation program under requested and well- known irradiation conditions. (author)

Monitoring the mechanical vibration of in-core detector tubes and fuel channels via ICFD noise analysis

International Nuclear Information System (INIS)

Glockler, O.; Cooke, D.F.; Czuppon, G.J.; Kapoor, K.K.

2000-01-01

Vibrations of core internals are regularly monitored in the CANDU nuclear generating stations of Ontario Power Generation (OPG) via the noise analysis of in-core flux detectors (ICFDs). Voltage signals of standard station instrumentation are recorded by portable multi-channel high-speed high-resolution data acquisition systems, then statistical parameters are derived from the multi-channel time series measurements. Reactor noise analysis is a non-intrusive statistical technique regularly used in system surveillance, diagnostics and in actual operational I and C problems. It utilizes the dynamic information carried by the small fluctuations (noise) of station signals measured around their mean values during steady-state operation. The present paper discusses specific results related to the flow-induced mechanical vibrations of detector tubes and fuel channels. (author)

Progress of the DUPIC fuel compatibility analysis (II) - thermal-hydraulics

Energy Technology Data Exchange (ETDEWEB)

Park, Joo Hwan; Choi, Hang Bok

2005-03-01

Thermal-hydraulic compatibility of the DUPIC fuel bundle with a 713 MWe Canada deuterium uranium (CANDU-6) reactor was studied by using both the single channel and sub-channel analysis methods. The single channel analysis provides the fuel channel flow rate, pressure drop, critical channel power, and the channel exit quality, which are assessed against the thermal-hydraulic design requirements of the CANDU-6 reactor. The single channel analysis by the NUCIRC code showed that the thermal-hydraulic performance of the DUPIC fuel is not different from that of the standard CANDU fuel. Regarding the local flow characteristics, the sub-channel analysis also showed that the uncertainty of the critical channel power calculation for the DUPIC fuel channel is very small. As a result, both the single and sub-channel analyses showed that the key thermal-hydraulic parameters of the DUPIC fuel channel do not deteriorate compared to the standard CANDU fuel channel.

Numerical simulations of carbon monoxide poisoning in high temperature proton exchange membrane fuel cells with various flow channel designs

International Nuclear Information System (INIS)

Jiao, Kui; Zhou, Yibo; Du, Qing; Yin, Yan; Yu, Shuhai; Li, Xianguo

2013-01-01

Highlights: ► Simulations of CO poisoning in HT-PEMFC with different flow channels are conducted. ► Parallel and serpentine designs result in least and most CO effects, respectively. ► General CO distributions in CLs are similar with different flow channel designs. - Abstract: The performance of high temperature proton exchange membrane fuel cell (HT-PEMFC) is significantly affected by the carbon monoxide (CO) in hydrogen fuel, and the flow channel design may influence the CO poisoning characteristics by changing the reactant flow. In this study, three-dimensional non-isothermal simulations are carried out to investigate the comprehensive flow channel design and CO poisoning effects on the performance of HT-PEMFCs. The numerical results show that when pure hydrogen is supplied, the interdigitated design produces the highest power output, the power output with serpentine design is higher than the two parallel designs, and the parallel-Z and parallel-U designs have similar power outputs. The performance degradation caused by CO poisoning is the least significant with parallel flow channel design, but the most significant with serpentine and interdigitated designs because the cross flow through the electrode is stronger. At low cell voltages (high current densities), the highest power outputs are with interdigitated and parallel flow channel designs at low and high CO fractions in the supplied hydrogen, respectively. The general distributions of absorbed hydrogen and CO coverage fractions in anode catalyst layer (CL) are similar for the different flow channel designs. The hydrogen coverage fraction is higher under the channel than under the land, and is also higher on the gas diffusion layer (GDL) side than on the membrane side; and the CO coverage distribution is opposite to the hydrogen coverage distribution

Neutron-physical characteristics of the TVRM-100 reactor with ten ring fuel channels

International Nuclear Information System (INIS)

Mikhajlov, V.M.; Myrtsymova, L.A.

1988-01-01

Three-dimensional heterogeneous calculations of TVRM-100 reactor which is a research reactor using enriched fuel with heavy-water moderator, coolant and reflector, are conducted. Achievable burnup depths depending on the number of removable FAs are presented. The maximum non-perturbed thermal neutron flux in the reflector is (2-1.8)x10 15 cm -2 c -1 ; mean flux on the fuel is 2.9x10 14 cm -2 c -1 . Energy release radial non-uniformity is 0.67, maximum bending by FA is ∼3.7. Reactivity temperature effect is negative and is equal to - 0.9x10 -4 grad -1 without accounting for experimental channels. Control rod efficiency in the radial reflector is high, but their location dose to experimental devices in the high neutron flux area is undesirable. 4 refs.; 5 figs

Mechanistic modeling of heat transfer process governing pressure tube-to-calandria tube contact and fuel channel failure

International Nuclear Information System (INIS)

Luxat, J.C.

2002-01-01

Heat transfer behaviour and phenomena associated with ballooning deformation of a pressure tube into contact with a calandria tube have been analyzed and mechanistic models have been developed to describe the heat transfer and thermal-mechanical processes. These mechanistic models are applied to analyze experiments performed in various COG funded Contact Boiling Test series. Particular attention is given in the modeling to characterization of the conditions for which fuel channel failure may occur. Mechanistic models describing the governing heat transfer and thermal-mechanical processes are presented. The technical basis for characterizing parameters of the models from the general heat transfer literature is described. The validity of the models is demonstrated by comparison with experimental data. Fuel channel integrity criteria are proposed which are based upon three necessary and sequential mechanisms: Onset of CHF and local drypatch formation at contact; sustained film boiling in the post-contact period; and creep strain to failure of the calandria tube while in sustained film boiling. (author)

A Study of the Influence of Gas Channel Parameters on HT-PEM Fuel Cell Performance Using FEM Analysis

Directory of Open Access Journals (Sweden)

Ionescu Viorel

2016-01-01

Full Text Available Proton Exchange Membrane Fuel Cells (PEMFC are highly efficient power generators, achieving up to 50–60% conversion efficiency, even in sizes of a few kilowatts. Comsol Multiphysics, a commercial solver based on the Finite Element Method (FEM was used for developing a three dimensional model of a high temperature PEMFC that can deal with both anode and cathode flow field for examining the micro flow channel with electrochemical reaction. Cathode gas flow velocity influence on the cell performance was investigated at first. Polarization curves for three different channel widths (0.8, 1.6 and 2.4 mm and three different channel depths (1, 2 and 3 mm were computed at a cathode inlet flow velocity of 0.06 m/s. Oxygen molar concentration at cathode catalyst layer-GDL channel interface and local current density variation along the cell length were also studied for specific gas channel geometries.

Investigation of flow blockage in a fuel channel with the ASSERT subchannel code

International Nuclear Information System (INIS)

Harvel, G.D.; Dam, R.; Soulard, M.

1996-01-01

On behalf of New Brunswick Power, a study was undertaken to determine if safe operation of a CANDU-6 reactor can be maintained at low reactor powers with the presence of debris in the fuel channels. In particular, the concern was to address if a small blockage due to the presence of debris would cause a significant reduction in dryout powers, and hence, to determine the safe operation power level to maintain dryout margins. In this work the NUCIRC(1,2), ASSERT-IV(3), and ASSERT-PV(3) computer codes are used in conjunction with a pool boiling model to determine the safe operation power level which maintains dryout safety margins. NUCIRC is used to provide channel boundary conditions for the ASSERTcodes and to select a representative channel for analysis. This pool boiling model is provided as a limiting lower bound analysis. As expected, the ASSERT results predict higher CHF ratios than the pool boiling model. In general, the ASSERT results show that as the model comes closer to modelling a complete blockage it reduces toward, but does not reach the pool boiling model. (author)

Assessment of conditions of the spent nuclear fuel stored in the stainless steel channel-holders

International Nuclear Information System (INIS)

Pesic, M.; Sotic, O.; Cupac, S.; Maksin, T.; Dasic, N.

2003-01-01

The IAEA technical co-operation project 'Safe Removal of Spent Fuel of the Vinca RA Research Reactor' is carried out at the Vinca Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since January 2003. Present activities will provide up-to-date information on the conditions of the spent nuclear fuel, stored in the stainless steel channel-holders ('chekhols') and on the water quality in the storage basins. Water samples taken out from the chekhols and the basins are measured to determine their activity and chemical parameters. Until September 2003, about 1/3 of the chekhols containing spent fuel elements with initial enrichment of 2% and 80% of uranium were inspected. High activity of Cs-137 was found in several water samples taken out from chekhols. All water samples show very high electrical conductivity, while those taken from the basins show the presence of chlorides and aluminium ions, too. Information on established procedures and measuring results are given in this paper. The obtained results, so far, show that the spent nuclear fuel elements are leaking in about 10% of chekhols. (author)

SARAPAN—A Simulated-Annealing-Based Tool to Generate Random Patterned-Channel-Age in CANDU Fuel Management Analyses

Directory of Open Access Journals (Sweden)

Doddy Kastanya

2017-02-01

Full Text Available In any reactor physics analysis, the instantaneous power distribution in the core can be calculated when the actual bundle-wise burnup distribution is known. Considering the fact that CANDU (Canada Deuterium Uranium utilizes on-power refueling to compensate for the reduction of reactivity due to fuel burnup, in the CANDU fuel management analysis, snapshots of power and burnup distributions can be obtained by simulating and tracking the reactor operation over an extended period using various tools such as the *SIMULATE module of the Reactor Fueling Simulation Program (RFSP code. However, for some studies, such as an evaluation of a conceptual design of a next-generation CANDU reactor, the preferred approach to obtain a snapshot of the power distribution in the core is based on the patterned-channel-age model implemented in the *INSTANTAN module of the RFSP code. The objective of this approach is to obtain a representative snapshot of core conditions quickly. At present, such patterns could be generated by using a program called RANDIS, which is implemented within the *INSTANTAN module. In this work, we present an alternative approach to derive the patterned-channel-age model where a simulated-annealing-based algorithm is used to find such patterns, which produce reasonable power distributions.

SARAPAN-A simulated-annealing-based tool to generate random patterned-channel-age in CANDU fuel management analyses

Energy Technology Data Exchange (ETDEWEB)

Kastanya, Doddy [Safety and Licensing Department, Candesco Division of Kinectrics Inc., Toronto (Canada)

2017-02-15

In any reactor physics analysis, the instantaneous power distribution in the core can be calculated when the actual bundle-wise burnup distribution is known. Considering the fact that CANDU (Canada Deuterium Uranium) utilizes on-power refueling to compensate for the reduction of reactivity due to fuel burnup, in the CANDU fuel management analysis, snapshots of power and burnup distributions can be obtained by simulating and tracking the reactor operation over an extended period using various tools such as the *SIMULATE module of the Reactor Fueling Simulation Program (RFSP) code. However, for some studies, such as an evaluation of a conceptual design of a next-generation CANDU reactor, the preferred approach to obtain a snapshot of the power distribution in the core is based on the patterned-channel-age model implemented in the *INSTANTAN module of the RFSP code. The objective of this approach is to obtain a representative snapshot of core conditions quickly. At present, such patterns could be generated by using a program called RANDIS, which is implemented within the *INSTANTAN module. In this work, we present an alternative approach to derive the patterned-channel-age model where a simulated-annealing-based algorithm is used to find such patterns, which produce reasonable power distributions.

Workbench experiments on interaction of nuclear fuel with channel reactor materials: the LFCM congestions criticality and accident scenario in both re-examinations

International Nuclear Information System (INIS)

Baryakhtar, V.; Gonchar, V.; Zhidkov, A.

2002-01-01

The heavy radioecological consequences of 1986 accident were mainly stipulated by destruction of both a significant part of fuel envelopes and fuel matrix due to high-temperature interaction with silicates, when nuclear fuel has lost an important property to keep the fission products inside its volume. In this respect the silicate material application in thermal-insulating filling is a crucial fault had been made when Chornobyl NPP channel reactor designing

Remarks to the hot channel power characteristics

International Nuclear Information System (INIS)

Tinka, I.; Tinkova, E.

2002-01-01

In connection with methodological improvements of safety analyses, some effects of detail power distributions, that should be taken into account for the hot channel characteristics determination, have been studied. This determination concerns the whole channel power (power of the fuel rod) and its axial (along the channel) and radial (across the fuel pellet radius) distribution. The total power of the channel is studied from the point of possible restrictions for different numbers of main cooling loops in operation. For radial power distribution the effect of burnup has been studied and for axial distribution the effect of the control rod vicinity (its coupler part) has been evaluated. The DNBR and fuel temperatures have been the key safety parameters influenced by these hot channel characteristics and have been evaluated in this study (Authors)

Electrochemical cell apparatus having axially distributed entry of a fuel-spent fuel mixture transverse to the cell lengths

Science.gov (United States)

Reichner, Philip; Dollard, Walter J.

1991-01-01

An electrochemical apparatus (10) is made having a generator section (22) containing axially elongated electrochemical cells (16), a fresh gaseous feed fuel inlet (28), a gaseous feed oxidant inlet (30), and at least one gaseous spent fuel exit channel (46), where the spent fuel exit channel (46) passes from the generator chamber (22) to combine with the fresh feed fuel inlet (28) at a mixing apparatus (50), reformable fuel mixture channel (52) passes through the length of the generator chamber (22) and connects with the mixing apparatus (50), that channel containing entry ports (54) within the generator chamber (22), where the axis of the ports is transverse to the fuel electrode surfaces (18), where a catalytic reforming material is distributed near the reformable fuel mixture entry ports (54).

Method for monitoring stability of channel within a core in a reactor

International Nuclear Information System (INIS)

Monta, Kazuo; Takigawa, Yukio.

1976-01-01

Object: To obtain a flow rate as a factor for determining a safety limit of hydraulic vibration in a fuel channel within a core from signals of an incore neutron detector every channel to thereby monitor stability of the fuel channel. Structure: On the basis of hydraulic data of fuel channels such as power distribution and flow distribution obtained in each fuel channel, average pressure of fuel channels, measured value relating to inlet sub-cleaning of recycling water and throttling of inlet and outlet orifices, discrimination of stability is effected by a channel stability monitoring device, or on the basis of comparison between the limit value of stability in connecting with those parameters designated among parameters and the actual value thereof, determination of stability allowance is carried out. (Yoshihara, H.)

Fuel assembly for BWR type reactor

International Nuclear Information System (INIS)

Ueda, Makoto

1990-01-01

Various considerations are applied to fuel rods for improving the fuel burnup degree. If a gap between the fuel rods is changed, this varies the easiness for the flow of coolants depending on places, to reduce the thermal margin. Then, it is noted for the distribution of stresses generated due to the difference of water pressure caused by the difference of water streams between the inside and the outside of a channel box, and composite value, of stresses upon occurrence of earthquakes, neutron irradiation and a channel creep phenomenon caused by the stresses of due to the water pressure difference described above, the thickness of the channel box is increased in the upstream and decreased toward the downstream. Further, fuel spacers at the position where the thickness of the channel box is changed are spaced apart from the channel box so as not to brought into contact with the channel box. This can contribute to the reduction of coolants pressure loss, improvement of critical power and improvement of reactivity, as well as remarkably moderate local stresses applied from the fuel spacers to the channel box due to horizontal vibrations upon occurrence of earthquakes to improve the integrity of fuel assembly. (N.H.)

HANARO core channel flow-rate measurement

Energy Technology Data Exchange (ETDEWEB)

Kim, Heon Il; Chae, Hee Tae; Im, Don Soon; Kim, Seon Duk [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-06-01

HANARO core consists of 23 hexagonal flow tubes and 16 cylindrical flow tubes. To get the core flow distribution, we used 6 flow-rate measuring dummy fuel assemblies (instrumented dummy fuel assemblies). The differential pressures were measured and converted to flow-rates using the predetermined relationship between AP and flow-rate for each instrumented dummy fuel assemblies. The flow-rate for the cylindrical flow channels shows +-7% relative errors and that for the hexagonal flow channels shows +-3.5% relative errors. Generally the flow-rates of outer core channels show smaller values compared to those of inner core. The channels near to the core inlet pipe and outlet pipes also show somewhat lower flow-rates. For the lower flow channels, the thermal margin was checked by considering complete linear power histories. From the experimental results, the gap flow-rate was estimated to be 49.4 kg/s (cf. design flow of 50 kg/s). 15 tabs., 9 figs., 10 refs. (Author) .new.

Experimental test results of multi-channel test rig of T1 test section, 5

International Nuclear Information System (INIS)

Hino, Ryutaro; Takase, Kazuyuki; Miyamoto, Yoshiaki

1990-09-01

Channel blockage test on a fuel column of the high temperature engineering test reactor (HTTR) has been performed under the helium gas atmosphere at a high temperature and a high pressure in order to obtain safety data on flow rate and temperature distributions in the fuel column with the multi-channel test rig of the fuel stack test section (T 1 ) in HENDEL. In the test, one of 12 fuel channels was blockaded to 90% of flow area at the channel inlet. Experimental results showed that the helium gas flow rate in the blockaded channel was 28%∼33% lower than the average flow rate for Reynolds number from 2300 to 14000 in isothermal flow. When simulated fuel rods were heated, the flow rate in the blockaded channel did not decrease down in comparison with the isothermal flow. This is due to that the heat generated in the fuel rods conducts to the other fuel channels in graphite fuel blocks, so that accelerated pressure losses in the fuel channels change with helium gas temperatures. (author)

Dynamic behavior of liquid water transport in a tapered channel of a proton exchange membrane fuel cell cathode

NARCIS (Netherlands)

Akhtar, N.; Kerkhof, P.J.A.M.

2011-01-01

A numerical model of a proton exchange membrane fuel cell (PEMFC) cathode with a tapered channel design has been developed in order to examine the dynamic behavior of liquid water transport. Three-dimensional, transient simulations employing the level-set method (available in COMSOL 3.5a, a

Fuel assemblies

International Nuclear Information System (INIS)

Nagano, Mamoru; Yoshioka, Ritsuo

1983-01-01

Purpose: To effectively utilize nuclear fuels by increasing the reactivity of a fuel assembly and reduce the concentration at the central region thereof upon completion of the burning. Constitution: A fuel assembly is bisected into a central region and a peripheral region by disposing an inner channel box within a channel box. The flow rate of coolants passing through the central region is made greater than that in the peripheral region. The concentration of uranium 235 of the fuel rods in the central region is made higher. In such a structure, since the moderating effect in the central region is improved, the reactivity of the fuel assembly is increased and the uranium concentration in the central region upon completion of the burning can be reduced, fuel economy and effective utilization of uranium can be attained. (Kamimura, M.)

Separating rings detection in fuel channels of Embalse NPP

International Nuclear Information System (INIS)

Obrutsky, L.S.; Otero, P.A.; Schmidt, O.A.

1988-01-01

The design specifications of Embalse Nuclear Power Plants (CANDU Type Reactor 600Mw) define the positions to be taken by 4 separating rings of the fuel channels. Experience has demonstrated the displacement possibility of the above mentioned rings. It means a risk of contact between pressure tube and calandria tube. In order to determine the position of separating rings, an inspection system based on Eddy Currents technique was developed by CNEA personnel. Detection is performed through two special probes operating according the ''emitter-receiver'' principle. Obtained signals and its relative position are recorded in a video tape and registered in paper. The probe is telecommanded by an automatic equipment. In this paper the construction and calibration of the detection equipment is described, as well as the propulsion. Final results are also outlined in the inspection carried out in November 1986 when an effective displacement of separating rings was verified from its design position in most of the inspected tubes

Review of the Thermal-Chemical Experiments for CANDU Fuel Channel

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyoung Tae; Min, Byung Joo; Park, Joo Hwan; Yoon, Churl; Rhee, Bo Wook

2005-08-15

In the present study, thermal-chemical experiments for CANDU channel analysis are reviewed. First, 11 experiments are identified from the present references and classified according to the number of heater rods, channel orientation, and degree of FES (Fuel Element Simulator) temperature rise during transient. The main configuration of the test rigs and the position of the measurement systems are identified. The experiments were generally conducted in three stages, a low-power, a high-power and a no-power stage. These test procedures are classified and described in this document. The experimental conditions for steam, coolant, and heat power are identified. The thermal properties of solid materials and fluids in the test apparatus are listed in the tables. From the review of the main test results, the following conclusions are to be obtained. Some of the reviewed experiments were not in the quasy-steady state conditions at a low-power stage and followed by a high-power stage. Zircaloy/steam reaction started when FES temperature were 800 .deg. C and escalated when temperature exceeded 1150 .deg. C. Uncontrolled temperature escalations due to Zircaloy/steam reaction were not observed when the FES temperature reached peak point (just below the melting point) and electric power to the test section shut off (self-sustaining Zircaloy/steam reaction). There were negligible circumferential temperature gradients in the FES bundle and pressure tube for the experiments performed in a vertical channel orientation. There were, however, noticeable circumferential gradients when the pressure tube was horizontal. These gradients were attributed to slumping of the FES bundle (sagging). Sagging of the bundle may have masked any buoyancy induced temperature gradients. Furthermore, the hot FES sagged towards the pressure tube transferring more heat to the pressure tube and increasing the temperature of the pressure tube.

Review of the Thermal-Chemical Experiments for CANDU Fuel Channel

International Nuclear Information System (INIS)

Kim, Hyoung Tae; Min, Byung Joo; Park, Joo Hwan; Yoon, Churl; Rhee, Bo Wook

2005-08-01

In the present study, thermal-chemical experiments for CANDU channel analysis are reviewed. First, 11 experiments are identified from the present references and classified according to the number of heater rods, channel orientation, and degree of FES (Fuel Element Simulator) temperature rise during transient. The main configuration of the test rigs and the position of the measurement systems are identified. The experiments were generally conducted in three stages, a low-power, a high-power and a no-power stage. These test procedures are classified and described in this document. The experimental conditions for steam, coolant, and heat power are identified. The thermal properties of solid materials and fluids in the test apparatus are listed in the tables. From the review of the main test results, the following conclusions are to be obtained. Some of the reviewed experiments were not in the quasy-steady state conditions at a low-power stage and followed by a high-power stage. Zircaloy/steam reaction started when FES temperature were 800 .deg. C and escalated when temperature exceeded 1150 .deg. C. Uncontrolled temperature escalations due to Zircaloy/steam reaction were not observed when the FES temperature reached peak point (just below the melting point) and electric power to the test section shut off (self-sustaining Zircaloy/steam reaction). There were negligible circumferential temperature gradients in the FES bundle and pressure tube for the experiments performed in a vertical channel orientation. There were, however, noticeable circumferential gradients when the pressure tube was horizontal. These gradients were attributed to slumping of the FES bundle (sagging). Sagging of the bundle may have masked any buoyancy induced temperature gradients. Furthermore, the hot FES sagged towards the pressure tube transferring more heat to the pressure tube and increasing the temperature of the pressure tube

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)

Numerical Study of Bubble Coalescence and Breakup in the Reactor Fuel Channel with a Vaned Grid

Directory of Open Access Journals (Sweden)

Tenglong Cong

2018-01-01

Full Text Available The characteristics of bubbles of different sizes in fuel assembly are vital to two-phase flow resistance and heat transfer capacity. However, due to the swirl flow caused by the mixing vane, bubbles can crowd at the heated surface, which may anticipate the occurrence of departure from nucleation boiling. In the current work, the adiabatic two-phase flow in a simplified fuel assembly was analyzed by using the Eulerian two-fluid model and the MUSIG (MUltiple SIze Group model. This computational domain consists of two coolant channels and two sets of vaned spacers, with three sets of periodic boundary conditions at the side faces of the domain. The distributions of vapor phase and bubble diameters were obtained, based on which the effects of mixing vanes on the bubble characteristics were analyzed. Vapor phase crowded at the rod surface in the higher inlet vapor fraction case, but crowded in the channel center in the lower inlet vapor fraction cases. This work can be used as a reference for the design of mixing vanes to avoid the anticipation of departure of nucleation boiling that may be caused by unreasonable design.

Tight connection between fission gas discharge channels

International Nuclear Information System (INIS)

Jung, W.; Peehs, M.; Rau, P.; Krug, W.; Stechemesser, H.

1978-01-01

The invention is concerned with the tight connection between the fission gas discharge channel, leading away from the support plate of a gas-cooled reactor, and the top of the fuel element suspended from this support plate. The closure is designed to be gas-tight for the suspended as well as for the released fuel element. The tight connection has got an annular body resting on the core support plate in the mouth region of the fission gas discharge channel. This body is connected with the fission gas discharge channel in the fuel element top fitting via a gas-tight part and supported by a compression spring. Care is taken for sealing if the fuel element is removal. (RW) [de

Fuel handling problems at KANUPP

International Nuclear Information System (INIS)

Ahmed, I.; Mazhar Hasan, S.; Mugtadir, A.

1991-01-01

KANUPP experienced two abnormal fuel and fuel handling related problems during the year 1990. One of these had arisen due to development of end plate to end plate coupling between the two bundles at the leading end of the fuel string in channel HO2-S. The incident occurred when attempts were being made to fuel this channel. Due to pulling of sticking bundles into the acceptor fuelling machine (north) magazine, which was not designed to accommodate two bundles, a magazine rotary stop occurred. The forward motion of the charge tube was simultaneously discovered to be restricted. The incident led to stalling of fuelling machine locked on to the channel HO2, necessitating a reactor shut down. Removal of the fuelling machine was accomplished sometime later after draining of the channel. The second incident which made the fuelling of channel KO5-N temporarily inexecutable, occurred during attempts to remove its north end shield plug when this channel came up for fuelling. The incident resulted due to breaking of the lugs of the shield plug, making its withdrawal impossible. The Plant however kept operating with suspended fuelling of channel KO5, until it could no longer sustain a further increase in fuel burnup at the maximum rating position. Resolving both these problems necessitated draining of the respective channels, leaving the resident fuel uncovered for the duration of the associated operation. Due to substantial difference in the oxidation temperatures Of UO 2 and Zircaloy and its influence as such on the cooling requirement, it was necessary either to determine explicitly that the respective channels did not contain defective fuel bundles or wait for time long enough to allow the decay heat to reduce to manageable proportions. This had a significant bearing on the Plant down time necessary for the rectification of the problems. This paper describes the two incidents in detail and dwells upon the measures adopted to resolve the related problems. (author)

Fuel handling problems at KANUPP

Energy Technology Data Exchange (ETDEWEB)

Ahmed, I; Mazhar Hasan, S; Mugtadir, A [Karachi Nuclear Power Plant (KANUPP), Karachi (Pakistan)

1991-04-01

KANUPP experienced two abnormal fuel and fuel handling related problems during the year 1990. One of these had arisen due to development of end plate to end plate coupling between the two bundles at the leading end of the fuel string in channel HO2-S. The incident occurred when attempts were being made to fuel this channel. Due to pulling of sticking bundles into the acceptor fuelling machine (north) magazine, which was not designed to accommodate two bundles, a magazine rotary stop occurred. The forward motion of the charge tube was simultaneously discovered to be restricted. The incident led to stalling of fuelling machine locked on to the channel HO2, necessitating a reactor shut down. Removal of the fuelling machine was accomplished sometime later after draining of the channel. The second incident which made the fuelling of channel KO5-N temporarily inexecutable, occurred during attempts to remove its north end shield plug when this channel came up for fuelling. The incident resulted due to breaking of the lugs of the shield plug, making its withdrawal impossible. The Plant however kept operating with suspended fuelling of channel KO5, until it could no longer sustain a further increase in fuel burnup at the maximum rating position. Resolving both these problems necessitated draining of the respective channels, leaving the resident fuel uncovered for the duration of the associated operation. Due to substantial difference in the oxidation temperatures Of UO{sub 2} and Zircaloy and its influence as such on the cooling requirement, it was necessary either to determine explicitly that the respective channels did not contain defective fuel bundles or wait for time long enough to allow the decay heat to reduce to manageable proportions. This had a significant bearing on the Plant down time necessary for the rectification of the problems. This paper describes the two incidents in detail and dwells upon the measures adopted to resolve the related problems. (author)

Three-dimensional single-channel thermal analysis of fully ceramic microencapsulated fuel via two-temperature homogenized model

International Nuclear Information System (INIS)

Lee, Yoonhee; Cho, Nam Zin

2014-01-01

Highlights: • Two-temperature homogenized model is applied to thermal analysis of fully ceramic microencapsulated (FCM) fuel. • Based on the results of Monte Carlo calculation, homogenized parameters are obtained. • 2-D FEM/1-D FDM hybrid method for the model is used to obtain 3-D temperature profiles. • The model provides the fuel-kernel and SiC matrix temperatures separately. • Compared to UO 2 fuel, the FCM fuel shows ∼560 K lower maximum temperatures at steady- and transient states. - Abstract: The fully ceramic microencapsulated (FCM) fuel, one of the accident tolerant fuel (ATF) concepts, consists of TRISO particles randomly dispersed in SiC matrix. This high heterogeneity in compositions leads to difficulty in explicit thermal calculation of such a fuel. For thermal analysis of a fuel element of very high temperature reactors (VHTRs) which has a similar configuration to FCM fuel, two-temperature homogenized model was recently proposed by the authors. The model was developed using particle transport Monte Carlo method for heat conduction problems. It gives more realistic temperature profiles, and provides the fuel-kernel and graphite temperatures separately. In this paper, we apply the two-temperature homogenized model to three-dimensional single-channel thermal analysis of the FCM fuel element for steady- and transient-states using 2-D FEM/1-D FDM hybrid method. In the analyses, we assume that the power distribution is uniform in radial direction at steady-state and that in axial direction it is in the form of cosine function for simplicity. As transient scenarios, we consider (i) coolant inlet temperature transient, (ii) inlet mass flow rate transient, and (iii) power transient. The results of analyses are compared to those of conventional UO 2 fuel having the same geometric dimension and operating conditions

Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

Directory of Open Access Journals (Sweden)

Itamar Iliuk

2016-01-01

Full Text Available Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermal-hydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was proposed. A partial loss of flow accident was simulated to show the behavior of the model under this type of accident. The results show that the critical heat flux was detected in the central region along the axial direction of the plate when the right water channel was blocked.

Opposed-Flow Flame Spread over Thin Solid Fuels in a Narrow Channel under Different Gravity

Science.gov (United States)

Zhang, Xia; Yu, Yong; Wan, Shixin; Wei, Minggang; Hu, Wen-Rui

Flame spread over solid surface is critical in combustion science due to its importance in fire safety in both ground and manned spacecraft. Eliminating potential fuels from materials is the basic method to protect spacecraft from fire. The criterion of material screening is its flamma-bility [1]. Since gas flow speed has strong effect on flame spread, the combustion behaviors of materials in normal and microgravity will be different due to their different natural convec-tion. To evaluate the flammability of materials used in the manned spacecraft, tests should be performed under microgravity. Nevertheless, the cost is high, so apparatus to simulate mi-crogravity combustion under normal gravity was developed. The narrow channel is such an apparatus in which the buoyant flow is restricted effectively [2, 3]. The experimental results of the horizontal narrow channel are consistent qualitatively with those of Mir Space Station. Quantitatively, there still are obvious differences. However, the effect of the channel size on flame spread has only attracted little attention, in which concurrent-flow flame spread over thin solid in microgravity is numerically studied[4], while the similarity of flame spread in different gravity is still an open question. In addition, the flame spread experiments under microgravity are generally carried out in large wind tunnels without considering the effects of the tunnel size [5]. Actually, the materials are always used in finite space. Therefore, the flammability given by experiments using large wind tunnels will not correctly predict the flammability of materials in the real environment. In the present paper, the effect of the channel size on opposed-flow flame spread over thin solid fuels in both normal and microgravity was investigated and compared. In the horizontal narrow channel, the flame spread rate increased before decreased as forced flow speed increased. In low speed gas flows, flame spread appeared the same trend as that in

Likelihood and consequences of fuel string compression at Point Lepreau Generating Station

International Nuclear Information System (INIS)

Reid, P.J.; Gibb, R.A.

1996-01-01

During an accident which results in fuel heatup, axial thermal expansion of the fuel string relative to the pressure tube will occur. If the temperature transient is sufficiently severe, the fuel string may contact the shieled plugs at both ends of the channel. Any additional axial thermal expansion will result in deformation of fuel and fuel channel components, leading to tensile or compressive stresses in the different fuel channel components. If these loads become sufficiently large, they could result in failure of a fuel channel component or to channel failure due to bending of a fuel element under load. The analysis described in this paper demonstrates that this process would not result in fuel channel failure for a design basis accident at Point Lepreau Generating Station (PLGS), even if the station were retubed to 'as-built' channel lengths. (author)

Defueled channel experiments in ZED-2 in support of ACR-1000 ROP analysis

International Nuclear Information System (INIS)

LaFontaine, M.W.R.; Zeller, M.B.; McPhee, G.P.

2011-01-01

Defueled channel experiments were performed in ZED-2 to help resolve discrepancies between calculated flux detector response during refueling in ACR-1000 according the reactor codes RFSP and MCNP. The data produced from these experiments was later used in a separate Regional-Over-Power (ROP) analysis to verify MCNP and RFSP neutron response predictions during refueling. These experiments provided information on thermal flux distributions interior and exterior to a fueled and defueled channel; and on epithermal absolute flux distributions exterior to the same channel. Critical height and moderator temperature data for fueled and defueled channel conditions were also measured. In addition, standard platinum-clad Inconel Self-Powered Detector (SPD) performance data was obtained. The following reactor physics and SPD parameters were measured in these experiments: C Radial flux distribution inside the channel of interest (fueled and defueled), C Radial flux distribution outside the channel of interest (fueled and defueled), C Epithermal radial flux distribution outside the channel of interest (fueled and defueled), and C SPD response parallel to and normal to the channel of interest (fueled and defueled).

Defueled channel experiments in ZED-2 in support of ACR-1000 ROP analysis

Energy Technology Data Exchange (ETDEWEB)

LaFontaine, M.W.R.; Zeller, M.B.; McPhee, G.P. [Atomic Energy of Canada Limited (Canada)

2011-07-01

Defueled channel experiments were performed in ZED-2 to help resolve discrepancies between calculated flux detector response during refueling in ACR-1000 according the reactor codes RFSP and MCNP. The data produced from these experiments was later used in a separate Regional-Over-Power (ROP) analysis to verify MCNP and RFSP neutron response predictions during refueling. These experiments provided information on thermal flux distributions interior and exterior to a fueled and defueled channel; and on epithermal absolute flux distributions exterior to the same channel. Critical height and moderator temperature data for fueled and defueled channel conditions were also measured. In addition, standard platinum-clad Inconel Self-Powered Detector (SPD) performance data was obtained. The following reactor physics and SPD parameters were measured in these experiments: C Radial flux distribution inside the channel of interest (fueled and defueled), C Radial flux distribution outside the channel of interest (fueled and defueled), C Epithermal radial flux distribution outside the channel of interest (fueled and defueled), and C SPD response parallel to and normal to the channel of interest (fueled and defueled).

Channel coincidence counter: version 1

International Nuclear Information System (INIS)

Krick, M.S.; Menlove, H.O.

1980-06-01

A thermal neutron coincidence counter has been designed for the assay of fast critical assembly fuel drawers and plutonium-bearing fuel rods. The principal feature of the detector is a 7-cm by 7-cm by 97-cm detector channel, which provides a uniform neutron detection efficiency of 16% along the central 40 cm of the channel. The electronics system is identical to that used for the High-Level Neutron Coincidence Counter

The fuel and channel thermal/mechanical behaviour code FACTAR 2.0 (LOCA)

International Nuclear Information System (INIS)

Westbye, C.J.; Mackinnon, J.C.; Gu, B.W.

1996-01-01

The computer code FACTAR 2.0 (LOCA) models the thermal and mechanical response of components within a single CANDU fuel channel under loss-of-coolant accident conditions. This code version is the successor to the FACTAR 1.x code series, and features many modelling enhancements over its predecessor. In particular, the thermal hydraulic treatment has been extended to model reverse and bi-directional coolant flow, and the axial variation in coolant flow rate. Thermal radiation is calculated by a detailed surface-to-surface model, and the ability to represent a greater range of geometries (including experimental configurations employed in code validation) has been implemented. Details of these new code treatments are described in this paper. (author)

Numerical predictions of a PEM fuel cell performance enhancement by a rectangular cylinder installed transversely in the flow channel

International Nuclear Information System (INIS)

Perng, Shiang-Wuu; Wu, Horng-Wen; Jue, Tswen-Chyuan; Cheng, Kuo-Chih

2009-01-01

This paper numerically investigates the installation of the transverse rectangular cylinder along the gas diffusion layer (GDL) in the flow channel for the cell performance enhancement of a proton exchange membrane fuel cell (PEMFC). The effects of the blockage at various gap sizes and the width of the cylinder on the cell performance enhancement have been studied with changing the gap ratios λ = 0.05-0.3, for the same cylinder) and the width-to-height ratios (WR = 0.66-1.66, for the same cylinder height and gap ratio). The results show that the transverse installation of a rectangular cylinder in the fuel flow channel effectively enhances the cell performance of a PEMFC. In addition, the influence of the width of the cylinder on the cell performance is obvious, and the best cell performance enhancement occurs at the gap ratio 0.2 among the gap ratios of 0.05, 0.1, 0.2, and 0.3.

Manufacture of fuel and fuel channels and their performance in Indian PHWRs'

International Nuclear Information System (INIS)

Kalidas, R.

2005-01-01

Nuclear Fuel Complex (NFC) at Hyderabad is conglomeration of chemical, metallurgical and mechanical plants, processing uranium and zirconium in two separate streams and culminating in the fuel assembly plant. Apart from manufacturing fuel for Pressurised Heavy Water Reactors (PHWRs) and Boiling Water Reactors (BWRs), NFC is also engaged in the manufacture of reactor core structurals for these reactors. NFC has carried our several technological developments over the years and implemented them for the manufacture of fuel, calandria tubes and pressure tubes for PHWRs. Keeping in pace with the Nuclear Power Programme envisaged by the Department of Atomic Energy, NFC had augmented its production capacities in all these areas. The paper highlights several actions initiated in the areas of fuel design, fuel manufacturing, manufacturing of zirconium alloy core structurals, fuel clad tubes and components and their performance in Indian PHWRs. (author)

Conservatism in methodologies for moderator subcooling sufficiency for fuel channel integrity upon pressure tube and calandria tube contact

Energy Technology Data Exchange (ETDEWEB)

Sun, L., E-mail: LSun@nbpower.com [Point Lepreau Generating Station, Lepreau, NB, (Canada)

2015-07-01

During a postulated large LOCA event in CANDU reactors, the pressure tube may balloon to contact with its surrounding calandria tube to transfer heat to the moderator. To confirm the integrity of the fuel channel in this case, many experiments have been performed in the last three decades. Based on the extant database of the pressure tube/calandria tube (PT/CT) contact, an analytical methodology was developed by Canadian Nuclear Industry to determine the sufficiency of moderator subcooling for fuel channel integrity. At the same time a semi-empirical methodology with an idea of Equivalent Moderator Subcooling (EMS) was also developed to judge the sufficiency of the moderator. In this work, some discussions were made over the two methodologies on their conservatism and it is demonstrated that the analytical approach is over conservative comparing with the EMS methodology. By using the EMS methodology, it is demonstrated that applying glass-peened calandria tubes, the requirement to moderator subcooling can be reduced by 10{sup o}C from that for smooth calandria tubes. (author)

The effect of channel flow pattern on internal properties distribution of a proton exchange membrane fuel cell for cathode starvation conditions

International Nuclear Information System (INIS)

Ko, Dong Soo; Kang, Young Min; Yang, Jang Sik; Jeong, Ji Hwan; Choi, Gyung Min; Kim, Duck Jool

2010-01-01

The effect of channel flow pattern on the internal properties distribution of a proton exchange membrane fuel cell (PEMFC) for cathode starvation conditions in a unit cell was investigated through numerical studies and experiments. The polarization curves of a lab-scale mixed serpentine PEMFC were measured with increasing current loads for different cell temperatures (40, 50, and 60 .deg. C) at a relative humidity of 100%. To study the local temperature on the membrane, the water content in the MEA, and the gas velocity in terms of the channel type of the PEMFC with operating characteristics, numerical studies using the es-pemfc module of STAR-CD, which have been matched to the experimental data, were conducted in detail. The water content and velocity at the cathode channel bend of the mixed serpentine channel were relatively higher than those at the single and double channels. Conversely, the local temperature and mean temperature on the membrane of a single serpentine channel were the highest among all channels. These results can be used to design the PEMFC system, the channel flow field, and the cooling device

Compatibility analysis of DUPIC fuel(4) - thermal hydraulic analysis

Energy Technology Data Exchange (ETDEWEB)

Park, Jee Won; Chae, Kyung Myung; Choi, Hang Bok

2000-07-01

Thermal-hydraulic compatibility of the DUPIC fuel bundle in the CANDU reactor has been studied. The critical channel power, the critical power ratio, the channel exit quality and the channel flow are calculated for the DUPIC and the standard fuels by using the NUCIRC code. The physical models and associated parametric values for the NUCIRC analysis of the fuels are also presented. Based upon the slave channel analysis, the critical channel power and the critical power ratios have been found to be very similar for the two fuel types. The same dryout model is used in this study for the standard and the DUPIC fuel bundles. To assess the dryout characteristics of the DUPIC fuel bundle, the ASSERT-PV code has been used for the subchannel analysis. Based upon the results of the subchannel analysis, it is found that the dryout location and the power for the two fuel types are indeed very similar. This study shows that thermal performance of the DUPIC fuel is not significantly different from that of the standard fuel.

Temperature measurement of the reactor materials samples irradiated in the fuel channels of the RA reactor - Annex 16

International Nuclear Information System (INIS)

Nikolic, M.; Djalovic, M.

1964-01-01

Reactor materials as graphite, stainless steel, magnox, zirconium alloys, etc. were exposed to fast neutron flux inside the fuel elements specially adapted for this purpose. Samples in the form ampoules were placed in capsules inside the fuel channels and cooled by heavy water which cools the fuel elements. In order to monitor the samples temperature 42 thermocouples were placed in the samples. That was necessary for reactor safety reasons and for further interpretation of measured results. Temperature monitoring was done continuously by multichannel milivoltmeters. This paper describes the technique of introducing the thermocouples, compensation instruments, control of the cold ends and adaptation of the instruments for precision (0.5%) temperature measurement in the range 30 deg - 130 deg C; 30 deg - 280 deg C and 30 deg - 80 deg C [sr

Nuclear fuel assembly

International Nuclear Information System (INIS)

Betten, P.R.

1976-01-01

Under the invention the fuel assembly is particularly suitable for liquid metal cooled fast neutron breeder reactors. Hence, according to the invention a fuel assembly cladding includes inward corrugations with respect to the remainder of the cladding according to a recurring pattern determined by the pitch of the metal wire helically wound round the fuel rods of the assembly. The parts of the cladding pressed inwards correspond to the areas in which the wire encircling the peripheral fuel rods is generally located apart from the cladding, thereby reducing the play between the cladding and the peripheral fuel rods situated in these areas. The reduction in the play in turn improves the coolant flow in the internal secondary channels of the fuel assembly to the detriment of the flow in the peripheral secondary channels and thereby establishes a better coolant fluid temperature profile [fr

Improved Direct Methanol Fuel Cell Stack

Science.gov (United States)

Wilson, Mahlon S.; Ramsey, John C.

2005-03-08

A stack of direct methanol fuel cells exhibiting a circular footprint. A cathode and anode manifold, tie-bolt penetrations and tie-bolts are located within the circular footprint. Each fuel cell uses two graphite-based plates. One plate includes a cathode active area that is defined by serpentine channels connecting the inlet and outlet cathode manifold. The other plate includes an anode active area defined by serpentine channels connecting the inlet and outlet of the anode manifold, where the serpentine channels of the anode are orthogonal to the serpentine channels of the cathode. Located between the two plates is the fuel cell active region.

Simplified CFD model of coolant channels typical of a plate-type fuel element: an exhaustive verification of the simulations

Energy Technology Data Exchange (ETDEWEB)

Mantecón, Javier González; Mattar Neto, Miguel, E-mail: javier.mantecon@ipen.br, E-mail: mmattar@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

2017-07-01

The use of parallel plate-type fuel assemblies is common in nuclear research reactors. One of the main problems of this fuel element configuration is the hydraulic instability of the plates caused by the high flow velocities. The current work is focused on the hydrodynamic characterization of coolant channels typical of a flat-plate fuel element, using a numerical model developed with the commercial code ANSYS CFX. Numerical results are compared to accurate analytical solutions, considering two turbulence models and three different fluid meshes. For this study, the results demonstrated that the most suitable turbulence model is the k-ε model. The discretization error is estimated using the Grid Convergence Index method. Despite its simplicity, this model generates precise flow predictions. (author)

Simplified CFD model of coolant channels typical of a plate-type fuel element: an exhaustive verification of the simulations

International Nuclear Information System (INIS)

Mantecón, Javier González; Mattar Neto, Miguel

2017-01-01

The use of parallel plate-type fuel assemblies is common in nuclear research reactors. One of the main problems of this fuel element configuration is the hydraulic instability of the plates caused by the high flow velocities. The current work is focused on the hydrodynamic characterization of coolant channels typical of a flat-plate fuel element, using a numerical model developed with the commercial code ANSYS CFX. Numerical results are compared to accurate analytical solutions, considering two turbulence models and three different fluid meshes. For this study, the results demonstrated that the most suitable turbulence model is the k-ε model. The discretization error is estimated using the Grid Convergence Index method. Despite its simplicity, this model generates precise flow predictions. (author)

Fuel assembly

International Nuclear Information System (INIS)

Ueda, Makoto.

1991-01-01

In a fuel assembly in which spectral shift type moderator guide members are arranged, the moderator guide member has a flow channel resistance member, that provides flow resistance against the moderators, in the upstream of a moderator flowing channel, by which the ratio of removing coolants is set greater at the upstream than downstream. With such a constitution, the void distribution increasing upward in the channel box except for the portion of the moderator guide member is moderated by the increase of the area of the void region that expands downward in the guide member. Accordingly, the axial power distribution is flattened throughout the operation cycle and excess distortion is eliminated to improve the fuel integrity. (T.M.)

Detection of failed fuel rods in shrouded BWR fuel assemblies

International Nuclear Information System (INIS)

Baero, G.; Boehm, W.; Goor, B.; Donnelly, T.

1988-01-01

A manipulator and an ultrasonic testing (UT) technique were developed to identify defective fuel rods in shrouded BWR fuel assemblies. The manipulator drives a UT probe axially through the bottom tie plate into the water channels between the fuel rods. The rotating UT probe locates defective fuel rods by ingressed water which attenuates the UT-signal. (author)

Applying alpha-channeling to mirror machines

Energy Technology Data Exchange (ETDEWEB)

Zhmoginov, A. I.; Fisch, N. J. [Department of Astrophysical Sciences, Princeton University, Princeton, New Jersey 08544 (United States)

2012-05-15

The {alpha}-channeling effect entails the use of radio-frequency waves to expel and cool high-energetic {alpha} particles born in a fusion reactor; the device reactivity can then be increased even further by redirecting the extracted energy to fuel ions. Originally proposed for tokamaks, this technique has also been shown to benefit open-ended fusion devices. Here, the fundamental theory and practical aspects of {alpha} channeling in mirror machines are reviewed, including the influence of magnetic field inhomogeneity and the effect of a finite wave region on the {alpha}-channeling mechanism. For practical implementation of the {alpha}-channeling effect in mirror geometry, suitable contained weakly damped modes are identified. In addition, the parameter space of candidate waves for implementing the {alpha}-channeling effect can be significantly extended through the introduction of a suitable minority ion species that has the catalytic effect of moderating the transfer of power from the {alpha}-channeling wave to the fuel ions.

On slip correlations used in the determination of the void distribution of BWR fuel channels

International Nuclear Information System (INIS)

Oelgaard, P.L.

1995-01-01

When performing calculations on boiling water reactors it is necessary to establish a relation between the average velocity of the steam, Vg, and that of the water, Vf, in the fuel channels. This is usually done through establishing an expression for the slip ratio S=V g /V f . In the literature a number of such slip ratio correlations -based on measurements - has been presented. A comparison between some of these correlations has been performed in this paper. While the correlations have some general trends in common, the numerical values of S obtained with the correlations may vary significantly. Further, in spite of the fact that various flow regimes exist in a boiling channel none of the correlations considered take the change of flow regime into account. This raises the question: How reliable is the use of slip correlation? (orig.) (9 refs., 5 figs.)

Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

OpenAIRE

Itamar Iliuk; José Manoel Balthazar; Ângelo Marcelo Tusset; José Roberto Castilho Piqueira

2016-01-01

Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermal-hydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was prop...

Perspective channel-type reactor with enhanced safety

International Nuclear Information System (INIS)

Adamov, E.O.; Grozdov, I.I.; Kuznetsov, S.P.; Petrov, A.A.; Rozhdestvensky, M.I.; Cherkashov, Yu.M.

1994-01-01

Following the search for new design solutions to develop within the framework of channel trends the reactor with enhanced safety the Research and Development Institute of Power Engineering has developed the design of the multiloop boiling water reactor (MKER). The MKER enhanced safety is attained when involving the inherent safety features, passive safety systems as well as the accident consequences confinement devices. The design realizes several advantages which are typical of the channel-type reactors, namely: The design desintegration simplifying the manufacture, control, equipment delivery and decreasing, versus the pressure vessel reactors, the accident effect if it proceeds in an explosive manner; small operating reactivity margin and fuel burnup increased due to continuous refuelling; fuel cycle flexibility allowing comparatively easily to adopt the reactor to the conjuncture of the country fuel balance; multiloop circuit of the main coolant which reduces the degree and effect of the accidents connected with the equipment and pipings rupture; monitoring of the channels and fuel assemblies leak-tightness. (orig.)

Advanced CANDU reactors fuel analysis through optimal fuel management at approach to refuelling equilibrium

International Nuclear Information System (INIS)

Tingle, C.P.; Bonin, H.W.

1999-01-01

The analysis of alternate CANDU fuels along with natural uranium-based fuel was carried out from the view point of optimal in-core fuel management at approach to refuelling equilibrium. The alternate fuels considered in the present work include thorium containing oxide mixtures (MOX), plutonium-based MOX, and Pressurised Water Reactor (PWR) spent fuel recycled in CANDU reactors (Direct Use of spent PWR fuel in CANDU (DUPIC)); these are compared with the usual natural UO 2 fuel. The focus of the study is on the 'Approach to Refuelling Equilibrium' period which immediately follows the initial commissioning of the reactor. The in-core fuel management problem for this period is treated as an optimization problem in which the objective function is the refuelling frequency to be minimized by adjusting the following decision variables: the channel to be refuelled next, the time of the refuelling and the number of fresh fuel bundles to be inserted in the channel. Several constraints are also included in the optimisation problem which is solved using Perturbation Theory. Both the present 37-rod CANDU fuel bundle and the proposed CANFLEX bundle designs are part of this study. The results include the time to reach refuelling equilibrium from initial start-up of the reactor, the average discharge burnup, the average refuelling frequency and the average channel and bundle powers relative to natural UO 2 . The model was initially tested and the average discharge burnup for natural UO 2 came within 2% of the industry accepted 199 MWh/kgHE. For this type of fuel, the optimization exercise predicted the savings of 43 bundles per full power year. In addition to producing average discharge burnups and other parameters for the advanced fuels investigated, the optimisation model also evidenced some problem areas like high power densities for fuels such as the DUPIC. Perturbation Theory has proven itself to be an accurate and valuable optimization tool in predicting the time between

Means for supporting nuclear fuel

International Nuclear Information System (INIS)

Cocker, P.; Price, M.A.

1975-01-01

Reference is made to means for supporting nuclear fuel pins in a reactor coolant channel and the problems that arise in this connection. For reasons of nuclear reactivity and neutron economy 'parasitic' material in a reactor core must be kept to a minimum, whilst for heat transfer reasons the use of fuel pins of large cross-sectional areas should be avoided. Fuel pins tend to be long thin objects having a can of minimum thickness and typically a pin may have a length/diameter ratio of about 500/1 and for fast reactor fuel pins, the outside diameter may be about 0.2 inch. The long slender pins must also be spaced very close together. A fast reactor fuel assembly may involve 200 to 300 fuel pins, each a few tenths of an inch in diameter, supported end on to coolant flowing up a channel of about 22 square inches in total area. The pins have a heavy metal oxide filling and require support. Details are given of a suitable method of support. Such support also allows withdrawal of pins from a fuel channel without the risk of breach of the can, after irradiation. (U.K.)

Eddy current detection of spacers in the fuel channels of CANDU nuclear reactors

International Nuclear Information System (INIS)

Krause, T.W.; Schankula, J.; Sullivan, S.P.

2002-01-01

Garter Spring (GS) spacers in the fuel channels of CANDU nuclear reactors maintain separation between the hot pressure tube and surrounding moderator cooled calandria tube. Eddy current detection of the four GSs provides assurance that spacers are at or close to design positions and are performing their intended function of maintaining a non-zero gap between pressure tube and calandria tube. Pressure tube constrictions, resulting from relatively less diametral creep at end-of-fuel bundle locations, also produce large eddy current signals. Large constrictions, present in higher service pressure tubes, can produce signals that are 10 times larger than GS signals, reducing GS detectability to 30% in standard GS-detect probes. The introduction of field-focussing elements into the design of the standard GS detection eddy current probe has been used to recover the detectability of GS spacers by increasing the signal amplitude obtained from GSs relative to that from constrictions by a factor of 10. The work presented here compares laboratory, modelling and in-reactor measurements of GS and constriction signals obtained from the standard probe with that obtained from field-focussed eddy current probe designs. (author)

Large-break LOCA assessment for the highly advanced core design

International Nuclear Information System (INIS)

Doria, F.J.; Nath, V.I.; Hau, K.F.; Dam, R.F.; Vecchiarelli, J.

1997-01-01

Over the course of the years, a conceptual highly advanced core (HAC) reactor has been designed for Japan Electric Power Development Company Limited (EPDC). The HAC reactor, which is capable of generating 1326 MW of electrical power, consists of 640 CANDU-type fuel channels with each fuel channel containing twelve 61-element fuel bundles. As part of the conceptual design study, the performance of the HAC reactor during a large loss-of-coolant accident (LOCA) was assessed with the use of several computer codes. The SOPHT, CATHENA, ELOCA and ELESTRES computer codes were used to predict the thermalhydraulic behaviour of the circuit, thermalhydraulic behaviour of a single high-power channel, thermal-mechanical behaviour of the outer fuel elements contained in the high-powered channel and the steady-state fuel-element conditions respectively. The LOCAs that were analyzed include 100% reactor outlet header (ROH) break, and a survey of reactor inlet header (RIH) breaks ranging from 5% to 25%. The conceptual feasibility of the HAC design was evaluated against two criteria; namely, maximum sheath temperature less than 1200 deg C and AECL's 5% sheath straining criterion to assess failure by excessive straining. For the cases analyzed, the analysis predicted a maximum sheath temperature of 820 deg C and a maximum sheath strain of 1.5% (the maximum pressure-tube temperature was 515 deg C). Although the maximum element-burnup of the HAC design is extended beyond the CANDU 6 burnup, the maximum linear power of HAC (40 kW/m) is significantly lower than the maximum linear power of a CANDU 6 reactor (60 kW/m). The reduced element-power level in conjunction with internal design modification for the HAC design has resulted m significantly lower internal gas pressures under steady-state conditions, as compared with the CANDU 6 design. During a LOCA, the low linear powers and zero-void reactivity associated with the HAC design has increased the safety margin. In addition, the cases

Feasibility investigations on a novel micro-manufacturing process for fabrication of fuel cell bipolar plates: Internal pressure-assisted embossing of micro-channels with in-die mechanical bonding

Energy Technology Data Exchange (ETDEWEB)

Koc, Muammer [NSF I/UCR Center for Precision Forming (CPF), Department of Mechanical Engineering, Virginia Commonwealth University (VCU), Richmond, VA (United States); Mahabunphachai, Sasawat [NSF I/UCR Center for Precision Forming (CPF), Department of Mechanical Engineering, Virginia Commonwealth University (VCU), Richmond, VA (United States); Department of Mechanical Engineering, University of Michigan, Ann Arbor, MI (United States)

2007-10-25

In this paper, we present the results of our studies on conceptual design and feasibility experiments towards development of a novel hybrid manufacturing process to fabricate fuel cell bipolar plates that consists of multi-array micro-channels on a large surface area. The premises of this hybrid micro-manufacturing process stem from the use of an internal pressure-assisted embossing process (cold or warm) combined with mechanical bonding of double bipolar plates in a single-die and single-step operation. Such combined use of hydraulic and mechanical forming forces and in-process bonding will (a) enable integrated forming of micro-channels on both surfaces (as anode and cathode flow fields) and at the middle (as cooling channels), (b) reduce the process steps, (c) reduce variation in dimensional tolerances and surface finish, (d) increase the product quality, (e) increase the performance of fuel cell by optimizing flow-field designs and ensuring consistent contact resistance, and (f) reduce the overall stack cost. This paper explains two experimental investigations that were performed to characterize and evaluate the feasibility of the conceptualized manufacturing process. The first investigation involved hydroforming of micro-channels using thin sheet metals of SS304 with a thickness of 51 {mu}m. The width of the channels ranged from 0.46 to 1.33 mm and the height range was between 0.15 and 0.98 mm. Our feasibility experiments resulted in that different aspect ratios of micro-channels could be fabricated using internal pressure in a controllable manner although there is a limit to very sharp channel shapes (i.e., high aspect ratios with narrow channels). The second investigation was on the feasibility of mechanical bonding of thin sheet metal blanks. The effects of different process and material variables on the bond quality were studied. Successful bonding of various metal blanks (Ni201, Al3003, and SS304) was obtained. The experimental results from both

Fuel assembly

International Nuclear Information System (INIS)

Ishibashi, Yoko; Aoyama, Motoo; Oyama, Jun-ichi.

1995-01-01

Burnable poison-incorporating fuel rods of a first group are disposed in a region in adjacent with a water rod having a large diameter (neutron moderator rod) disposed to the central portion of a fuel assembly. Burnable poison-incorporating fuel rods of a second group are disposed to a region other than peripheral zone in adjacent with a channel box and corners positioned at an inner zone, in adjacent with the channel box. The average concentration of burnable poisons of the burnable poison-incorporating fuel rods of the first group is made greater than that of the second group. With such a constitution, when the burnable poisons of the first group are burnt out, the burnable poisons of the second group are also burnt out at the same time. Accordingly, an amount of burnable poisons left unburnt at the final stage of the operation cycle is reduced, to improve the reactivity. This can improve the economical property. (I.N.)

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

Enhanced heat transfer with corrugated flow channel in anode side of direct methanol fuel cells

International Nuclear Information System (INIS)

Heidary, H.; Abbassi, A.; Kermani, M.J.

2013-01-01

Highlights: • Effect of corrugated flow channel on the heat exchange of DMFC is studied. • Corrugated boundary (except rectangular type) increase heat transfer up to 90%. • Average heat transfer in rectangular-corrugated boundary is less than straight one. • In Re > 60, wavy shape boundary has highest heat transfer. • In Re < 60, triangular shape boundary has highest heat transfer. - Abstract: In this paper, heat transfer and flow field analysis in anode side of direct methanol fuel cells (DMFCs) is numerically studied. To enhance the heat exchange between bottom cold wall and core flow, bottom wall of fluid delivery channel is considered as corrugated boundary instead of straight (flat) one. Four different shapes of corrugated boundary are recommended here: rectangular shape, trapezoidal shape, triangular shape and wavy (sinusoidal) shape. The top wall of the channel (catalyst layer boundary) is taken as hot boundary, because reaction occurs in catalyst layer and the bottom wall of the channel is considered as cold boundary due to coolant existence. The governing equations are numerically solved in the domain by the control volume approach based on the SIMPLE technique (1972). A wide spectrum of numerical studies is performed over a range of various shape boundaries, Reynolds number, triangle block number, and the triangle block amplitude. The performed parametric studies show that corrugated channel with trapezoidal, triangular and wavy shape enhances the heat exchange up to 90%. With these boundaries, cooling purpose of reacting flow in anode side of DMFCs would be better than straight one. Also, from the analogy between the heat and mass transfer problems, it is expected that the consumption of reacting species within the catalyst layer of DMFCs enhance. The present work provides helpful guidelines to the bipolar plate manufacturers of DMFCs to considerably enhance heat transfer and performance of the anode side of DMFC

Consideration of hot channel factors in design for providing operating margins on coolant channel outlet temperature

International Nuclear Information System (INIS)

Sharma, V.K.; Surendar, C.; Bapat, C.N.

1994-01-01

The Indian Pressurized Heavy Water Reactors (IPHWR) are horizontal pressure tube reactors using natural uranium oxide fuel in the form of short (495 mm) clusters. The fuel clusters in the Zr-Nb pressure tubes are cooled by high pressure, high temperature and subcooled circulating heavy water. Coolant flow distribution to individual channels is designed to match the power distribution so as to obtain uniform coolant outlet temperature. However, during operation, the coolant outlet temperature in individual channels deviate from their nominal value due to: tolerances in process design; effects of grid frequency on the pump speed; deviation in channel powers from the nominal values due to on-power fuelling and movement of reactivity devices, and so on. Thus an operating margin, between the highest permissible and nominal coolant outlet temperatures, is required taking into account various hot channel factors that contribute to higher coolant outlet temperatures. The paper discusses the methodology adopted to assess various hot channel factors which would provide optimum operating margins while ensuring sub-cooling. (author)

Effect of power variations across a fuel bundle and within a fuel element on fuel centerline temperature in PHWR bundles in uncrept and crept pressure tubes

International Nuclear Information System (INIS)

Onder, E.N.; Roubtsov, D.; Rao, Y.F.; Wilhelm, B.

2017-01-01

Highlights: • Pressure tube creep effect on fuel pin power and temperatures was investigated. • Noticeable effects were observed for 5.1% crept pressure tube. • Bundle eccentricity effect on power variations was insignificant for uncrept channels. • Difference of 112 °C was observed between top & bottom elements in 5.1% crept channel. • Not discernible fission gas release was expected with temperature difference of 112 °C. - Abstract: The neutron flux and fission power profiles through a fuel bundle and across a fuel element are important aspects of nuclear fuel analysis in multi-scale/multi-physics modelling of Pressurized Heavy Water Reactors (PHWRs) with advanced fuel bundles. Fuel channels in many existing PHWRs are horizontal. With ageing, pressure tubes creep and fuel bundles in these pressure tubes are eccentrically located, which results in an asymmetric coolant flow distribution between the top and bottom of the fuel bundles. The diametral change of the pressure tube due to creep is not constant along the fuel channel; it reaches a maximum in the vicinity of the maximum neutron flux location. The cross-sectional asymmetric positioning of fuel bundles in a crept pressure tube contributes to an asymmetric power distribution within a ring of fuel elements. Modern reactor physics lattice codes (such as WIMS-AECL) are capable of predicting the details of power distribution from basic principles. Thermalhydraulics subchannel codes (such as ASSERT-PV) use models to describe inhomogeneous power distribution within and across fuel elements (e.g., flux tilt model, different powers in different ring elements, or radial power profiles). In this work, physics and thermalhydraulics codes are applied to quantify the effect of eccentricity of a fuel bundle on power variations across it and within a fuel element, and ultimately on the fuel temperature distribution and fuel centerline temperature, which is one of the indicators of fuel performance under normal

Equations of macrotransport in reactor fuel assemblies

International Nuclear Information System (INIS)

Sorokin, A.P.; Zhukov, A.V.; Kornienko, Yu.N.; Ushakov, P.A.

1986-01-01

The rigorous statement of equations of macrotransport is obtained. These equations are bases for channel-by-channel methods of thermohydraulic calculations of reactor fuel assemblies within the scope of the model of discontinuous multiphase coolant flow (including chemical reactions); they also describe a wide range of problems on thermo-physical reactor fuel assembly justification. It has been carried out by smoothing equations of mass, momentum and enthalpy transfer in cross section of each phase of the elementary fuel assembly subchannel. The equation for cross section flows is obtaind by smoothing the equation of momentum transfer on the interphase. Interaction of phases on the channel boundary is described using the Stanton number. The conclusion is performed using the generalized equation of substance transfer. The statement of channel-by-channel method without the scope of homogeneous flow model is given

Method and device for cleaning fuel pins

International Nuclear Information System (INIS)

Matsumoto, Kaname; Oohigashi, Yoshiaki.

1985-01-01

Purpose: To remove clads or scales deposited on the outer surface of fuel pins in BWR type reactors. Method: A fuel assembly taken out of a reactor core is vertically contained without detaching a channel box in a scrubber tower disposed in a liquid tight manner within a fuel pool. Then, a specifically prepared slurry is caused to flow and uprise from the bottom of the scrubber tower into the channel box and then discharged from the top of the tower. The slurry is prepared by mixing pure water and granules (for example, as activated carbon, ion exchanger resin, iron and molecular sieve) of such a granular size as not causing clogging in the channel box of the fuel assembly and having a larger specific gravity than pure water. The slurry flown into the channel box scrubs the surface of fuel pins to scrape off clads or scales. Then, discharged slurry is sent to a hydraulic cyclone to separate the granules from the clads or scales. (Ikeda, J.)

Numerical prediction of turbulent heat transfer augmentation in an annular fuel channel with two-dimensional square ribs

International Nuclear Information System (INIS)

Takase, Kazuyuki

1996-01-01

The square-ribbed fuel rod for high temperature gas-cooled reactors was developed in order to enhance the turbulent heat transfer in comparison with the standard fuel rod. To evaluate the heat transfer performance of the square-ribbed fuel rod, the turbulent heat transfer coefficients in an annular fuel channel with repeated two-dimensional square ribs were analyzed numerically on a fully developed incompressible flow using the k - ε turbulence model and the two-dimensional axisymmetrical coordinate system. Numerical analyses were carried out for a range of Reynolds numbers from 3000 to 20000 and ratios of square-rib pitch to height of 10, 20 and 40, respectively. The predicted values of the heat transfer coefficients agreed within an error of 10% for the square-rib pitch to height ratio of 10, 20% for 20 and 25% for 40, respectively, with the heat transfer empirical correlations obtained from the experimental data. It was concluded by the present study that the effect of the heat transfer augmentation by square ribs could be predicted sufficiently by the present numerical simulations and also a part of its mechanism could be explained by means of the change in the turbulence kinematic energy distribution along the flow direction. (author)

Numerical prediction of augmented turbulent heat transfer in an annular fuel channel with repeated two-dimensional square ribs

International Nuclear Information System (INIS)

Takase, K.

1996-01-01

The square-ribbed fuel rod for high temperature gas-cooled reactors was designed and developed so as to enhance the turbulent heat transfer in comparison with the previous standard fuel rod. The turbulent heat transfer characteristics in an annular fuel channel with repeated two-dimensional square ribs were analysed numerically on a fully developed incompressible flow using the k-ε turbulence model and the two-dimensional axisymmetrical coordinate system. Numerical analyses were carried out under the conditions of Reynolds numbers from 3000 to 20000 and ratios of square-rib pitch to height of 10, 20 and 40 respectively. The predictions of the heat transfer coefficients agreed well within an error of 10% for the square-rib pitch to height ratio of 10, 20% for 20 and 25% for 40 respectively, with the heat transfer empirical correlations obtained from the experimental data due to the simulated square-ribbed fuel rods. Therefore it was found that the effect of heat transfer augmentation due to the square ribs could be predicted by the present numerical simulations and the mechanism could be explained by the change in the turbulence kinematic energy distribution along the flow direction. (orig.)

BWR fuel assembly with improved spacer and fuel bundle design for enhanced thermal-hydraulic performance

International Nuclear Information System (INIS)

Mildrum, C.M.; Taleyarkhan, R.P.

1987-01-01

In a fuel assembly having a bundle of elongated fuel rods disposed in side-by-side relationship so as to form an array of spaced fuel rods, an outer tubular flow channel surrounding the fuel rods so as to direct flow of coolant/moderator fluid along the fuel rods, a hollow water cross extending centrally through and interconnected with the outer flow channel so as to divide the channel into separate compartments and the bundle of fuelrods into a plurality of mini-bundles thereof being disposed in the compartments, and spacers axially displaced along the fuel rods in each of the mini-bundles thereof. Each spacer is composed of inner and outer means which together define spacer cells at corner, side and interior locations of the spacer and have respective protrusions formed thereon which extend into cells so as to maintain the fuel rods received through the spacer cells in laterally spaced relationships. The improvement is described which comprises: (a) a generally uniform poison coating within at least a majority of the fuel rods; (b) a predetermined pattern of fuel enrichment with respect to the fuel rods of each mini-bundle thereof which together with the uniform poison coating within the fuel rods ensures that the packing powers of the fuel rods in the corner and side cells of the spacers are less than the peaking power of a leading one of the fuel rods in the interior cells of the spacers; and (c) each of the fuel rods being received through the cells of each spacer having a diametric size smaller than that of each of the fuel rods received through the side and interior cells of each spacer, the diametric sizes of each of the fuel rods received through the side and interior cells of each spacer being generally equal

DETAILS OF OPERATIONS PERFORMED BY THE REMOTE CONTROL ROBOT (CONCEPT TO THE HORIZONTAL FUEL CHANNEL DURING DECOMMISSIONING PHASE OF NUCLEAR REACTOR CALANDRIA STRUCTURE. PART II: INSIDE OPERATIONS

Directory of Open Access Journals (Sweden)

Constantin POPESCU

2017-05-01

Full Text Available The authors contribution to this paper is to present a concept solution of a remote control robot (RCR used for decommissioning of the horizontal fuel channels pressure tube in the CANDU nuclear reactor. In this paper the authors highlight few details of geometry, operations, constraints by kinematics and dynamics of the robot movement inside of the reactor fuel channel. Inside operations performed has as the main steps of dismantling process the followings: unblock and extract the channel closure plug (from End Fitting - EF, unblock and extract the channel shield plug (from Lattice Tube - LT, cut the ends of the pressure tube, extract the pressure tube and cut it in small parts, sorting and storage extracted items in the safe robot container. All steps are performed in automatic mode. The remote control robot (RCR represents a safety system controlled by sensors and has the capability to analyze any error registered and decide next activities or abort the inside decommissioning procedure in case of any risk rise in order to ensure the environmental and workers protection.

Coupled 3D neutron kinetics and thermalhydraulic characteristics of the Canadian supercritical water reactor

Energy Technology Data Exchange (ETDEWEB)

Hummel, David William, E-mail: hummeld@mcmaster.ca; Novog, David Raymond

2016-03-15

Highlights: • A coupled spatial kinetics and thermalhydraulics model of the PT-SCWR was created. • Positive power excursions were demonstrated during accident-like transients. • The reactor will inherently self-shutdown in such transients with some delay. • A fast-acting shutdown system would limit the consequences of the power pulse. - Abstract: The Canadian Supercritical Water-cooled Reactor concept, as an evolution of the CANada Deuterium Uranium (CANDU) reactor, includes both pressure tubes and a low temperature heavy water moderator. The current Pressure Tube type SCWR (PT-SCWR) concept features 64-element fuel assemblies placed within High Efficiency Re-entrant Channels (HERCs) that connect to core inlet and outlet plena. Among current SCWR concepts the PT-SCWR is unique in that the HERC separates multiple coolant and moderator regions, giving rise to coupled neutronic-thermalhydraulic feedbacks beyond those present in CANDU or contemporary Light Water Reactors. The objective of this work was thus to model the coupled neutronic-thermal hydraulic properties of the PT-SCWR to establish the impact of these multiple regions on the core's transient behavior. To that end, the features of the PT-SCWR were first modeled with the neutron transport code DRAGON to create a database of homogenized and condensed cross-sections and thermalhydraulic feedback coefficients. These were used as input to a core-level neutron diffusion model created with the code DONJON. The behavior of the primary heat transport system was modeled with the thermalhydraulic system code CATHENA. A procedure was developed to couple the outputs of DONJON and CATHENA, facilitating three-dimensional spatial neutron kinetics and coupled thermalhydraulic analysis of the PT-SCWR core. Several postulated transients were initiated within the coupled model by changing the core inlet and outlet boundary conditions. Decreasing coolant density around the fuel was demonstrated to produce positive

Manufacture of fuel and fuel channels and their performance in Indian PHWRS - an overview

International Nuclear Information System (INIS)

Kalidas, R.

2005-01-01

Nuclear Fuel Complex (NFC) at Hyderabad is a conglomeration of chemical, metallurgical and mechanical plants, processing uranium and zirconium in two separate streams and culminating in the fuel assembly plant. Apart from manufacturing fuel for Pressurised Heavy Water Reactors (PHWRs) and Boiling Water Reactors (BWRs), NFC is also engaged in the manufacture of reactor core structurals for these reactors. NFC has carried out several technological developments over the years and implemented them for the manufacture of fuel, calandria tubes and pressure tubes for PHWRs. Keeping in pace with the Nuclear Power Programme envisaged by the Department of Atomic Energy, NFC had augmented its production capacities in all these areas. The paper highlights several actions initiated in the areas of fuel design, fuel manufacturing, manufacturing of zirconium alloy core structurals, fuel clad tubes and components and their performance in Indian PHWRs. (author)

Evaluation of a thermal SCWR core with sub-channel analysis

International Nuclear Information System (INIS)

Liu Xiaojing; Cheng Xu

2008-01-01

A previous study shows that the two-row fuel assembly has much more favorable neutron-physical and thermal-hydraulic behaviour than the existing one-row fuel assemblies. With this new developed two-row fuel assembly, a thermal SCWR core design is proposed Assessment of this design is carried out in this paper. The performance of this new core design is investigated with 3-D coupled thermal-hydraulic/neutronic calculations. During the coupling procedure, the thermal-hydraulic behaviour is analyzed using a single-channel code and the neutron-physical performance is computed with a 3-D reactor physical code. This paper presents the main results achieved so far related to the distribution of some neutronic and thermal-hydraulic parameters. Since the power distribution in some fuel assemblies is extremely uneven, sub-channel analysis is applied to the hottest and most non-uniform assembly in the core. The sub-channel analysis is performed with the power and thermal hydraulic parameters from the coupling results. It provides the hot channel factor and the maximal cladding surface temperature more precisely. The power and mass flux distribution in these assemblies are illustrated in detail for the demonstration purpose. The difference of the results evaluated with two different methods, i.e. sub-channel analysis and single-channel analysis, shows the importance of applying sub-channel analysis. A sensitivity analysis of some important parameters is also carried out. (author)

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

Low hydrostatic head electrolyte addition to fuel cell stacks

International Nuclear Information System (INIS)

Kothmann, R.E.

1983-01-01

A fuel cell and system for supply electrolyte, as well as fuel and an oxidant to a fuel cell stack having at least two fuel cells, each of the cells having a pair of spaced electrodes and a matrix sandwiched therebetween, fuel and oxidant paths associated with a bipolar plate separating each pair of adjacent fuel cells and an electrolyte fill path for adding electrolyte to the cells and wetting said matrices. Electrolyte is flowed through the fuel cell stack in a back and forth fashion in a path in each cell substantially parallel to one face of opposite faces of the bipolar plate exposed to one of the electrodes and the matrices to produce an overall head uniformly between cells due to frictional pressure drop in the path for each cell free of a large hydrostatic head to thereby avoid flooding of the electrodes. The bipolar plate is provided with channels forming paths for the flow of the fuel and oxidant on opposite faces thereof, and the fuel and the oxidant are flowed along a first side of the bipolar plate and a second side of the bipolar plate through channels formed into the opposite faces of the bipolar plate, the fuel flowing through channels formed into one of the opposite faces and the oxidant flowing through channels formed into the other of the opposite faces

Measurement of the Velocity and Pressure Drop in a Tubular Type Fuel

International Nuclear Information System (INIS)

Jonghark Park; Heetaek Chae; Cheol Park; Heonil Kim

2006-01-01

We have developed a tubular type fuel assembly design as one of candidates for fuel to be used in the Advanced HANARO Reactor (AHR). The tubular type fuel has several merits over a rod type fuel with respect to the thermal-hydraulic and structural safety; the larger ratio of surface area to volume makes the surface temperature of a fuel element become lower, and curved plate is stronger against longitudinal bending and vibration. In the other side, a disadvantage is expected such that the flow velocity can be distributed unevenly channel by channel because the flow channels are isolated from each other in a tubular type fuel assembly. In addition to the design development, we also investigated the flow characteristics of the tubular fuel experimentally. To examine the flow velocity distribution and pressure drop, we made an experiment facility and a mockup of the tubular fuel assembly. The fuel assembly consists of 6 concentric fuel tubes so that 7 layers are made between fuel tubes. Since each layer is divided into three sections by stiffeners, 21 isolated flow channels are made in total. We employed pitot-tubes to measure the coolant velocity in each channel. The maximum velocity was measured as large as about 28% of the average velocity. It was observed in the innermost channel contrarily to the expectation from the hydraulic diameter. A change in the total flow rate did not affect the flow distribution. Meanwhile, the pressure drop was measured as about 70% of the drop in the rod type fuel assembly in use in HANARO. (authors)

Method of removing crud deposited on fuel element clusters

International Nuclear Information System (INIS)

Yokota, Tokunobu; Yashima, Akira; Tajima, Jun-ichiro.

1982-01-01

Purpose: To enable easy elimination of claddings deposited on the surface of fuel element. Method: An operator manipulates a pole from above a platform, engages the longitudinal flange of the cover to the opening at the upper end of a channel box and starts up a suction pump. The suction amount of the pump is set such that water flow becomes within the channel box at greater flow rate than the operational flow rate in the channel box of the fuel element clusters during reactor operation. This enables to remove crud deposited on the surface of individual fuel elements with ease and rapidly without detaching the channel box. (Moriyama, K.)

Assessment of Fuel Analysis Methodology and Fission Product Release for 37-Element Fuel by Using the Latest IST Codes during Stagnation Feeder Break in CANDU

International Nuclear Information System (INIS)

Park, Joo Hwan; Jung, Jong Yeob

2009-09-01

Feeder break accident is regarded as one of the design basis accident in CANDU reactor which results in a fuel failure. For a particular range of inlet feeder break sizes, the flow in the channel is reduced sufficiently that the fuel and fuel channel integrity can be significantly affected to have damage in the affected channel, while the remainder of the core remains adequately cooled. The flow in the downstream channel can be more or less stagnated due to a balance between pressure at the break on the upstream side and the reverse driving pressure between the break and the downstream end. In the extreme, this can lead to rapid fuel heatup and fuel damage and failure of the fuel channel similar to that associated with a severe channel flow blockage. Such an inlet feeder break scenario is called a stagnation break. In this report, the fuel analysis methodology and the assessment results of fission product inventory and release during the stagnation feeder break are described for conservatively assumed limiting channel. The accident was assumed to be occurred in the refurbished Wolsong unit 1 and the latest safety codes were used in the analysis. Fission product inventories during the steady state were calculated by using ELESTRES-IST 1.2 code. The whole analysis process was carried out by a script file which was programmed by Perl language. The perl script file was programmed to make all ELESTRES input files for each bundle and each ring based on the given power-burnup history and thermal-hydraulic conditions of the limiting channel and to perform the fuel analysis automatically. The fission product release during the transient period of stagnation feeder break was evaluated by applying Gehl model. The amounts of each isotope's release are conservatively evaluated for additional 2 seconds after channel failure. The calculated fission product releases are provided to the following dose assessment as a source term

Sensitivity Analysis of Gap Conductance for Heat Split in an Annular Fuel Rod

International Nuclear Information System (INIS)

Chun, Kun Ho; Chun, Tae Hyun; In, Wang Kee; Song, Keun Woo

2006-01-01

To increase of the core power density in the current PWR cores, an annular fuel rod was proposed by MIT. This annular fuel rod has two coolant channels and two cladding-pellet gaps unlike the current solid fuel rod. It's important to predict the heat split reasonably because it affects coolant enthalpy rise in each channel and Departure from Nuclear Boiling Ratio (DNBR) in each channel. Conversely, coolant conditions affect fuel temperature and heat split. In particular if the heat rate leans to either inner or outer channel, it is out of a thermal equilibrium. To control a thermal imbalance, placing another gap in the pellet is introduced. The heat flow distribution between internal and external channels as well as fuel and cladding temperature profiles is calculated with and without the fuel gap between the inner and outer pellets

BWR fuel performance

International Nuclear Information System (INIS)

Baily, W.E.; Armijo, J.S.; Jacobson, J.; Proebstle, R.A.

1979-01-01

The General Electric experience base on BWR fuel includes over 29,000 fuel assemblies which contain 1,600,000 fuel rods. Over the last five years, design, process and operating changes have been introduced which have had major effects in improving fuel performance. Monitoring this fuel performance in BWRs has been accomplished through cooperative programs between GE and utilities. Activities such as plant fission product monitoring, fuel sipping and fuel and channel surveillance programs have jointly contributed to the value of this extensive experience base. The systematic evaluation of this data has established well-defined fuel performance trends which provide the assurance and confidence in fuel reliability that only actual operating experience can provide

Standard model of WWER-440 fuel rod for Transuranus and its application for RELAP5 hot channel validation

International Nuclear Information System (INIS)

Hatala, B.; Cvan, M.

2001-01-01

Within the PECO European Commission project of 'Extension of the validation matrix of the TRANSURANUS code' is developed a generic model of WWER-440 fuel rod. The model is intended to be applied for both realistic and licensing, conservative analysis. For such an application the TRANSURANUS code would be complementary tool to generally used system codes, e.g. RELAP5, providing realistic, more detailed insight into processes and safety criteria, relevant to the fuel rod. The paper presents general description of the model for TRANSURANUS code, brief discussion of approaches used in TRANSURANUS and RELAP5 code safety analysis, accompanied with information about RELAP5 model (whole scope unit model, used for licensing analysis). The existing model for RELAP5 code for WWER-440/V-213 Bohunice V2 unit is checked and modified in hot channel part to allow transparent comparison with the TRANSURANUS code. The results from comparison calculations of the both codes are presented for fresh fuel and quasi steady state scenario and are in good agreement, almost identical. These results might be used as a basis for transient analysis

Fuel cell membrane humidification

Science.gov (United States)

Wilson, Mahlon S.

1999-01-01

A polymer electrolyte membrane fuel cell assembly has an anode side and a cathode side separated by the membrane and generating electrical current by electrochemical reactions between a fuel gas and an oxidant. The anode side comprises a hydrophobic gas diffusion backing contacting one side of the membrane and having hydrophilic areas therein for providing liquid water directly to the one side of the membrane through the hydrophilic areas of the gas diffusion backing. In a preferred embodiment, the hydrophilic areas of the gas diffusion backing are formed by sewing a hydrophilic thread through the backing. Liquid water is distributed over the gas diffusion backing in distribution channels that are separate from the fuel distribution channels.

BWR fuel assembly bottom nozzle with one-way coolant flow valve

International Nuclear Information System (INIS)

Taleyarkhan, R.P.

1987-01-01

In a nuclear reactor having a flow of coolant/moderator fluid therein, at least one fuel assembly installed in the fluid flow, the fuel assembly is described comprising in combination: a bundle of elongated fuel rods disposed in side-by-side relationship so as to form an array of spaced fuel rods; an outer tubular flow channel surrounding the fuel rods so as to direct the flow of coolant/moderator fluid along the fuel rods; bottom and top nozzles mounted at opposite ends of the flow channel and having an inlet and outlet respectively for allowing entry and exit of the flow of coolant/moderator fluid into and from the flow channel and along the fuel rods therein; and a coolant flow direction control device operatively disposed in the bottom nozzle so as to open the inlet thereof to the flow of coolant/moderator fluid in an inflow direction into the flow channel through the bottom nozzle inlet but close the inlet to the flow of coolant/moderator fluid from the flow channel through the bottom nozzle inlet upon reversal of coolant/moderator fluid flow from the inflow direction

Thermal-Hydraulic Analysis of Coolant Flow Decrease in Fuel Channels of Smolensk-3 RBMK during GDH Blockage Event

International Nuclear Information System (INIS)

Costa, A. L.; Cherubini, M.; D'Auria, F.; Giannotti, W.; Moskalev, A.

2007-01-01

One of the transients that have received considerable attention in the safety evaluation of RBMK reactors is the partial break of a group distribution header (GDH). The coolant flow rate blockage in one GDH might lead to excessive heat-up of the pressure tubes and can result in multiple fuel channels (FC) ruptures. In this work, the GDH flow blockage transient has been studied considering the Smolensk-3 RBMK NPP (nuclear power plant). In the RBMK, each GDH distributes coolant to 40-43 FC. To investigate the behavior of each FC belonging to the damaged GDH and to have a more realistic trend, one (affected) GDH has been schematised with its forty-two FC, one by one. The calculations were performed using the 0-D NK (neutron kinetic) model of the RELAP5-3.3 stand-alone code. The results show that, during the event, the mass flow rate is disturbed differently according to the power distribution established for each FC in the schematization. The start time of the oscillations in mass flow rate depends strongly on the attributed power to each FC. It was also observed that, during the event, the fuel channels at higher thermal power values tend to undergo first cladding rupture leaving the reactor to scram and safeguarding all the other FCs connected to the affected GDH.

Fuel assembly for a nuclear reactor

International Nuclear Information System (INIS)

Ferrari, H.M.; Miller, D.L.; Tong, L.S.

1975-01-01

A description is given of a fuel assembly including multiple open channel grids for holding fuel rods and control rod guide thimbles in predetermined fixed relationship with each other. Metallic straps are interwoven to form a grid or egg crate configuration having openings which receive the fuel rods and guide thimbles. To properly support and cool the fuel rods near the grid-fuel rod interface, springs and dimples on the grid straps project into each opening, the dimples being oriented in a direction to permit flow of coolant upwardly therethrough. To minimize turbulence in coolant flow, the leading edge of each grid strap is provided with cutout sections which form scallops effective in channeling coolant in a uniform flow path through the network of grid openings

Water Transport and Removal in PEMFC Gas Flow Channel with Various Water Droplet Locations and Channel Surface Wettability

Directory of Open Access Journals (Sweden)

Yanzhou Qin

2018-04-01

Full Text Available Water transport and removal in the proton exchange membrane fuel cell (PEMFC is critically important to fuel cell performance, stability, and durability. Water emerging locations on the membrane-electrode assembly (MEA surface and the channel surface wettability significantly influence the water transport and removal in PEMFC. In most simulations of water transport and removal in the PEMFC flow channel, liquid water is usually introduced at the center of the MEA surface, which is fortuitous, since water droplet can emerge randomly on the MEA surface in PEMFC. In addition, the commonly used no-slip wall boundary condition greatly confines the water sliding features on hydrophobic MEA/channel surfaces, degrading the simulation accuracy. In this study, water droplet is introduced with various locations along the channel width direction on the MEA surface, and water transport and removal is investigated numerically using an improved model incorporating the sliding flow property by using the shear wall boundary condition. It is found that the water droplet can be driven to the channel sidewall by aerodynamics when the initial water location deviates from the MEA center to a certain amount, forming the water corner flow in the flow channel. The channel surface wettability on the water transport is also studied and is shown to have a significant impact on the water corner flow in the flow channel.

A Comparison of Materials Issues for Cermet and Graphite-Based NTP Fuels

Science.gov (United States)

Stewart, Mark E.; Schnitzler, Bruce G.

2013-01-01

This paper compares material issues for cermet and graphite fuel elements. In particular, two issues in NTP fuel element performance are considered here: ductile to brittle transition in relation to crack propagation, and orificing individual coolant channels in fuel elements. Their relevance to fuel element performance is supported by considering material properties, experimental data, and results from multidisciplinary fluid/thermal/structural simulations. Ductile to brittle transition results in a fuel element region prone to brittle fracture under stress, while outside this region, stresses lead to deformation and resilience under stress. Poor coolant distribution between fuel element channels can increase stresses in certain channels. NERVA fuel element experimental results are consistent with this interpretation. An understanding of these mechanisms will help interpret fuel element testing results.

Reliability testing of failed fuel location system

International Nuclear Information System (INIS)

Vieru, G.

1996-01-01

This paper presents the experimental reliability tests performed in order to prove the reliability parameters for Failed Fuel Location System (FFLS), equipment used to detect in which channel of a particular heat transport loop a fuel failure is located, and to find in which channel what particular bundle pair is failed. To do so, D20 samples from each reactor channel are sequentially monitored to detect a comparatively high level of delayed neutron activity. 15 refs, 8 figs, 2 tabs

Leak-before-break assessment of RBMK-1500 fuel channel in case of delayed hydride cracking

International Nuclear Information System (INIS)

Klimasauskas, A.; Grybenas, A.; Makarevicius, V.; Nedzinskas, L.; Levinskas, R.; Kiselev, V.

2003-01-01

One of the factors determining remaining lifetime of Zr-2.5% Nb fuel channel (FC) is the amount of hydrogen dissolved during corrosion process. When the concentration of hydrogen exceeds the terminal solid solubility limit zirconium hydrides are precipitated. As a result form necessary conditions for delayed hydride cracking (DHC). Data from the RBMK-1500 fuel channel tubes (removed from service) shows that hydrogen in some cases distributes unevenly and hydrogen concentration can differ several times between individual FC tubes or separate zones of the same tube and possibly, can reach dangerous levels in the future. Consequently, lacking statistical research data, it is difficult to forecast increase of hydrogen concentration and formation of DHC. So it is important to verify if under the most unfavorable situation leak before break condition will be satisfied in the case of DHC. To estimate possible DHC rates in RBMK 1500 FC pressure tubes experiments were done in the following order: hydriding of the Zr-2.5Nb pressure tube material to the required hydrogen concentration; hydrogen analysis; machining of specimens, fatigue crack formation in the axial direction, DHC testing; average crack length measurement and DHC velocity calculation. During the tests in average DHC values were determined at 283, 250 and 144 degC (with hydrogen concentrations correspondingly 76, 54 and 27 ppm). The fracture resistance dependence from hydrogen concentration was measured at 20 degC. To calculate leak through the postulated flaw, statistical distribution of DHC surface irregularity was determined. Leak before break analysis was carried out according to requirements of RBMK 1500 regulatory documents. J integral and crack opening were calculated using finite element method. Loading of the FC was determined using RELAP5 code. Critical crack length was calculated using R6 and J-integral methods. Coolant flow rate through the postulated crack was estimated using SQUIRT software

Grid spacers for use in a nuclear fuel assembly

International Nuclear Information System (INIS)

Kuwako, Akira.

1987-01-01

Purpose: To obtain spacers capable of reducing the pressure loss by enlarging coolant flow channels when the fuel temperature is high, while capable of reliably maintaining the fuel pins with no vibrations when the fuel temperature is low. Constitution: This invention concerns grid spacers for constituting fuel assemblies for use in water cooled reactors. Memory shape alloys are disposed at least a portion of a spacer element that takes such a shape as urging the pin when the fuel temperature is low, while enlarging the coolant flow channel to reduce the pressure loss when the fuel temperature is high. (Ikeda, J.)

Influence of single-phase heat transfer correlations on safety analysis of research reactors with narrow rectangular fuel channels

International Nuclear Information System (INIS)

Rawashdeh, A.; Altamimi, R.; Lee, B.; Chung, Y. J.; Park, S.

2013-01-01

The influence of different single-phase heat transfer correlations on the fuel temperature and minimum critical heat flux ratio (MCHFR) during a typical accident of a 5 MW research reactor is investigated. A reactor uses plate type fuel, of which the cooling channels have a narrow rectangular shape. RELAP5/MOD3.3 tends to over-predict the Nusselt number (Nu) at a low Reynolds number (Re) region, and therefore the correlation set is modified to properly describe the thermal behavior at that region. To demonstrate the effect of Nu at a low-Re region on an accident analysis, a two-pump failure accident was chosen as a sample problem. In the accident, the downward core flow decreases by a pump coast-down, and then reverses upward by natural convection. During the pump coast-down and flow reversal, the flow undergoes a laminar flow regime which has a different Nu with respect to the correlation sets. Compared to the results by the original RELAP5/MOD3.3, the modified correlation set predicts the fuel temperature to be a little higher than the original value, and the MCHFR to be a little lower than the original value. Although the modified correlation set predicts the fuel temperature and the MCHFR to be less conservative than those calculated from the original correlation of RELAP5/MOD3.3, the maximum fuel temperature and the MCHFR still satisfy the safety acceptance criteria

Parallel channel effects under BWR LOCA conditions

International Nuclear Information System (INIS)

Suzuki, H.; Hatamiya, S.; Murase, M.

1988-01-01

Due to parallel channel effects, different flow patterns such as liquid down-flow and gas up-flow appear simultaneously in fuel bundles of a BWR core during postulated LOCAs. Applying the parallel channel effects to the fuel bundle, water drain tubes with a restricted bottom end have been developed in order to mitigate counter-current flow limiting and to increase the falling water flow rate at the upper tie plate. The upper tie plate with water drain tubes is an especially effective means of increasing the safety margin of a reactor with narrow gaps between fuel rods and high steam velocity at the upper tie plate. The characteristics of the water drain tubes have been experimentally investigated using a small-scaled steam-water system simulating a BWR core. Then, their effect on the fuel cladding temperature was evaluated using the LOCA analysis program SAFER. (orig.)

Fuel assembly for BWR type reactor

International Nuclear Information System (INIS)

Kato, Shigeru.

1993-01-01

In the fuel assembly of the present invention, a means for mounting and securing short fuel rods is improved. Not only long fuel rods but also short fuel rods are disposed in channel of the fuel assembly to improve reactor safety. The short fuel rods are supported by a screw means only at the lower end plug. The present invention prevents the support for the short fuel rod from being unreliable due to the slack of the screw by the pressure of inflowing coolants. That is, coolant abutting portions such as protrusions or concave grooves are disposed at a portion in the channel box where coolants flowing from the lower tie plate, as an uprising stream, cause collision. With such a constitution, a component caused by the pressure of the flowing coolants is formed. The component acts as a rotational moment in the direction of screwing the male threads of the short fuel rod into the end plug screw hole. Accordingly, the screw is not slackened, and the short fuel rods are mounted and secured certainly. (I.S.)

Apparatus for mixing fuel in a gas turbine

Science.gov (United States)

Uhm, Jong Ho; Johnson, Thomas Edward

2015-04-21

A combustor nozzle includes an inlet surface and an outlet surface downstream from the inlet surface, wherein the outlet surface has an indented central portion. A plurality of fuel channels are arranged radially outward of the indented central portion, wherein the plurality of fuel channels extend through the outlet surface.

Fuel assembly

International Nuclear Information System (INIS)

Chaki, Masao; Nishida, Koji; Karasawa, Hidetoshi; Kanazawa, Toru; Orii, Akihito; Nagayoshi, Takuji; Kashiwai, Shin-ichi; Masuhara, Yasuhiro

1998-01-01

The present invention concerns a fuel assembly, for a BWR type nuclear reactor, comprising fuel rods in 9 x 9 matrix. The inner width of the channel box is about 132mm and the length of the fuel rods which are not short fuel rods is about 4m. Two water rods having a circular cross section are arranged on a diagonal line in a portion of 3 x 3 matrix at the center of the fuel assembly, and two fuel rods are disposed at vacant spaces, and the number of fuel rods is 74. Eight fuel rods are determined as short fuel rods among 74 fuel rods. Assuming the fuel inventory in the short fuel rod as X(kg), and the fuel inventory in the fuel rods other than the short fuel rods as Y(kg), X and Y satisfy the relation: X + Y ≥ 173m, Y ≤ - 9.7X + 292, Y ≤ - 0.3X + 203 and X > 0. Then, even when the short fuel rods are used, the fuel inventory is increased and fuel economy can be improved. (I.N.)

Water transport in the cathode channels of direct methanol fuel cells; Wasseraustrag aus den Kathodenkanaelen von Direkt-Methanol-Brennstoffzellen

Energy Technology Data Exchange (ETDEWEB)

Schroeder, Alexander

2011-10-26

Mass transport phenomena are vital for the operating performance of direct methanol fuel cells. In particular, the discharge of liquid water from the cathode channels is crucial for the supply of oxygen to the cathode and thus for operational stability. Droplets of water in the pores of the the diffusion layer and the cathode channels may lower the power output and induce locally negative current densities as they considerably limit the oxygen supply. This work investigates the water discharge from the cathode channels using neutron radiography, synchrotron radiography and locally resolved current density measurements and it identifies ways of improving the operational stability. Neutron radiography is a measuring technique suitable for detecting the water distribution in fuels cells under operating conditions. Synchrotron radiography is a method complementary to neutron radiography, allowing a more detailed analysis of smaller areas. Special test cells adapted to both measuring methods are developed. Their electrode areas are radiographed either frontally or laterally. To enable locally resolved current density measurements, a printed circuit board with a segmented contact area is integrated into each of the test cells. The measuring technique used is based on compensated sensor resistors, which ensure a reactionless measurement. In addition, the temperature distribution and the pressure drop on the cathod side are recorded. In order to correlated the water distribution, the current density distribution and the pressure drop, neutron radiography and synchrotron radiography are both combined with locally resolved current density measurements. Furthermore, current density measurements are performed under constant laboratory conditions to study the variation of paramenters. A measurement with a stack is also performed. The experiments reveal fundamental interdependencies between different factors and the discharge of water. At a given air ratio, the geometry and the

TAPIR, Thermal Analysis of HTGR with Graphite Sleeve Fuel Elements

International Nuclear Information System (INIS)

Weicht, U.; Mueller, W.

1983-01-01

1 - Nature of the physical problem solved: Thermal analysis of a reactor core containing internally and/or externally gas cooled prismatic fuel elements of various geometries, rating, power distribution, and material properties. 2 - Method of solution: A fuel element in this programme is regarded as a sector of a fuelled annulus with graphite sleeves of any shape on either side and optional annular gaps between fuel and graphite and/or within the graphite. It may have any centre angle and the fuelled annulus may become a solid cylindrical rod. Heat generation in the fuel is assumed to be uniform over the cross section and peripheral heat flux into adjacent sectors is ignored. Fuel elements and coolant channels are treated separately, then linked together to fit a specified pattern. 3 - Restrictions on the complexity of the problem: Maxima of: 50 fuel elements; 50 cooled channels; 25 fuel geometries; 25 coolant channel geometries; 10 axial power distributions; 10 graphite conductivities

Fuel spacer

International Nuclear Information System (INIS)

Nishida, Koji; Yokomizo, Osamu; Kanazawa, Toru; Kashiwai, Shin-ichi; Orii, Akihito.

1992-01-01

The present invention concerns a fuel spacer for a fuel assembly of a BWR type reactor and a PTR type reactor. Springs each having a vane are disposed on the side surface of a circular cell which supports a fuel rods. A vortex streams having a vertical component are formed by the vanes in the flowing direction of a flowing channel between adjacent cylindrical cells. Liquid droplets carried by streams are deposited on liquid membrane streams flowing along the fuel rod at the downstream of the spacer by the vortex streams. In view of the above, the liquid droplets can be deposited to the fuel rod without increasing the amount of metal of the spacer. Accordingly, the thermal margin of the fuel assembly can be improved without losing neutron economy. (I.N.)

Fast leak in channel H9

International Nuclear Information System (INIS)

Bauer, E.

1987-01-01

The loss of seal of the H9 channel in vacuum, freeing the entire cross section of the front part, leads to a fast leak that progresses rapidly. The effect of depressurizing the reflector can leads to shutdown of the shutdown rod pumps. The source changer associated with the channel fills completely before the valve closes. All of the leak water remains contained within the source changer containment. After the valves open, cooling of the fuel element is handled by natural convection, requiring a reversal of the flow between the plates. This changeover, which takes place at a relatively low pressure level, could lead to local boiling in the fuel element. Consequently, irreversible transformations cannot be excluded as possibilities for the fuel element and even for the control rod. Subsequently, the can is refilled with heavy water with establishment of the usual pressure levels

Hot channel calculation methodologies in case of Gd burnable poison

International Nuclear Information System (INIS)

Panka, I.; Kereszturi, A.

2008-01-01

The final step in the safety analysis is the investigation of the fulfilment of the acceptance criteria using hot channel calculations. Recently, there has been under way at Paks NPP to introduce a new, higher enriched (4.2 %) fuel type containing Gd burnable poison. To do that, for some transients the DBA analyses must be repeated and last year, as one of the first steps in this process, it was needed to review the hot channel calculation methodologies used in the analyses. The goal of the paper is to summarize some aspects of the hot channel calculation methodologies using different lattice pitches and different fuel types (Gd or non Gd and different enrichments). Mainly, three topics are discussed. First, the influence of the radial power distribution (and other burnup dependent parameters) inside the fuel pin are investigated, and then we discuss the problem of the selection of the appropriate 'frame parameter' in connection with the initial power level at the initial stationary state of DBA transients. Finally, we are trying to answer the question: is it possible to build up a conservative single closed sub-channel approach against multi channel approach?(Authors)

Measurement of the heavy water level in the fuel channels of the RA reactor - Annex 11; Prilog 11- Merenje nivoa teske vode u gorivnim kanalima reaktora RA

Energy Technology Data Exchange (ETDEWEB)

Nikolic, M [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

1964-12-15

The objective of measuring the heavy water level in the reactor channels was to verify experimentally the possibilities of reactor cooling with parallel operation of heavy water pumps od 1500 rotations/min at nominal power of 6.5 MW. Measurements were done in 2 periphery and 2 central fuel channels with pumps speed 1500, 1800 and 3000 rotations/min by a contact probe with electric resistance measuring device. precision of the measurement was {+-}1 cm.

Advanced combinational microfluidic multiplexer for fuel cell reactors

International Nuclear Information System (INIS)

Lee, D W; Kim, Y; Cho, Y-H; Doh, I

2013-01-01

An advanced combinational microfluidic multiplexer capable to address multiple fluidic channels for fuel cell reactors is proposed. Using only 4 control lines and two different levels of control pressures, the proposed multiplexer addresses up to 19 fluidic channels, at least two times larger than the previous microfluidic multiplexers. The present multiplexer providing high control efficiency and simple structure for channel addressing would be used in the application areas of the integrated microfluidic systems such as fuel cell reactors and dynamic pressure generators

Basic experimental study with visual observation on elimination of the re-criticality issue using the MELT-II facility. Simulated fuel-escape behavior through a coolant channel

International Nuclear Information System (INIS)

Matsuba, Ken-ichi; Imahori, Shinji; Isozaki, Mikio

2004-11-01

In a core disruptive accident of fast reactors, fuel escape from the reactor core is a key phenomenon for prevention of re-criticality with significant mechanical-energy release subsequent to formation of a large-scale fuel pool with high mobility. Therefore, it is effective to study possibility of early fuel escape through probable escape paths such as a control-rod-guide-tube space well before high-mobility-pool formation. The purpose of the present basic experimental study is to clarify the mechanism of fuel-escape under a condition expected in the reactor situation, in which some amount of coolant may be entrapped into the molten-fuel pool. The following results have been obtained through basic experiments in which molten Wood's metal (components: 60wt%Bi-20wt%Sn-20wt%In, density at the room temperature: 8700 kg/m 3 , melting point: 78.8degC) is ejected into an coolant channel filled with water. (1) In the course of melt ejection, a small quantity of coolant is forced to be entrapped into the melt pool as a result of thermal interactions leading to high-pressure rise within the coolant channel. (2) Melt ejection is accelerated by pressure build-up which results from vapor pressure of entrapped coolant within the melt pool. (3) Average melt-ejection rate tends to increase in lower coolant-subcooling conditions, in which pressure build-up within the melt pool is enhanced. These results indicate a probability of a phenomenon in which melt ejection is accelerated by entrapment of coolant within a melt pool. Through application of the mechanism of confirmed phenomenon into the reactor condition, it is suggested that fuel escape is enhanced by entrapment of coolant within a fuel pool. (author)

An Evaluation of a Fission Product Inventory for CANDU Fuels

International Nuclear Information System (INIS)

Jung, Jong Yeob; Park, Joo Hwan

2007-01-01

Fission products are released by two processes when a single channel accident occurs. One is a 'prompt release' and the other is a 'delayed release'. Prompt release assumes that the gap inventory of the fuel elements is released by a fuel element failure at the time of an accident. Delayed release assumes that the inventories within the grain or at the grain boundary are released after a accident due to a diffusion through grains, an oxidation of the fuel and an interaction between the fuel and the Zircaloy sheath. Therefore, the calculation of a fission product inventory and its distribution in a fuel during a normal operating is the starting point for the assessment of a fission product release for single channel accidents. In this report, the fission product inventories and their distributions within s fuel under a normal operating condition are evaluated for three types of CANDU fuels such as the 37 element fuel, CANFLEX-NU and CANFLEX-RU fuel bundles in the 'limiting channel'. To accomplish the above mentioned purposes, the basic power histories for each type of CANDU fuel were produced and the fission product inventories were calculated by using the ELESTRES code

On slip correlations used in the determination of the void distribution of BWR fuel channels

International Nuclear Information System (INIS)

Oelgaard, P.L.

1995-05-01

When performing calculations on boiling water reactors it is necessary to establish a relation between the average velocity of the steam, V g , and that of the water, V f , in the fuel channels. This is usually done through establishing an expression for the slip ratio S=V g /V f . In the literature a number of such slip correlations - all based on measurements - has been presented. A comparison between some of these correlations has been performed in this paper. While the correlations have some general trends in common, the numerical values of S obtained with the correlations may vary significantly. Further, in spite of the fact that various flow regimes into account. This raises the question: How reliable is the use of slip correlations? (au)

Criticality evaluations of scrambled fuel in water basin storage

International Nuclear Information System (INIS)

Fast, E.

1989-01-01

Fuel stored underwater in the Idaho Chemical Processing Plant basins has been subjected to the usual criticality safety evaluations to assure safe storage configurations. Certain accident or emergency conditions, caused by corrosion or a seismic event, could change the fuel configuration and environment to invalidate previous calculations. Consideration is given here to such contingencies for fuel stored in three storage basins. One basin has fuel stored in racks, on a generally flat floor. In the other two basins, the fuel is stored on yokes and in baskets suspended from a monorail system. The floor is ribbed with 30.48-cm-thick and 80-cm-high concrete barriers across the basin width and spaced 30.48 cm apart. The suspended fuel is typically down to 15 cm above the floor of the channel between the concrete barriers. These basins each have 29 channels of 18 positions maximum per channel for a total of 522 possible positions, which are presently 77 and 49% occupied. The three basins are hydraulically interconnected. Several scenarios indicate possible changes in the fuel configuration. An earthquake could rupture a basin wall or floor, allowing the water to drain from all basins. All levels of water would fall to the completely drained condition. Suspended fuel could drop and fall over within the channel. Corrosion might weaken the support systems or cause leaks in sealed fuel canisters. Calculations were made with the KENO-IV criticality program and the library of mostly Hansen-Roach 16-energy-group neutron cross sections

Linear stability of a fuel channel uniformly heated considering retrofeeding by vacuum. Theoretical study

International Nuclear Information System (INIS)

Gonzalez M, V.; Salinas H, J.G.; Espinosa P, G.

2000-01-01

The Technology, Regulation and Services Management of the National Commission of Nuclear Safety and Safeguards in coordinated form with the IPH Department of the Metropolitan Autonomous-Iztapalapa University, developed the present project to study the linear stability in a fuel channel uniformly heated with effects of retrofeeding by vacuums. In this study the methodology used in the analysis of linear stability of the nuclear reactor unit 1 at Laguna Verde power plant is described which represented by an average channel uniformly heated. The conceptual model consists of two cells which represent the two regions in which is divided the channel according to the cooling is in one and two phases, considering the boiling length dependent in the time. It is used the homogeneous flux models for describing the thermohydraulic behavior of the cooling in the two phases region. The neutron processes with the punctual model of the neutron kinetics with a group of retarded neutrons precursors are described. It is studied the behavior of the system in the frequency domain with the transfer functions obtained and it is characterized in four operation states corresponding to the four corners of the low stability zone in the map power-flow Laguna Verde power plant. For these operation states the characteristic frequency is determined and the corresponding Nyquist diagrams are obtained. The results show that the system stability depends on the power-flow relation and that the operations which implicate a reduction of this relation improve the stability of the system (reducing the power introducing control bars with constant cooling flow or increase cooling flow with bars pattern established). The obtained results with effects of retrofeeding by vacuums show that the value of the characteristic frequency is modified very little with respect to the model without retrofeeding, therefore the thermohydraulic processes seem to determine the response of the stability of the system

Gas Test Loop Booster Fuel Hydraulic Testing

International Nuclear Information System (INIS)

Gas Test Loop Hydraulic Testing Staff

2006-01-01

The Gas Test Loop (GTL) project is for the design of an adaptation to the Advanced Test Reactor (ATR) to create a fast-flux test space where fuels and materials for advanced reactor concepts can undergo irradiation testing. Incident to that design, it was found necessary to make use of special booster fuel to enhance the neutron flux in the reactor lobe in which the Gas Test Loop will be installed. Because the booster fuel is of a different composition and configuration from standard ATR fuel, it is necessary to qualify the booster fuel for use in the ATR. Part of that qualification is the determination that required thermal hydraulic criteria will be met under routine operation and under selected accident scenarios. The Hydraulic Testing task in the GTL project facilitates that determination by measuring flow coefficients (pressure drops) over various regions of the booster fuel over a range of primary coolant flow rates. A high-fidelity model of the NW lobe of the ATR with associated flow baffle, in-pile-tube, and below-core flow channels was designed, constructed and located in the Idaho State University Thermal Fluids Laboratory. A circulation loop was designed and constructed by the university to provide reactor-relevant water flow rates to the test system. Models of the four booster fuel elements required for GTL operation were fabricated from aluminum (no uranium or means of heating) and placed in the flow channel. One of these was instrumented with Pitot tubes to measure flow velocities in the channels between the three booster fuel plates and between the innermost and outermost plates and the side walls of the flow annulus. Flow coefficients in the range of 4 to 6.5 were determined from the measurements made for the upper and middle parts of the booster fuel elements. The flow coefficient for the lower end of the booster fuel and the sub-core flow channel was lower at 2.3

Gas Test Loop Booster Fuel Hydraulic Testing

Energy Technology Data Exchange (ETDEWEB)

Gas Test Loop Hydraulic Testing Staff

2006-09-01

The Gas Test Loop (GTL) project is for the design of an adaptation to the Advanced Test Reactor (ATR) to create a fast-flux test space where fuels and materials for advanced reactor concepts can undergo irradiation testing. Incident to that design, it was found necessary to make use of special booster fuel to enhance the neutron flux in the reactor lobe in which the Gas Test Loop will be installed. Because the booster fuel is of a different composition and configuration from standard ATR fuel, it is necessary to qualify the booster fuel for use in the ATR. Part of that qualification is the determination that required thermal hydraulic criteria will be met under routine operation and under selected accident scenarios. The Hydraulic Testing task in the GTL project facilitates that determination by measuring flow coefficients (pressure drops) over various regions of the booster fuel over a range of primary coolant flow rates. A high-fidelity model of the NW lobe of the ATR with associated flow baffle, in-pile-tube, and below-core flow channels was designed, constructed and located in the Idaho State University Thermal Fluids Laboratory. A circulation loop was designed and constructed by the university to provide reactor-relevant water flow rates to the test system. Models of the four booster fuel elements required for GTL operation were fabricated from aluminum (no uranium or means of heating) and placed in the flow channel. One of these was instrumented with Pitot tubes to measure flow velocities in the channels between the three booster fuel plates and between the innermost and outermost plates and the side walls of the flow annulus. Flow coefficients in the range of 4 to 6.5 were determined from the measurements made for the upper and middle parts of the booster fuel elements. The flow coefficient for the lower end of the booster fuel and the sub-core flow channel was lower at 2.3.

Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

Energy Technology Data Exchange (ETDEWEB)

Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro, E-mail: duvan.castellanos@ufabc.edu.br, E-mail: joao.moreira@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: pedro.rossi@ufabc.edu.br, E-mail: pedro.carajilescov10@gmail.com [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil). Centro de Engenharias, Modelagem e Ciências Sociais Aplicadas

2017-07-01

To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

International Nuclear Information System (INIS)

Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro

2017-01-01

To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

Relativistic laser channeling in plasmas for fast ignition

Science.gov (United States)

Lei, A. L.; Pukhov, A.; Kodama, R.; Yabuuchi, T.; Adumi, K.; Endo, K.; Freeman, R. R.; Habara, H.; Kitagawa, Y.; Kondo, K.; Kumar, G. R.; Matsuoka, T.; Mima, K.; Nagatomo, H.; Norimatsu, T.; Shorokhov, O.; Snavely, R.; Yang, X. Q.; Zheng, J.; Tanaka, K. A.

2007-12-01

We report an experimental observation suggesting plasma channel formation by focusing a relativistic laser pulse into a long-scale-length preformed plasma. The channel direction coincides with the laser axis. Laser light transmittance measurement indicates laser channeling into the high-density plasma with relativistic self-focusing. A three-dimensional particle-in-cell simulation reproduces the plasma channel and reveals that the collimated hot-electron beam is generated along the laser axis in the laser channeling. These findings hold the promising possibility of fast heating a dense fuel plasma with a relativistic laser pulse.

A fast-running fuel management program for a CANDU reactor

International Nuclear Information System (INIS)

Choi, Hangbok

2000-01-01

A fast-running fuel management program for a CANDU reactor has been developed. The basic principle of this program is to select refueling channels such that the reference reactor conditions are maintained by applying several constraints and criteria when selecting refueling channels. The constraints used in this program are the channel and bundle power and the fuel burnup. The final selection of the refueling channel is determined based on the priority of candidate channels, which enhances the reactor power distribution close to the time-average model. The refueling simulation was performed for a natural uranium CANDU reactor and the results were satisfactory

TRANTHAC-1: transient thermal-hydraulic analysis code for HTGR core of multi-channel model

International Nuclear Information System (INIS)

Sato, Sadao; Miyamoto, Yoshiaki

1980-08-01

The computer program TRANTHAC-1 is for predicting thermal-hydraulic transient behavior in HTGR's core of pin-in-block type fuel elements, taking into consideration of the core flow distribution. The program treats a multi-channel model, each single channel representing the respective column composed of fuel elements. The fuel columns are grouped in flow control regions; each region is provided with an orifice assembly. In the region, all channels are of the same shape except one channel. Core heat is removed by downward flow of the control through the channel. In any transients, for given time-dependent power, total core flow, inlet coolant temperature and coolant pressure, the thermal response of the core can be determined. In the respective channels, the heat conduction in radial and axial direction are represented. And the temperature distribution in each channel with the components is calculated. The model and usage of the program are described. The program is written in FORTRAN-IV for computer FACOM 230-75 and it is composed of about 4,000 cards. The required core memory is about 75 kilowords. (author)

Apparatus for loading fuel pellets in fuel rods

International Nuclear Information System (INIS)

Tedesco, R.J.

1976-01-01

An apparatus is disclosed for loading fuel pellets into fuel rods for a nuclear reactor including a base supporting a table having grooves therein for holding a multiplicity of pellets. Multiple fuel rods are placed in alignment with grooves in the pellet table and a guide member channels pellets from the table into the corresponding fuel rods. To effect movement of pellets inside the fuel rods without jamming, a number of electromechanical devices mounted on the base have arms connected to the lower surface of the fuel rod table which cyclically imparts a reciprocating arc motion to the table for moving the fuel pellets longitudinally of and inside the fuel rods. These electromechanical devices include a solenoid having a plunger therein connected to a leaf type spring, the arrangement being such that upon energization of the solenoid coil, the leaf spring moves the fuel rod table rearwardly and downwardly, and upon deenergization of the coil, the spring imparts an upward-forward movement to the table which results in physical displacement of fuel pellets in the fuel rods clamped to the table surface. 8 claims, 6 drawing figures

Experiment on transient heat transfer in closed narrow channel

International Nuclear Information System (INIS)

Ochiai, Masaaki

1985-01-01

Heat transfer coefficients and transient pressures in closed narrow channels were obtained experimentally, in order to assess the gap heat transfer models in the computer code WTRLGD which were devised to analyze the internal pressure behavior of waterlogged fuel rods. Gap widths of channels are 0.1--0.5mm to simulate the gap region of waterlogged fuel rods, and test fluids are water (7--89.2 0 C) and Freon-113 (9.2 0 C). The results show that the heater temperature and the pressure measured in the experiments without the DNB occurrence are simulated fairly well by the calculational model of WTRLGD where the heat transfer in a closed narrow channel is evaluated with one-dimensional transient thermal conduction equation and Jens and Lottes' correlation for nucleate boiling. Consequently, it is also suggested that the above equations are available for evaluation of heat flux from fuel to internal water of waterlogged fuel rods. The film boiling heat transfer coefficient was in the same order of that evaluated by Bromley's correlation and the DNB heat flux was smaller than that obtained in quasi-steady experiments with ordinary systems, although the experimental data for them were not enough. (author)

Fuel temperature characteristics of the 37-element and CANFLEX fuel bundle

International Nuclear Information System (INIS)

Bae, Jun Ho; Rho, Gyu Hong; Park, Joo Hwan

2009-10-01

This report describes the fuel temperature characteristics of CANFLEX fuel bundles and 37-element fuel bundles for a different burnup of fuel. The program was consisted for seeking the fuel temperature of fuel bundles of CANFLEX fuel bundles and 37-element fuel bundles by using the method in NUCIRC. Fuel temperature has an increasing pattern with the burnup of fuel for CANFLEX fuel bundles and 37-element fuel bundles. For all the case of burnup, the fuel temperature of CANFLEX fuel bundles has a lower value than that of 37-element fuel bundles. Especially, for the high power channel, the CANFLEX fuel bundles show a lower fuel temperature as much as about 75 degree, and the core averaged fuel temperature has a lower fuel temperature of about 50 degree than that of 37-element fuel bundles. The lower fuel temperature of CANFLEX fuel bundles is expected to enhance the safety by reducing the fuel temperature coefficient. Finally, for each burnup of CANFLEX fuel bundles and 37-element fuel bundles, the equation was present for predicting the fuel temperature of a bundle in terms of a coolant temperature and bundle power

Axial shuffling fuel-management schemes for 1.2% SEU in CANDU

International Nuclear Information System (INIS)

Younis, M.H.; Boczar, P.G.

1989-11-01

The use of slightly enriched uranium (SEU) in CANDU (CANada Deuterium Uranium) requires a different fuel-management strategy than that usually employed with natural uranium fuel. Axial shuffling is a fuel-management strategy in which some or all of the fuel bundles are removed from the channel, rearranged, and reinserted into the same channel, along with fresh fuel. An axial shuffling scheme has been devised for 1.2% SEU which results in excellent power profiles, from the perspectives of both good axial flattening and power histories. With the CANFLEX (CANdu FLEXible fuelling) advanced fuel bundle, fuel rating can be reduced to below 40kW/m, with consequent safety benefits

1200 FPD refuelling simulation of RUFIC fuel in a CANDU 6 reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Soon Young; Jeong, Chang Joon; Min, Byung Joo; Suk, Ho Chun

2001-07-01

The refuelling strategy of RUFIC (Recovered Uranium Fuel in CANDU) fuel as a high-burnup fuel for a CANDU 6 reactor is studied to determine the achievable operation characteristics of the fuel and reactor. In this study, three refuelling schemes of 4-, 2-, and 3-bundle shift for 0.92 w/o RUFIC fuel in an CANDU 6 reactor were individually evaluated through 1200 FPD(Full Power Day)refuelling simulaltions where the 0.92 w/o RUFIC is equivalent to CANFLEX 0.9 w/o SEU(Slightly Enriched Uranium) in reactivity and burnup respects. The computer code system used for this study is WIMS-AECL/DRAGON/RFSP. The results simulated for the case of 4-bundle shift refueling scheme shows that the peak maximum channel power and peak maximum CPPF(Channel Power Peaking Factor)of 7228 kW and 1.175, respectively, seems too high to maintain the available operating margins, because some data of the maximum channel power exceed the operating limit(7070 kW based on the Technical Specifications of Wolsong 3 and 4 Units). Whereas, the results simulated for the case of 2-bundle shift refuelling scheme shows that sufficient operating margin could be secured where the peak maximum channel power and peak maximum CPPF were 6889 kW and 1.094, respectively. However, the channel refuelling rate (channels/day) of the 2-bundle shift refuelling scheme is twice that of the 4-bundle shift refuelling scheme, and hence the 2-bundle shift refuelling would not be an economical refuelling scheme for the RUFIC fuel bundles. Therefore, a 3-bundle shift refuelling scheme for the RUFIC fuel in CANDU 6 reactor was also studied by the 1200 FPD refuelling simulation. As a result, it is found that all the operating parameters in the 3-bundle shift case are achivable for the CANDU 6 reactor operation, and the channel refuelling rate of 2.88 channels/day seems to be attractive compared to the refuelling rate of 4.32 channels/day in the 2-bundle shift case.

Fuel cell membrane hydration and fluid metering

Science.gov (United States)

Jones, Daniel O.; Walsh, Michael M.

1999-01-01

A hydration system includes fuel cell fluid flow plate(s) and injection port(s). Each plate has flow channel(s) with respective inlet(s) for receiving respective portion(s) of a given stream of reactant fluid for a fuel cell. Each injection port injects a portion of liquid water directly into its respective flow channel in order to mix its respective portion of liquid water with the corresponding portion of the stream. This serves to hydrate at least corresponding part(s) of a given membrane of the corresponding fuel cell(s). The hydration system may be augmented by a metering system including flow regulator(s). Each flow regulator meters an injecting at inlet(s) of each plate of respective portions of liquid into respective portion(s) of a given stream of fluid by corresponding injection port(s).

Investigation on thermo-acoustic instability dynamic characteristics of hydrocarbon fuel flowing in scramjet cooling channel based on wavelet entropy method

Science.gov (United States)

Zan, Hao; Li, Haowei; Jiang, Yuguang; Wu, Meng; Zhou, Weixing; Bao, Wen

2018-06-01

As part of our efforts to find ways and means to further improve the regenerative cooling technology in scramjet, the experiments of thermo-acoustic instability dynamic characteristics of hydrocarbon fuel flowing have been conducted in horizontal circular tubes at different conditions. The experimental results indicate that there is a developing process from thermo-acoustic stability to instability. In order to have a deep understanding on the developing process of thermo-acoustic instability, the method of Multi-scale Shannon Wavelet Entropy (MSWE) based on Wavelet Transform Correlation Filter (WTCF) and Multi-Scale Shannon Entropy (MSE) is adopted in this paper. The results demonstrate that the developing process of thermo-acoustic instability from noise and weak signals is well detected by MSWE method and the differences among the stability, the developing process and the instability can be identified. These properties render the method particularly powerful for warning thermo-acoustic instability of hydrocarbon fuel flowing in scramjet cooling channels. The mass flow rate and the inlet pressure will make an influence on the developing process of the thermo-acoustic instability. The investigation on thermo-acoustic instability dynamic characteristics at supercritical pressure based on wavelet entropy method offers guidance on the control of scramjet fuel supply, which can secure stable fuel flowing in regenerative cooling system.

Irradiation of novel MTR fuel plates in BR2

International Nuclear Information System (INIS)

Verboomen, B.; Aoust, Th.; Beeckmans De Westmeerbeeck, A.; De Raedt, Ch.

2000-01-01

Since the end of 1999, novel MTR fuel plates with very high-density meat are being irradiated in BR2. The purpose of the irradiation is to investigate the behaviour of these fuel plates under very severe reactor operation conditions. The novel fuel plates are inserted in two standard six-tube BR2 fuel elements in the locations normally occupied by the standard outer fuel plates. The irradiation in BR2 was prepared by carrying out detailed neutron Monte Carlo calculations of the whole BR2 core containing the two experimental fuel elements for various positions in the reactor and for various azimuthal orientations of the fuel elements. Comparing the thus determined fission density levels and azimuthal profiles in the new MTR fuel plates irradiated in the various channels allowed the experimenters to choose the most appropriate BR2 channel and the most appropriate fuel element orientation. (author)

Statistical calculation of hot channel factors

International Nuclear Information System (INIS)

Farhadi, K.

2007-01-01

It is a conventional practice in the design of nuclear reactors to introduce hot channel factors to allow for spatial variations of power generation and flow distribution. Consequently, it is not enough to be able to calculate the nominal temperature distributions of fuel element, cladding, coolant, and central fuel. Indeed, one must be able to calculate the probability that the imposed temperature or heat flux limits in the entire core is not exceeded. In this paper, statistical methods are used to calculate hot channel factors for a particular case of a heterogeneous, Material Testing Reactor (MTR) and compare the results obtained from different statistical methods. It is shown that among the statistical methods available, the semi-statistical method is the most reliable one

Boiling transition phenomenon in BWR fuel assemblies effect of fuel spacer shape on critical power

International Nuclear Information System (INIS)

Yamamoto, Yasushi; Morooka, Shin-ichi; Mitsutake, Toru; Yokobori, Seiichi; Kimura, Jiro.

1996-01-01

A thorough understanding of the thermal-hydraulic phenomena near fuel spacer is necessary for the accurate prediction of the critical power of BWR fuel assemblies, and is thus essential for effective developments of a new BWR fuel assembly. The main purpose of this study is to develop an accurate method for predicting the effect of spacer shapes on critical power. Tests have been conducted under actual BWR operating conditions, using an annulus flow channel consisting of a heated rod and circular-tube channel, and BWR simulated 4x4 rod bundles with heater rods unheated just upsteam of spacer. The effect of spacer shapes on critical power was predicted analytically based on the droplet deposition rate estimation. The droplet deposition rate for different spacer shapes was calculated using a single-phase flow model. The prediction results were compared with the test results for the annulus flow channel using ring-type spacers. Analytical results of critical power agreed with measured critical power from point of the effects of changes in the rod-spacer clearance and the spacer thickness on critical power. (author)

Multi-channel grouping techniques for conducting reactor safety studies

International Nuclear Information System (INIS)

Waltar, A.E.; Wilburn, N.P.

1975-01-01

In conducting safety studies for postulated unprotected accidents in an LMFBR system, it is common practice to employ multi-channel coupled neutronics, thermal hydraulics computer programs such as SAS3A or MELT-III. The multichannel feature of such code systems is important if the natural fuel failure incoherencies and the resulting sodium void/fuel motion reactivity feedbacks--which have strong spatial variations--are to be properly modeled. Because of the large amounts of computer time associated with many channel runs, however, there is a strong incentive to conduct parametric studies with as few channels as possible. The paper presented is focused on methods successfully employed to accomplish this end for a study of the hypothetical unprotected transient overpower accident conducted for the FFTF

Fluid moderator control system fuel assembly seal connector

International Nuclear Information System (INIS)

Veronesi, L.; Tower, S.N.; Klassen, W.E.

1987-01-01

A nuclear reactor is described comprising fuel assemblies having one or more flow channels therethrough, a core support plate having one or more flow channels therethrough, and seal connectors for sealingly connecting the one or more flow channels in the core support plate with the one or more flow channels in the fuel assemblies. The seal connectors each comprises a first portion and second portion each comprising an elongated member having a flow channel therethrough and being in substantial axial alignment with each other and being separated by a space therebetween, means for sealingly connecting the first portion o one or the one or more flow channels in the fuel assemblies, means for sealingly connecting the second portion to the first portion and for allowing relative motion between the portions, means for limiting the relative motion of the first and second portion in directions toward and away from each other, means for reconnectingly connecting and resealingly sealing the second portion to one of the one or more flow channels in the core support plate. It comprises a slip fit connection whereby the remote end of the second portion fits within whereby the remote end of the second portion fits within an opening in the core support plate which is in flow communication with the one or more flow channels in the core support plate and further comprises a ball and cones seal in series with axially spaced ring seals

Variegated operation of MAPS reactors after enmasse' coolant channel replacement: a tale-tell signature of high standard fuel bundle production quality

International Nuclear Information System (INIS)

Jena, J.K.; Sahu, J.K.; Arularasan, V.; Sivagurnathan, D.; Rathakrishnan, S.; Ramamurthy, K.

2009-01-01

After the Enmasse' Coolant Channel Replacement (EMCCR) of both the reactors of Madras Atomic Power Station (MAPS), they have put up a good performance, as far as core integrity is considered. This is a tale-tell signature of the high quality of the fuel bundles manufactured by Nuclear Fuel Complex (NFC), Hyderabad. Both the reactor cores have been loaded with various types of fuel bundles viz. Natural Uranium (NU), Depleted Uranium (DU), and Deeply Depleted Uranium (DDU) and were operated at different power level with different flux configuration at different stages of operation. Even around 1026 low burn up bundle (<2500 MWD/TeU) were transferred from MAPS-1 to MAPS-2, first time in the history of PHWRS. During all such variegated operations, the Primary Heat Transport (PHT) system 131 I activity, which is synonymous with the core integrity, was maintaining low for most of the reactor operation period. However, recently a low burn up fuel bundle failure has been observed in MAPS-1. Even though the overall failure rate is very low, the cause of such failure needs to be ascertained for taking appropriate action to maintain the high standards of quality in the manufacturing process of the fuel bundles. (author)

Reactor and fuel assembly

International Nuclear Information System (INIS)

Ishii, Yoshihiko; Bessho, Yasunori; Sano, Hiroki; Yokomizo, Osamu; Yamashita, Jun-ichi.

1990-01-01

The present invention realizes an effective spectral operation by applying an optimum pressure loss coefficient while taking the characteristics of a lower tie plate into consideration. That is, the pressure loss coefficient of the lower tie plate is optimized by varying the cross sectional area of a fuel assembly flow channel in the lower tie plate or varying the surface roughness of a coolant flow channel in the lower tie plate. Since there is a pressure loss coefficient to optimize the moderator density over a flow rate change region, the effect of spectral shift rods can be improved by setting the optimum pressure loss coefficient of the lower tie plate. According to the present invention, existent fuel assemblies can easily be changed successively to fuel assemblies having spectral shift rods of a great spectral shift effect by using existent reactor facilities as they are. (I.S.)

Numerical simulation for rib and channel position effect on PEMFC performances

Energy Technology Data Exchange (ETDEWEB)

Liu, Zhixiang; Wang, Cheng; Mao, Zongqiang [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Zhang, Haiyong [Lab of Sail, Cable and Paints, Navy Equipment Technology Institute, Beijing, 102442 (China)

2010-04-15

Simulation is an important method for engineers to probe the detailed transportation and reaction information inside fuel cells and guide their designs without large amount of experiments. Although many papers discussing fuel cell flow fields design could be found in documents, relative positions of the ribs and channels in the anode and cathode flow field plates haven't been paid attention to surprisingly. In this paper, simulation results were given to explain the influences of relative positions of the ribs and channels in the anode and cathode flow field plates on the proton exchange membrane fuel cell (PEMFC) performances. It is interesting that the influence differs with several factors and the information will be helpful for fuel cell design. (author)

Study on the HTGR axial fuel loading

International Nuclear Information System (INIS)

Tanaka, Ryokichi

1981-01-01

In the nuclear and thermal design of reactor cores, it is one of the important targets for reactor safety to flatten fuel temperature distribution as far as possible to prevent local peaking. As a macroscopic method to prevent temperature peaking, it is considered to give exponential type power output distribution in coolant flow direction, while flattening radial power output distribution. Assuming rod-shaped fuel, the distribution of fuel heat generation is given by an exponential function under constant maximum fuel temperature condition in the direction of channel. By applying this function to neutron source distribution, and in a premise that U-235 loading can be changed continuously, the preliminary investigation on no-reflector core by one-dimensional one-group consideration, and then the analytical solution of the diffusion equation for a core with reflectors by two group one-dimensional approximation were carried out. The results of these investigations revealed that the U-235 concentration required for achieving exponential type power output distribution is necessary to have large concentration gradient up to the distance equivalent to the length of a few fuel elements from the core inlet, but it is sufficient to have constant concentration in downstream fuel elements, which is 0.8 to 0.9 times as much as the average value along the channel, except for large flow rate channel. (Wakatsuki, Y.)

Experimental study of flow induced vibration of the planar fuel assembly

International Nuclear Information System (INIS)

Wang Jinhua; Bo Hanliang; Jiang Shengyao; Jia Haijun; Zheng Wenxiang; Min Gang; Qu Xinxing

2005-01-01

The paper studied the flow-induced vibration of the planar fuel assembly under scour of coolant through experiments, the study includes: the characteristics of the inherent vibration, the response to the flow-induced vibration in rating condition and the confirmation of the critical flow velocity's scope of the flow flexible instability. The velocity distributions in different flow channels formed by fuel plates in the assembly were measured, and the velocity distribution in the same flow channel was also measured. The experimental conclusions includes: the inherent vibration frequency of the planar fuel assembly is different for a little in each direction. The damp ratio corresponding to the assembly each rank's inherent frequency is small, and the damp ratio decreased with the increase of the corresponding inherent frequency. The velocity in different flow channels decreased from outside to inside, and the velocity in the middle channel was the least; the velocity in the same channel decreased from inside to outside, and the velocity in the middle position was the most. The vibration swing of the fuel assembly was small at rating condition, and the vibration swing of the fuel plates was larger than side plates. The vibration of the fuel assembly increased with the increase of the velocity, the vibration of the middle fuel plate were larger than the border fuel plate, and the vibration of the border fuel plate was larger than the side plate. The large scale vibration of the flow flexible instability didn't occur in the velocity scope of 0-18.8 m/s in the experiment, so the critical flow velocity of the flow flexible instability was not in the flow velocity scope of the experiment. (authors)

CANDU reactor experience: fuel performance

International Nuclear Information System (INIS)

Truant, P.T.; Hastings, I.J.

1985-07-01

Ontario Hydro has more than 126 reactor-years experience in operating CANDU reactors. Fuel performance has been excellent with 47 000 channel fuelling operations successfully completed and 99.9 percent of the more than 380 000 bundles irradiated operating as designed. Fuel performance limits and fuel defects have had a negligible effect on station safety, reliability, the environment and cost. The actual incapability charged to fuel is less than 0.1 percent over the stations' lifetimes, and more recently has been zero

Dimensional measurement of fresh fuel bundle for CANDU reactor

International Nuclear Information System (INIS)

Jo, Chang Keun; Cho, Moon Sung; Suk, Ho Chun; Koo, Dae Seo; Jun, Ji Su; Jung, Jong Yeob

2005-01-01

This report describes the results of the dimensional measurement of fresh fuel bundles for the CANDU reactor in order to estimate the integrity of fuel bundle in two-phase flow in the CANDU-6 fuel channel. The dimensional measurements of fuel bundles are performed by using the 'CANDU Fuel In-Bay Inspection and Dimensional Measurement System', which was developed by this project. The dimensional measurements are done from February 2004 to March 2004 in the CANDU fuel storage of KNFC for the 36 fresh fuel bundles, which are produced by KNFC and are waiting for the delivery to the Wolsong-3 plant. The detail items of dimensional measurements are included fuel rod and bearing pad profiles of the outer ring in fuel bundle, diameter of fuel bundle, bowing of fuel bundle, fuel rod length, and surface profile of end plate profile. The measurement data will be compared with those of the post-irradiated bundles cooled in Wolsong-3 NPP spent fuel pool by using the same bundles and In-Bay Measurement System. So, this analysis of data will be applied for the evaluation of fuel bundle integrity in two-phase flow of the CANDU-6 fuel channel

Effect of channel aspect ratio on chemical recuperation process in advanced aeroengines

International Nuclear Information System (INIS)

Zhang, Silong; Cui, Naigang; Xiong, Yuefei; Feng, Yu; Qin, Jiang; Bao, Wen

2017-01-01

The working process of an advanced aeroengine such as scramjet with endothermic hydrocarbon fuel cooling is a chemical recuperative cycle. The design of cooling channel in terms of engine real working conditions is very important for the chemical recuperation process. To study the effects of channel aspect ratio (AR) on chemical recuperation process of advanced aeroengines, three dimensional model of pyrolysis coolant flow inside asymmetrical rectangular cooling channels with fins is introduced and validated through experiments. Cases when AR varies from 1 to 8 are carried out. In the pyrolysis zone of the cooling channel, decreasing the channel aspect ratio can reduce the temperature difference and non-uniformity of fuel conversion in the channel cross section, and it can also increase the final conversion and corresponding chemical heat absorption. A small channel aspect ratio is beneficial for the chemical recuperation process and can guarantee the engine cooling performance in the pyrolysis zone of the cooling channel. - Highlights: • Large non-uniformity of conversion is bad for the chemical recuperation. • Small channel aspect ratio is beneficial for improving the chemical recuperation effectiveness. • Small channel aspect ratio is also beneficial for reducing the engine wall temperature.

Effect of humidity content and direction of the flow of reactant gases on water management in the 4-serpentine and 1-serpentine flow channel in a PEM (proton exchange membrane) fuel cell

International Nuclear Information System (INIS)

Khazaee, I.; Sabadbafan, H.

2016-01-01

The performance of a PEM (proton exchange membrane) fuel cell depends on design and operating parameters such as relative humidity, operation pressure, and number of channels and direction of the flow of reactant gases. In this study, a three-dimensional, two-phase model has been established to investigate the water management and performance of PEM fuel cell with rectangular geometry and 1-serpentine and 4-serpentine with parallel flow, counter flow and cross flow for hydrogen and oxygen. The numerical simulation was realized with a PEM fuel cell model based on the FLUENT. The active area of each cell is 24.8 cm 2 that its weight is 1300 gr. The material of the gas diffusion layer is carbon clothes, the membrane is nafion117 and the catalyst layer is a plane with 0.004 g cm −2 platinum. Pure hydrogen is used on the anode side and oxygen on the cathode side. Simulation results are obtained for voltage as a function of current density at different humidity. The simulation results are compared with the experimental data, and the agreement is found to be good. The results show that the cell performance at lower voltages increases with increasing humidity in cell with 4-Serpentine flow channel and also in cell with 1-Serpentine flow channel, cell performance at all voltages increases with increasing humidity. In cell with 4-Serpentine and parallel flow channel cell performance is better than counter and cross flow in low voltage and in cell with 1-Serpentine and parallel flow, performance is better than counter and cross flow in high voltage. - Highlights: • Investigation new geometries of a fuel cell. • The effect of geometry on current density, oxygen and water distribution. • The effect of humidity on current density, oxygen and water distribution. • Seeing the interacting and complex electrochemical phenomena.

Radial heat transfer from fuel to moderator during LOCAs for CANDU PHW reactors

International Nuclear Information System (INIS)

Hildebrandt, J.G.; So, C.B.; Gillespie, G.E.; MacLean, G.

1983-01-01

In a postulated CANDU-PHW loss-of-coolant accident (LOCA) with coincident impaired emergency cooling, the axial transport of heat from the fuel by convection is reduced. This reduction in heat removal causes the fuel to heat up and the radial heat transfer to the moderator to become significant. This paper deals with two codes that predict the thermal response of fuel channels under LOCA conditions. New channel thermal radiation models in both RAMA, a thermalhydraulic code, and CHAN II, a fuel channel thermo-chemical code, are presented and their predictions are compared with the experimental results of an electrically heated bundle of 37 fuel pins. A second experiment, involving a single heated pin in a channel with flowing steam, is presented. The predictions of RAMA and CHAN II are compared with this experiment to verify the codes' thermo-chemical models. There is good agreement between the predictions of both codes and the experimental results

Development and application of sub-channel analysis code based on SCWR core

International Nuclear Information System (INIS)

Fu Shengwei; Xu Zhihong; Yang Yanhua

2011-01-01

The sub-channel analysis code SABER was developed for thermal-hydraulic analysis of supercritical water-cooled reactor (SCWR) fuel assembly. The extended computational cell structure, a new boundary conditions, 3 dimensional heat conduction model and water properties package were implemented in SABER code, which could be used to simulate the thermal fuel assembly of SCWR. To evaluate the applicability of the code, a steady state calculation of the fuel assembly was performed. The results indicate good applicability of the SABER code to simulate the counter-current flow and the heat exchange between coolant and moderator channels. (authors)

Fuel channel life limiting factors that dictate fuel channel maintenance requirements

International Nuclear Information System (INIS)

Richinson, P.J.; Wong, H.W.; Ellis, P.J.

1995-01-01

CANDU reactors have been operating for 33 years. The Nuclear Power Demonstration (NPD) Unit started up in 1962 and the prototype of CANDU, Douglas Point, started in 1967. The first commercial reactors, Pickering Units 1 and 2 both went into service in 1971 closely followed by Units 3 and 4 in 1972 and 1973 respectively. Operating commercial reactor experience represents over 10,000 pressure tubes, not including the replaced channels in all the Pickering A Units, and nearly 130,000 pressure tube operating years. No pressure tube has yet operated for its 30 year design lifetime of 210 KEFPH at 80% capacity factor. The longest operating time for pressure tubes to-date is about 120 KEFPH in Pickering Unit 4. Many lessons have been learned regarding pressure tube life limiting factors from the early CANDU units and these, together with the information obtained from an extensive pressure tube R and D program, have resulted in many design changes and improvements in material properties, mainly from manufacturing route changes. Reactors built recently are expected to achieve their 30 year design life. The development of Periodic and In-service Inspection programs and equipment, assessment methodologies and acceptance criteria, and the development of maintenance tooling and procedures are enabling the life limiting factors to be addressed in the currently operating units. The life limiting factors in currently operating Units are reviewed in relation to the experience gained from the early units, the R and D programs and the inspection and maintenance performed to date. (author)

Channel box dimension measuring method

International Nuclear Information System (INIS)

Oshima, Hirotake; Jo, Hiroto.

1994-01-01

The present invention provides a method for measuring the entire length of a channel box of a fuel assembly of a BWR type reactor. Namely, four sensors are used as one set that generate ultrasonic waves from oblique upper portion, oblique lower portion, upper portion and lower portion of the channel box respectively. The distances between the four sensors and each of the portions of the channel box are measured respectively for both of a reference member and a member to be measured. The entire length of the channel box is measured by calculating the measured values and the angles of the obliquely disposed sensors according to a predetermined formula. According to the method of the present invention, the inclination of the channel box to be measured can be corrected. In addition, accuracy of the measurement is improved and the measuring time is saved as well as the measuring device and operation can be simplified. (I.S.)

Flow-induced plastic collapse of stacked fuel plates

Energy Technology Data Exchange (ETDEWEB)

Davis, D C; Scarton, H A

1985-03-01

Flow-induced plastic collapse of stacked fuel plate assemblies was first noted in experimental reactors such as the ORNL High Flux Reactor Assembly and the Engineering Test Reactor (ETR). The ETR assembly is a stack of 19 thin flat rectangular fuel plates separated by narrow channels through which a coolant flows to remove the heat generated by fission of the fuel within the plates. The uranium alloyed plates have been noted to buckle laterally and plastically collapse at the system design coolant flow rate of 10.7 m/s, thus restricting the coolant flow through adjacent channels. A methodology and criterion are developed for predicting the plastic collapse of ETR fuel plates. The criterion is compared to some experimental results and the Miller critical velocity theory.

Fuel-element vibration and bearing pad to pressure tube fretting

International Nuclear Information System (INIS)

Fisher, N.J.; Taylor, C.E.; Pettigrew, M.J.

1990-08-01

Fuel channel operation under boiling condition results in increased flow velocities, which may lead to unacceptable fuel-element vibration and bearing pad to pressure tube fretting. The existing endurance test database does not fully cover the range of future channel operating conditions. In particular, after refuelling, some channels for future designs may operate with two-phase flow conditions outside the range of endurance test conditions. Full-scale endurance testing at realistic steam-water conditions involves substantial energy costs. Therefore, fundamental laboratory investigations were conducted to define and endurance test matrix which adequately envelops the future range of operating conditions while minimizing both the number of tests and the energy requirement of individual tests. The main focus of the laboratory investigations was to establish the relationships between: fuel channel flow conditions and fuel-element vibration; and fuel-element vibration and bearing pad to pressure tube fretting. The vibration response of a single fuel element was measured over a wide range of operating conditions covering realistic fuel channel conditions and simulated endurance testing conditions. For higher void fractions, the vibration amplitudes measured in air/water were much higher than in steam/water, while for low void fractions, the amplitudes were similar. The measured amplitudes in steam/water varied very little over the range of temperature and pressure investigated. The effects of temperature, pressure tube oxide thickness, vibration amplitude and bearing pad manufacturer on pressure tube fretting were investigated. The fretting rate is extremely temperature dependent. For vibration amplitudes about three or four times greater than expected in-reactor conditions, peak fretting rates were observed in the 225 to 286 degrees C temperature range. Fretting rates were seven times less at the higher temperatures of 300 and 315 degrees C, and the lower temperatures

Fuel assembly in a reactor

International Nuclear Information System (INIS)

Saito, Shozo; Kawahara, Akira.

1975-01-01

Object: To provide a fuel assembly in a reactor which can effectively prevent damage of the clad tube caused by mutual interference between pellets and the clad tube. Structure: A clad tube for a fuel element, which is located in the outer peripheral portion, among the fuel elements constituting fuel assemblies arranged in assembled and lattice fashion within a channel box, is increased in thickness by reducing the inside diameter thereof to be smaller than that of fuel elements internally located, thereby preventing damage of the clad tube resulting from rapid rise in output produced when control rods are removed. (Kamimura, M.)

Nuclear reactor core and fuel element therefor

International Nuclear Information System (INIS)

Fortescue, P.

1986-01-01

This patent describes a nuclear reactor core. This core consists of vertical columns of disengageable fuel elements stacked one atop another. These columns are arranged in side-by-side relationship to form a substantially continuous horizontal array. Each of the fuel elements include a block of refractory material having relatively good thermal conductivity and neutron moderating characteristics. The block has a pair of parallel flat top and bottom end faces and sides which are substantially prependicular to the end faces. The sides of each block is aligned vertically within a vertical column, with the sides of vertically adjacent blocks. Each of the blocks contains fuel chambers, including outer rows containing only fuel chambers along the sides of the block have nuclear fuel material disposed in them. The blocks also contain vertical coolant holes which are located inside the fuel chambers in the outer rows and the fuel chambers which are not located in the outer rows with the fuel chambers and which extend axially completely through from end face to end face and form continuous vertical intracolumn coolant passageways in the reactor core. The blocks have vertical grooves extending along the sides of the blocks form interblock channels which align in groups to form continuous vertical intercolumn coolant passsageways in the reactor core. The blocks are in the form of a regular hexagonal prism with each side of the block having vertical gooves defining one half of one of the coolant interblock channels, six corner edges on the blocks have vertical groves defining one-third of an interblock channel, the vertical sides of the blocks defining planar vertical surfaces

Numerical Study for Turbulent Heat Transfer in Helical Wired Sub-channel Flow Regime of Duct-less Assembly in SFR

International Nuclear Information System (INIS)

You, Byunghyun; Jeong, Yong Hoon

2014-01-01

A fuel assembly had hexagonal structure adjacent to 6 fuel assemblies, which influence to the target fuel assembly due to elimination of duct. For calculating the influence, 6 additional channels were generated between the adjacent fuel assemblies and cross flow model was applied to the channels. The adjacent fuel assemblies were analyzed and the results were used in the additional channel as boundary condition of the target fuel assembly. To design the specifications of duct-less assembly, modified or brand-new thermal-hydraulic methodology is needed which is using MATRA-LMR and CFD codes in this study. The MATRA-LMR is a sub-channel analysis code for LMR that has been developed in Korea Atomic Energy Research Institute. It is designed to analyze a fuel assembly with wire-wrap and duct structure. However, the duct-less core is not able to be analyzed by the MATRA-LMR which doesn't consider cross flow between the fuel assemblies and effect of grid spacer. The code need improvement by editing source code to eliminate effect of duct and analyze pressure drop and mixing between the sub-channels caused by grid spacer and cross flow between the fuel assemblies. To validate reformed pressure drop model and cross flow model in MATRA-LMR, CFD analysis is performed. For verifying the results of CFD, LMR subchannel experimental data is benchmarked which is done by ORNL. The verified CFD for thermalhydraulic analysis calculated pressure drop and mixing caused by grid spacer and cross flow between fuel assemblies

Effect of emergency core cooling system flow reduction on channel temperature during recirculation phase of large break loss-of-coolant accident at Wolsong unit 1

Energy Technology Data Exchange (ETDEWEB)

Yu, Seon Oh; Cho, Yong Jin [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kim, Sung Joong [Dept. of Nuclear Engineering, Hanyang University, Seoul (Korea, Republic of)

2017-08-15

The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

Spent fuel disassembly hardware and other non-fuel bearing components: characterization, disposal cost estimates, and proposed repository acceptance requirements

Energy Technology Data Exchange (ETDEWEB)

Luksic, A.T.; McKee, R.W.; Daling, P.M.; Konzek, G.J.; Ludwick, J.D.; Purcell, W.L.

1986-10-01

There are two categories of waste considered in this report. The first is the spent fuel disassembly (SFD) hardware. This consists of the hardware remaining after the fuel pins have been removed from the fuel assembly. This includes end fittings, spacer grids, water rods (BWR) or guide tubes (PWR) as appropriate, and assorted springs, fasteners, etc. The second category is other non-fuel-bearing (NFB) components the DOE has agreed to accept for disposal, such as control rods, fuel channels, etc., under Appendix E of the standard utiltiy contract (10 CFR 961). It is estimated that there will be approximately 150 kg of SFD and NFB waste per average metric ton of uranium (MTU) of spent uranium. PWR fuel accounts for approximately two-thirds of the average spent-fuel mass but only 50 kg of the SFD and NFB waste, with most of that being spent fuel disassembly hardware. BWR fuel accounts for one-third of the average spent-fuel mass and the remaining 100 kg of the waste. The relatively large contribution of waste hardware in BWR fuel, will be non-fuel-bearing components, primarily consisting of the fuel channels. Chapters are devoted to a description of spent fuel disassembly hardware and non-fuel assembly components, characterization of activated components, disposal considerations (regulatory requirements, economic analysis, and projected annual waste quantities), and proposed acceptance requirements for spent fuel disassembly hardware and other non-fuel assembly components at a geologic repository. The economic analysis indicates that there is a large incentive for volume reduction.

Fuel injection nozzle and method of manufacturing the same

Science.gov (United States)

Monaghan, James Christopher; Johnson, Thomas Edward; Ostebee, Heath Michael

2017-02-21

A fuel injection head for use in a fuel injection nozzle comprises a monolithic body portion comprising an upstream face, an opposite downstream face, and a peripheral wall extending therebetween. A plurality of pre-mix tubes are integrally formed with and extend axially through the body portion. Each of the pre-mix tubes comprises an inlet adjacent the upstream face, an outlet adjacent the downstream face, and a channel extending between the inlet and the outlet. Each pre-mix tube also includes at least one fuel injector that at least partially extends outward from an exterior surface of the pre-mix tube, wherein the fuel injector is integrally formed with the pre-mix tube and is configured to facilitate fuel flow between the body portion and the channel.

A system extinguishing a fire by insulating a liquid fuel

International Nuclear Information System (INIS)

Colome, Jacques; Duchene, Alain; Regnier, Jean.

1975-01-01

The invention refers to a system for quickly extinguishing a liquid fuel body on fire by insulating it completely from the ambient air. It applies particularly to the case of a high temperature liquid sodium sheet flowing accidentally from a circuit belonging to a fast neutron reactor. The system in question includes a lower receptacle for collecting the liquid fuel and a top cover shutting off the receptacle. This cover has inclined channels to take the liquid fuel flow and openings to allow this liquid through at the bottom end of the channels. These openings are closed by retractable shutters moving away under the pressure of the liquid in the channels and closing automatically after the liquid has flowed into the receptacle [fr

Development of the advanced CANDU technology

International Nuclear Information System (INIS)

Suk, Soo Dong; Min, Byung Joo; Na, Y. H.; Lee, S. Y.; Choi, J. H.; Lee, B. C.; Kim, S. N.; Jo, C. H.; Paik, J. S.; On, M. R.; Park, H. S.; Kim, S. R.

1997-07-01

The purpose of this study is to develop the advanced design technology to improve safety, operability and economy and to develop and advanced safety evaluation system. More realistic and reasonable methodology and modeling was employed to improve safety margin in containment analysis. Various efforts have been made to verify the CATHENA code which is the major safety analysis code for CANDU PHWR system. Fully computerized prototype ECCS was developed. The feasibility study and conceptual design of the distributed digital control system have been performed as well. The core characteristics of advanced fuel cycle, fuel management and power upgrade have been studied to determine the advanced core. (author). 77 refs., 51 tabs., 108 figs

Development of the advanced CANDU technology

Energy Technology Data Exchange (ETDEWEB)

Suk, Soo Dong; Min, Byung Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Na, Y H; Lee, S Y; Choi, J H; Lee, B C; Kim, S N; Jo, C H; Paik, J S; On, M R; Park, H S; Kim, S R [Korea Electric Power Co., Taejon (Korea, Republic of)

1997-07-01

The purpose of this study is to develop the advanced design technology to improve safety, operability and economy and to develop and advanced safety evaluation system. More realistic and reasonable methodology and modeling was employed to improve safety margin in containment analysis. Various efforts have been made to verify the CATHENA code which is the major safety analysis code for CANDU PHWR system. Fully computerized prototype ECCS was developed. The feasibility study and conceptual design of the distributed digital control system have been performed as well. The core characteristics of advanced fuel cycle, fuel management and power upgrade have been studied to determine the advanced core. (author). 77 refs., 51 tabs., 108 figs.

Enthalpy and void distributions in subchannels of PHWR fuel bundles

Energy Technology Data Exchange (ETDEWEB)

Park, J W; Choi, H; Rhee, B W [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1999-12-31

Two different types of the CANDU fuel bundles have been modeled for the ASSERT-IV code subchannel analysis. From calculated values of mixture enthalpy and void fraction distribution in the fuel bundles, it is found that net buoyancy effect is pronounced in the central region of the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. It is also found that the central region of the DUPIC fuel bundle can be cooled more efficiently than that of the standard fuel bundle. From the calculated mixture enthalpy distribution at the exit of the fuel channel, it is found that the mixture enthalpy and void fraction can be highest in the peripheral region of the DUPIC fuel bundle. On the other hand, the enthalpy and the void fraction were found to be highest in the central region of the standard CANDU fuel bundle at the exit of the fuel channel. This study shows that the subchannel analysis is very useful in assessing thermal behavior of the fuel bundle that could be used in CANDU reactors. 10 refs., 4 figs., 2 tabs. (Author)

Enthalpy and void distributions in subchannels of PHWR fuel bundles

Energy Technology Data Exchange (ETDEWEB)

Park, J. W.; Choi, H.; Rhee, B. W. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

Two different types of the CANDU fuel bundles have been modeled for the ASSERT-IV code subchannel analysis. From calculated values of mixture enthalpy and void fraction distribution in the fuel bundles, it is found that net buoyancy effect is pronounced in the central region of the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. It is also found that the central region of the DUPIC fuel bundle can be cooled more efficiently than that of the standard fuel bundle. From the calculated mixture enthalpy distribution at the exit of the fuel channel, it is found that the mixture enthalpy and void fraction can be highest in the peripheral region of the DUPIC fuel bundle. On the other hand, the enthalpy and the void fraction were found to be highest in the central region of the standard CANDU fuel bundle at the exit of the fuel channel. This study shows that the subchannel analysis is very useful in assessing thermal behavior of the fuel bundle that could be used in CANDU reactors. 10 refs., 4 figs., 2 tabs. (Author)

CARA design criteria for HWR fuel burnup extension

International Nuclear Information System (INIS)

Florido, P.C.; Cirimello, R.O.; Bergallo, J.E.; Marino, A.C.; Delmastro, D.F.; Brasnarof, D.O.; Gonzalez, J.H.; Juanico, L.A.

2002-01-01

A new concept for HWR fuel bundles, namely CARA, is presented. The CARA design allows to improve all the major performances in the PHWR fuel technology. Among others, it reaches higher burnup and thermohydraulic safety margins, together with lower fuel pellet temperatures and Zry/HM mass ratio. Moreover, it keeps the fuel mass content per unit length and the channel pressure drop by using a single diameter of fuel rods. (author)

An assessment of thermal behavior of the DUPIC fuel bundle by subchannel analysis

International Nuclear Information System (INIS)

Park, Jee Won.

1997-12-01

Thermal behavior of the standard DUPIC fuel has been assessed. The DUPIC fuel bundle has been modeled for a subchannel analysis using the ASSERT-IV code which was developed by AECL. From the calculated mixture enthalpy, equilibrium quality and void fraction distributions of the DUPIC fuel bundle, it is found that net buoyancy effect is pronounced in the central region of the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. It is also found that the central region of the DUPIC fuel bundle can be cooled more efficiently than that of the standard fuel bundle. Based upon the subchannel modeling used in this study, the location of minimum CHFR in the DUPIC fuel bundle has been found to be very similar to that of the standard fuel. From the calculated mixture enthalpy distribution at the exit of the fuel channel, it is found that the mixture enthalpy and void fraction can be highest in the peripheral region of the DUPIC fuel bundle. On the other hand, the enthalpy and the void fraction was found to be highest in the central region of the standard CANDU fuel bundle at the exit of the fuel channel. Since the transverse interchange model between subchannels is important for the behavior of these variables, it is needed to put more effort in validating the transverse interchange model. For the purpose of investigating influence of thermal-hydraulic parameter variations of the DUPIC fuel bundle, four different values of the channel flow rates were used in the subchannel analysis. The effect of the channel flow reduction on thermal-hydraulic parameters have been presented. This study shows that the subchannel analysis is very useful in assessing thermal behavior of the fuel bundles in CANDU reactors. (author). 12 refs., 3 tabs., 17 figs

The post-irradiation examination of fuel in support of Bruce A Nuclear Division fueling with flow program

International Nuclear Information System (INIS)

Montin, J.; Sagat, S.

1995-10-01

Bruce A Nuclear Division (BAND) units are operating at ∼ 75% of full power, because of the potential of a power pulse in the event of an inlet header break. As a result, BAND is converting to fueling with flow, to eliminate the potential of a power pulse and to allow for full-power operation. Concerns regarding the integrity of the end-of-life (EOL) bundles interacting with the latch at the downstream end of the fuel channel were raised. BAND carried out a test program in which EOL bundles in the upstream position of 13 of Unit 2 were cascaded into the downstream latch position 1 of another channel. Six of twelve cascaded bundles and two typical EOL position 13 (benchmark) bundles were selected for post-irradiation examination (PIE). Incipient cracks were found in the benchmark bundles. Metallographic and fractographic examination, along with crack dating, and hydrogen and deuterium analyses, indicated that the incipient cracks were the result of delayed-hydride assisted cracking at the EOL. Consequently, Ontario Hydro changed the design of the outlet shield plug to support all three rings of the fuel bundle, to minimize stress and prevent end plate cracking. Also, an ultrasonic end plate inspection tool (UT) was developed and located in the fuel bay, to inspect fuel-bundle end plates for cracks. A second test was done involving a series of four bundle cascades in BAND Unit 4 channels that had new outlet shield plugs. The latch bundles were discharged after a hot shutdown. The cascaded Unite 2 and Unit 4 latch bundles were checked for cracks using the UT. The PIE found incipient cracks or less-than-ideal welds in the assembly welds of fuel elements from Unit 2 (latch-supported fuel bundles) that had been identified by the UT as having incipient cracks. No incipient cracks were found in the assemble welds of fuel elements from Unit 4 (new outlet shield-supported fuel bundles) confirming the UT results. (author). 5 refs., 8 figs

Finite element analysis of advanced neutron source fuel plates

International Nuclear Information System (INIS)

Luttrell, C.R.

1995-08-01

The proposed design for the Advanced Neutron Source reactor core consists of closely spaced involute fuel plates. Coolant flows between the plates at high velocities. It is vital that adjacent plates do not come in contact and that the coolant channels between the plates remain open. Several scenarios that could result in problems with the fuel plates are studied. Finite element analyses are performed on fuel plates under pressure from the coolant flowing between the plates at a high velocity, under pressure because of a partial flow blockage in one of the channels, and with different temperature profiles

Hydraulic Design of the CARA Fuel Assembly for Atucha-I

International Nuclear Information System (INIS)

Juanico, Luis; Brasnarof, Daniel

2000-01-01

In this paper a hydraulic model of the CARA fuel assembly within the Atucha I fuel channel is developed. Besides, a experimental test running in the CBP low pressure loop have been designed.This model is used for design purpose of the assembly system such as the whole channel pressure drop remains the same that it is at the present.It is observed that choosing the right thickness and hole surface of the assembly system, it is possible tune up the CARA pressure drop, releases the azimuth alignment condition on the fuel element neighbors

Water removal characteristics of parallel serpentine channels

Energy Technology Data Exchange (ETDEWEB)

Jiao, K.; Zhou, B.; Quan, P. [Windsor Univ., ON (Canada). Dept. of Mechanical, Automotive and Materials Engineering

2005-07-01

A study was conducted in which the liquid water behaviours in parallel serpentine channels with manifolds on the cathode side of a proton exchange membrane (PEM) fuel cell stack were examined. A 3-dimensional, unsteady two-phase flow model within the commercial computational fluid dynamics software package FLUENT was used. Membrane electrode assemblies (MEA) were placed on different sides in the numerical analysis. Several water management issues were identified for this type of fuel cell stack by examining the flow behaviours of liquid water and airflow velocity fields. It was shown that water in the outflow manifold could be blocked by air streams from the gas flow channels, with water flowing continuously into the outflow manifold. It was also shown that the pressure drop along all the unit cells can never increase or decrease at the same pace. Water which adheres to the end wall of both the inlet and outlet manifolds is difficult to remove. It was suggested that faster water drainage can be achieved by keeping the MEA side of the gas flow channels close to the outlet of the outflow manifold. It was also suggested that the collecting and separating effect at the serpentine gas flow channels could improve the water drainage. 8 refs., 10 figs.

Effect of emergency core cooling system flow reduction on channel temperature during recirculation phase of large break loss-of-coolant accident at Wolsong unit 1

Directory of Open Access Journals (Sweden)

Seon Oh Yu

2017-08-01

Full Text Available The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

Core analysis during transition from 37-element fuel to CANFLEX-NU fuel in CANDU 6

Energy Technology Data Exchange (ETDEWEB)

Jeong, Chang Joon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1999-12-31

An 1200-day time-dependent fuel-management for the transition from 37-element fuel to CANFLEX-NU fuel in a CANDU 6 reactor has been simulated to show the compatibility of the CANFLEX-NU fuel with the reactor operation. The simulation calculations were carried out with the RFSP code, provided by cell averaged fuel properties obtained from the POWDERPUFS-V code. The refueling scheme for both fuels was an eight bundle shift at a time. The simulation results show that the maximum channel and bundle powers were maintained below the license limit of the CANDU 6. This indicates that the CANFLEX-NU fuel bundle is compatible with the CANDU 6 reactor operation during the transition period. 3 refs., 2 figs., 1 tab. (Author)

Core analysis during transition from 37-element fuel to CANFLEX-NU fuel in CANDU 6

Energy Technology Data Exchange (ETDEWEB)

Jeong, Chang Joon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

An 1200-day time-dependent fuel-management for the transition from 37-element fuel to CANFLEX-NU fuel in a CANDU 6 reactor has been simulated to show the compatibility of the CANFLEX-NU fuel with the reactor operation. The simulation calculations were carried out with the RFSP code, provided by cell averaged fuel properties obtained from the POWDERPUFS-V code. The refueling scheme for both fuels was an eight bundle shift at a time. The simulation results show that the maximum channel and bundle powers were maintained below the license limit of the CANDU 6. This indicates that the CANFLEX-NU fuel bundle is compatible with the CANDU 6 reactor operation during the transition period. 3 refs., 2 figs., 1 tab. (Author)

Reactor core with rod-shaped fuel cells

International Nuclear Information System (INIS)

Dworak, A.

1977-01-01

The aim is an optimization of load distribution in the core so that the load decreases in the direction of coolant flow (with gas cooling from above downwards) but so that it remains constant in horizontal layers to the edge of the core. The former produces optimum cooling, because the coolant has to take up decreasing heat output in the direction of flow. The latter simplifies refueling, because replacement of a whole layer having the same burn-up takes place. The upper two layers with the highest output and the shortest dwell time are replaced every 300 days, for example, the third layer is replaced after double this time and 5 more layers after four times this dwell time. After the simultaneous replacement of all layers, the reactor is in the same state as at commissioning. The fuel cells consist of hexagonal graphite blocks about 1.65 metres in height and 0.75 wide, for example. Each block contains about 100 through cooling channels and about 200 fuel channels closed on both sides. A large number of columns each consisting of 8 blocks is arranged in a tight honeycomb pattern and forms the core. Within each of the 8 horizontal layers of blocks, each fuel cell contains the same fuel mixture with predetermined dwell time. The fuel mixture is suited to the dwell time planned for each layer. The various fuel cells are kept at the same output by burnable neutron poisons in special channels provided for this purpose in the fuel cell and/or by absorber rods, or a planned load distribution is maintained. (HP) [de

Nuclear fuel assembly

International Nuclear Information System (INIS)

Hirano, Yasushi; Hirukawa, Koji; Sakurada, Koichi.

1994-01-01

A bundle of fuel rods is divided into four fuel rod group regions of small fuel rod bundles by a cross-shaped partitioning structure consisting of paired plate-like structures which connect two opposing surfaces of a channel box. A water removing material with less neutron absorption (for example, Zr or a Zr alloy) or a solid moderator is inserted and secured to a portion of a non-boiling water region interposed between the paired plate-like structure. It has a structure that light water flows to the region in the plate-like structure. The volume, density or composition of the water removing material is controlled depending on the composition of the fuels, to change the moderating characteristics of neutrons in the non-boiling water region. This can easily moderate the difference of nuclear characteristics between each of fuel assemblies using fuel materials of different fuel compositions. Further, the reactivity control effect of the burnable poisons can be enhanced without worsening fuel economy or linear power density. (I.N.)

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.

Experimental investigation of critical velocity in a parallel plate research reactor fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Castro, Alfredo J.A.; Scuro, Nikolas L.; Andrade, Delvonei A., E-mail: ajcastro@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNE-SP), Sao Paulo, SP (Brazil)

2017-07-01

The fuel elements of a MTR (Material Testing Reactor) type nuclear reactor are mostly composed of aluminum coated fuel plates containing the core of uranium silica (U{sub 3}Si{sub 2}) dispersed in an aluminum matrix. These plates have a thickness of the order of millimeters and are much longer in relation to their thickness. They are arranged in parallel in the assembly of the fuel element to form channels between them a few millimeters in thickness, through which there is a flow of the coolant. This configuration, combined with the need for a flow at high flow rates to ensure the cooling of the fuel element in operation, may create problems of mechanical failure of fuel plate due to the vibration induced by the flow in the channels. In the case of critical velocity excessive permanent deflections of the plates can cause blockage of the flow channel in the reactor core and lead to overheating in the plates. For this study an experimental bench capable of high volume flows and a test section that simulates a plate-like fuel element with three cooling channels were developed. The dimensions of the test section were based on the dimensions of the Fuel Element of the Brazilian Multipurpose Reactor (RMB), whose project is being coordinated by the National Commission of Nuclear Energy (CNEN). The experiments performed attained the objective of reaching Miller's critical velocity condition. The critical velocity was reached with 14.5 m/s leading to the consequent plastic deformation of the flow channel plates. (author)

Composition heterogeneity analysis for DUPIC fuel(I) - Statistical analysis

Energy Technology Data Exchange (ETDEWEB)

Choi, Hang Bok [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-08-01

The fuel composition heterogeneity effect on reactor performance parameters was assessed by refueling simulations for three DUPIC fuel options of fuel composition heterogeneity control: the fissile content adjustment, the reactivity control by slightly enriched and depleted uranium, and the reactivity control by natural uranium. For each DUPIC fuel option, the simulations were performed using 30 heterogeneous fuel types which were determined by the agglomerative hierarchical clustering method. The heterogeneity effect was considered during the refueling simulation by randomly selecting fuel types for the refueling operation. The refueling simulations of the heterogeneous core have shown that the key performance parameters such as the maximum channel power (MCP), maximum bundle power (MBP), and channel power peaking factor (CPPF) are close to those of the core that has single fuel type. For the three DUPIC fuel options, the uncertainties of MCP, MBP, and CPPF due to the fuel composition heterogeneity are less than 0.6, 1.5 and 0.8%, respectively, including the uncertainty of the group-average fuel property. This study has shown that the three DUPIC fuel options reduces the composition heterogeneity effectively and the zone power control system has a sufficient margin to adjust the perturbations cased by the fuel composition heterogeneity. 15 refs., 28 figs.,10 tabs. (Author)

Water removal characteristics of parallel serpentine channels. Paper no. IGEC-1-035

International Nuclear Information System (INIS)

Jiao, K.; Zhou, B.; Quan, P.

2005-01-01

Water management in a proton exchange membrane (PEM) fuel cell stack has been a challenging issue on the road to commercialization. This paper presents a numerical investigation of air-water flow in parallel serpentine channels on cathode side of a PEM fuel cell stack by use of the commercial Computational Fluid Dynamics (CFD) software package FLUENT. Different air-water flow behaviours inside the serpentine flow channels with inlet and outlet manifolds were discussed. The results showed that there were significant variations of water distribution and pressure drop in different cells at different times. The 'collecting-and-separating effect' due to the serpentine shape of the gas flow channels, the pressure drop change due to the water distribution inside the outlet manifold were observed. Several gas flow problems of this type of parallel serpentine channels were identified and useful suggestions were given through investigating the flow patterns inside the channels and manifolds. (author)

Experimental Study and Comparison of Various Designs of Gas Flow Fields to PEM Fuel Cells and Cell Stack Performance

International Nuclear Information System (INIS)

Liu, Hong; Li, Peiwen; Juarez-Robles, Daniel; Wang, Kai; Hernandez-Guerrero, Abel

2014-01-01

In this study, a significant number of experimental tests to proton exchange membrane (PEM) fuel cells were conducted to investigate the effect of gas flow fields on fuel cell performance. Graphite plates with various flow field or flow channel designs, from literature survey and also novel designs by the authors, were used for the PEM fuel cell assembly. The fabricated fuel cells have an effective membrane area of 23.5 cm 2 . The results showed that the serpentine flow channel design is still favorable, giving the best single fuel cell performance amongst all the studied flow channel designs. A novel symmetric serpentine flow field was proposed for a relatively large sized fuel cell application. Four fuel cell stacks each including four cells were assembled using different designs of serpentine flow channels. The output power performances of fuel cell stacks were compared and the novel symmetric serpentine flow field design is recommended for its very good performance.

Metabolism regulates the spontaneous firing of substantia nigra pars reticulata neurons via KATP and nonselective cation channels.

Science.gov (United States)

Lutas, Andrew; Birnbaumer, Lutz; Yellen, Gary

2014-12-03

Neurons use glucose to fuel glycolysis and provide substrates for mitochondrial respiration, but neurons can also use alternative fuels that bypass glycolysis and feed directly into mitochondria. To determine whether neuronal pacemaking depends on active glucose metabolism, we switched the metabolic fuel from glucose to alternative fuels, lactate or β-hydroxybutyrate, while monitoring the spontaneous firing of GABAergic neurons in mouse substantia nigra pars reticulata (SNr) brain slices. We found that alternative fuels, in the absence of glucose, sustained SNr spontaneous firing at basal rates, but glycolysis may still be supported by glycogen in the absence of glucose. To prevent any glycogen-fueled glycolysis, we directly inhibited glycolysis using either 2-deoxyglucose or iodoacetic acid. Inhibiting glycolysis in the presence of alternative fuels lowered SNr firing to a slower sustained firing rate. Surprisingly, we found that the decrease in SNr firing was not mediated by ATP-sensitive potassium (KATP) channel activity, but if we lowered the perfusion flow rate or omitted the alternative fuel, KATP channels were activated and could silence SNr firing. The KATP-independent slowing of SNr firing that occurred with glycolytic inhibition in the presence of alternative fuels was consistent with a decrease in a nonselective cationic conductance. Although mitochondrial metabolism alone can prevent severe energy deprivation and KATP channel activation in SNr neurons, active glucose metabolism appears important for keeping open a class of ion channels that is crucial for the high spontaneous firing rate of SNr neurons. Copyright © 2014 the authors 0270-6474/14/3416336-12$15.00/0.

The velocity measurement by LDV at the simulated plate fuel assembly

International Nuclear Information System (INIS)

Tae Sung Ha

2001-01-01

For a more accurate safety analysis for McMaster Nuclear Reactor (MNR), local velocity measurements in a mock-up of the 18-plate fuel assembly are conducted over the range of M=2.0kg/s to 5.0kg/s (u=0.59m/s to 1.48m/s). To enable the measurement of the mass flow distribution through the channels by Laser Doppler Velocimeter(LDV), the curved fuel plate assembly is modified to flat fuel plates. The experimental result shows that the velocity profile is fairly symmetric for the 1st channel to the 17th subchannel at its center. The velocity in the peripheral area is slightly decreased while that directly above the circular pipe is correspondingly increased due to the effect of blockage by the exit endfitting. The mass flow rate fraction is fairly well distributed from the 1st to the 9th channels; at the outmost channels (1st and 3rd subchannels) the flow is approximately 95-97% of the average channel flow and at the central channels (4th and 8th subchannels) the flow is about 102-105% of the average channel mass flow rate. It is shown that the measured mass flow distribution is consistent with the results of the numerical calculation except 1st and 17th channels. (author)

The post irradiation examination of fuel in support of Bruce A nuclear division fueling with flow program

International Nuclear Information System (INIS)

Montin, J.; Sagat, S.; Day, R.; Novak, J.; Bromfield, H.

1995-01-01

Bruce A Nuclear Division (BAND) units are operating at ∼ 75% of full power, because of the potential of a power pulse in the event of an inlet header break. As a result, BAND is converting to fueling with flow, to eliminate the potential of a power pulse and to allow for full-power operation. Concerns regarding the integrity of the end-of-life (EOL) bundles interacting with the latch at the downstream end of the fuel channel were raised. BAND carried out a test program in which EOL bundles in the upstream position 13 of Unit 2 were cascaded into the downstream latch position 1 of another channel. Six of twelve cascaded bundles and two typical EOL position 13 (benchmark) bundles were selected for post-irradiation examination (PIE). Incipient cracks were found in the assembly welds (endplateto-endcap welds) of all six cascaded bundles. No incipient cracks were found in the benchmark bundles. Metallographic and fractographic examination, along with crack dating, and hydrogen and deuterium analyses, indicated that the incipient cracks were the result of delayed-hydride assisted cracking at the EOL. Consequently, Ontario Hydro changed the design of the outlet shield plug to support all three rings of the fuel bundle, to minimize stress and prevent endplate cracking. Also, an ultrasonic endplate inspection tool (UT) was developed and located in the fuel bay. to inspect fuelbundle endplates for cracks. A second test was done involving a series of four bundle cascades in BAND Unit 4 channels that had new outlet shield plugs. The latch bundles were discharged after a hot shutdown. The cascaded Unit 2 and Unit 4 latch bundles were checked for cracks using the UT. The PIE found incipient cracks or less-than-ideal welds in the assembly welds of fuel elements from Unit 2 (latch-supported fuel bundles) that had been identified by the UT as having incipient cracks. No incipient cracks were found in the assembly welds of fuel elements from Unit 4 (new outlet shield

Air-cooled, hydrogen-air fuel cell

Science.gov (United States)

Shelekhin, Alexander B. (Inventor); Bushnell, Calvin L. (Inventor); Pien, Michael S. (Inventor)

1999-01-01

An air-cooled, hydrogen-air solid polymer electrolyte (SPE) fuel cell with a membrane electrode assembly operatively associated with a fluid flow plate having at least one plate cooling channel extending through the plate and at least one air distribution hole extending from a surface of the cathode flow field into the plate cooling channel.

Economic and environmental considerations of biomass fuels

International Nuclear Information System (INIS)

Booth, Roger

1992-01-01

The economic and environmental aspects of biomass fuels are considered. Close to source, the cost of useful energy in the form of lignocellulose is often competitive with fossil fuels, say $1-3 per GJ. There are three main options to divert this biomass into commercial energy channels: solid fuels for underboiler use; liquid fuels for automotive use and electric power generation, each of which is discussed. The social, economic and environmental advantages of an afforestation programme are highlighted. (Author)

Prediction for the flow distribution and the pressure drop of a plate type fuel assembly

International Nuclear Information System (INIS)

Park, Jong Hark; Jo, Dea Sung; Chae, Hee Taek; Lee, Byung Chul

2011-01-01

A plate type fuel assembly widely used in many research reactors does not allow the coolant to mix with neighboring fuel channels due to the completely separated flow channels. If there is a serious inequality of coolant distribution among channels, it can reduce thermal-hydraulic safety margin, as well as it can cause a deformation of fuel plates by the pressure difference between neighboring channels, thus the flow uniformity in the fuel assembly should be confirmed. When designing a primary cooling system (PCS), the pressure drop through a reactor core is a dominant value to determine the PCS pump size. The major portion of reactor core pressure drop is caused by the fuel assemblies. However it is not easy to get a reasonable estimation of pressure drop due to the geometric complexity of the fuel assembly and the thin gaps between fuel assemblies. The flow rate through the gap is important part to determine the total flow rate of PCS, so it should be estimated as reasonable as possible. It requires complex and difficult jobs to get useful data. In this study CFD analysis to predict the flow distribution and the pressure drop were conducted on the plate type fuel assembly, which results would be used to be preliminary data to determine the PCS flow rate and to improve the design of a fuel assembly

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)

Economic analyses of alpha channeling in tokamak power plants

International Nuclear Information System (INIS)

Ehst, D.A.

1998-01-01

The hot-ion-mode of operation [1] has long been thought to offer optimized performance for long-pulse or steady-state magnetic fusion power plants. This concept was revived in recent years when theoretical considerations suggested that nonthermal fusion alpha particles could be made to channel their power density preferentially to the fuel ions [2,3]. This so-called anomalous alpha particle slowing down can create plasmas with fuel ion temperate T i somewhat larger than the electron temperature T e , which puts more of the beta-limited plasma pressure into the useful fuel species (rather than non-reacting electrons). As we show here, this perceived benefit may be negligible or nonexistent for tokamaks with steady state current drive. It has likewise been argued [2,3] that alpha channeling could be arranged such that little or no external power would be needed to generate the steady state toroidal current. Under optimistic assumptions we show that such alpha-channeling current drive would moderately improve the economic performance of a first stability tokamak like ARIES-I [4], however a reversed-shear (advanced equilibrium) tokamak would likely not benefit since traditional radio-wave (rf) electron-heating current drive power would already be quite small

Impact of industrial nuclear releases into the English Channel

International Nuclear Information System (INIS)

Germain, P.; Guegueniat, P.

1992-01-01

The nuclear fuel reprocessing plant at La Hague is the main source of releases of weakly radioactive waste into the English Channel; there are also some contributions from nuclear power stations along the coast. Indicator species, seawater samples and sediments are used to study the distribution and transfer mechanisms of radionuclides in Channel waters. The observed pattern of radiolabelled zones is in good agreement with an hydrodynamic model for the Channel. The variations of activity with time are discussed in relation to releases from La Hague

The PACo channel strategy; La strategie du reseau PACo

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

The PACo channel was created in june 1999 to answer government demands in the domain of new energies, to stimulate the new technologies development and to participate to enterprises growth. The first year was devoted in particular to the fuel cell and the hydrogen fuel. This document presents an analysis of the first year of program management. (A.L.B.)

Safety analysis of JMTR LEU fuel core, (3)

International Nuclear Information System (INIS)

Tsuchida, Noboru; Shiraishi, Tadao; Takahashi, Yutaka; Inada, Seiji; Saito, Minoru; Futamura, Yoshiaki; Kitano, Kyoshiro.

1992-10-01

Dose analysis in the safety evaluation and the site evaluation were performed for the JMTR core conversion from MEU fuel to LEU fuel. In the safety evaluation, the effective dose equivalents for the public surrounding the site were estimated in fuel handling accident and flow blockage to coolant channel which were selected as the design basis accidents with release of radioactive fission products to the environment. In the site evaluation, the flow blockage to coolant channel was selected as siting basis events, since this accident had the possibility of spreading radioactive release. Maximum exposure doses for the public were estimated assuming large amounts of fission products to release. It was confirmed that risk of radiation exposure of the public is negligible and the siting is appropriate. (author)

Nuclear reactor coolant channels

International Nuclear Information System (INIS)

Macbeth, R.V.

1978-01-01

A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

Nuclear reactor fuel assembly

International Nuclear Information System (INIS)

Sakurai, Shungo; Ogiya, Shunsuke.

1990-01-01

In a fuel assembly, if the entire fuels comprise mixed oxide fuels, reactivity change in cold temperature-power operation is increased to worsen the reactor shutdown margin. The reactor shutdown margin has been improved by increasing the burnable poison concentration thereby reducing the reactivity of the fuel assembly. However, since unburnt poisons are present at the completion of the reactor operation, the reactivity can not be utilized effectively to bring about economical disadvantage. In view of the above, the reactivity change between lower temperature-power operations is reduced by providing a non-boiling range with more than 9.1% of cross sectional area at the inside of a channel at the central portion of the fuel assembly. As a result, the amount of the unburnt burnable poisons is decreased, the economy of fuel assembly is improved and the reactor shutdown margin can be increase. (N.H.)

Wood fuels sources and markets

International Nuclear Information System (INIS)

Koopmans, Auke

2003-01-01

Biomass energy is an important source of energy in most Asian countries. Households and industries use substantial amounts of fuel wood, charcoal and other biomass energy, such as agricultural residues, dung, leaves and sawmill residues. The main household applications are cooking and heating whereas industrial applications range widely. This paper provides an overview of estimates on the production and trade of biomass fuels in the South-east Asia region. The flows and channels used in the supply of wood fuels in different countries were analysed. This paper may help in identifying policy gaps with regards to the supply and consumption of wood fuels from both forest and non-forest sources. (Author)

An optimum fuel management method based on CANDU in-core detector readings

International Nuclear Information System (INIS)

Jeong, Chang Joon; Choi, Hang Bok

2001-01-01

In this study, a new optimal fuel management method is developed for a CANDU 600 MWe (CANDU-6) reactor. At first, an efficient power mapping method has been developed, which provides an accurate core status of an operating CANDU reactor. Secondly, an optimum refueling channel selection method has been developed by an optimization theory. For the power mapping method, the measured detector readings are used as boundary conditions of the diffusion theory calculation with the Kalman filtering (DIKAL) method. The performance of the DIKAL method was assessed for various core states and applied to the calculation of power and flux distribution in the CANDU 6 reactor. Sensitivity studies have shown that DIKAL method is insensitive to the detector random and systematic errors. An optimal refueling simulation method (OPTIMA), practically applicable to a CANDU 6 reactor, has also been developed. The objective of the optimization is to reproduce the reference core performance during refueling simulation, while satisfying the operation limits of channel and bundle powers. The optimization process consists of two stages: i) elimination of candidate refueling channels by several constraints and ii) selection of refueling channels by a direct search method that uses sensitivity coefficients of channel power generated for the reference core. The elimination process sorts out an appropriate number of fuel channels suitable for refueling, considering the channel power, bundle power and fuel burnup. The optimum refueling channels are then selected such that the difference of power distribution from the reference is minimized. In order to demonstrate the applicability of the overall fuel management methodology developed in this study, the DIKAL-OPTIMA method was applied to Wolsong-3 reactor refueling simulation, which is a typical CANDU-6 reactor. The results of refueling simulation have shown that the method can be efficiently used for the performance analysis of the operating

An optimum fuel management method based on CANDU in-core detector readings

Energy Technology Data Exchange (ETDEWEB)

Jeong, Chang Joon; Choi, Hang Bok

2001-01-01

In this study, a new optimal fuel management method is developed for a CANDU 600 MWe (CANDU-6) reactor. At first, an efficient power mapping method has been developed, which provides an accurate core status of an operating CANDU reactor. Secondly, an optimum refueling channel selection method has been developed by an optimization theory. For the power mapping method, the measured detector readings are used as boundary conditions of the diffusion theory calculation with the Kalman filtering (DIKAL) method. The performance of the DIKAL method was assessed for various core states and applied to the calculation of power and flux distribution in the CANDU 6 reactor. Sensitivity studies have shown that DIKAL method is insensitive to the detector random and systematic errors. An optimal refueling simulation method (OPTIMA), practically applicable to a CANDU 6 reactor, has also been developed. The objective of the optimization is to reproduce the reference core performance during refueling simulation, while satisfying the operation limits of channel and bundle powers. The optimization process consists of two stages: i) elimination of candidate refueling channels by several constraints and ii) selection of refueling channels by a direct search method that uses sensitivity coefficients of channel power generated for the reference core. The elimination process sorts out an appropriate number of fuel channels suitable for refueling, considering the channel power, bundle power and fuel burnup. The optimum refueling channels are then selected such that the difference of power distribution from the reference is minimized. In order to demonstrate the applicability of the overall fuel management methodology developed in this study, the DIKAL-OPTIMA method was applied to Wolsong-3 reactor refueling simulation, which is a typical CANDU-6 reactor. The results of refueling simulation have shown that the method can be efficiently used for the performance analysis of the operating

Analyses of subchannel velocity distribution for HANARO fuel assembly

International Nuclear Information System (INIS)

Chae, Hee Taek; Han, Gee Yang; Park, Cheol; Lim, In Cheol

1998-10-01

MATRA-h which is a subchannel analysis computer code is used to evaluate the thermal margin of HANARO core. To estimate core thermal margin, accurate prediction of subchannel velocity is very important. The average subchannel velocities of 18 element fuel assembly were obtained from the results of velocity measurement test. To validate the adequacy of the hydraulic model code predictions were compared with the experimental results for the subchannel velocity distribution in 18 element fuel channel. The calculated subchannel velocity distributions in the central channels were larger than those of experiment. On the other hand the subchannel velocities in the outer channels were smaller. It is speculated that the prediction like as above would make CHF value lower because CHF phenomena had been occurred in the outer fuel element in the bundle CHF test of AECL. The prediction for axial pressure distribution coincided with the experimental results well. (author). 9 refs., 9 tabs., 14 figs

Hydraulic Profiling of a Parallel Channel Type Reactor Core

International Nuclear Information System (INIS)

Seo, Kyong-Won; Hwang, Dae-Hyun; Lee, Chung-Chan

2006-01-01

An advanced reactor core which consisted of closed multiple parallel channels was optimized to maximize the thermal margin of the core. The closed multiple parallel channel configurations have different characteristics to the open channels of conventional PWRs. The channels, usually assemblies, are isolated hydraulically from each other and there is no cross flow between channels. The distribution of inlet flow rate between channels is a very important design parameter in the core because distribution of inlet flow is directly proportional to a margin for a certain hydraulic parameter. The thermal hydraulic parameter may be the boiling margin, maximum fuel temperature, and critical heat flux. The inlet flow distribution of the core was optimized for the boiling margins by grouping the inlet orifices by several hydraulic regions. The procedure is called a hydraulic profiling

Schemes for fuel conservation for PHWRs due for complete fuel discharge

International Nuclear Information System (INIS)

Bansal, Ravi; Kumar, Deepak; Tejram

2009-01-01

Narora Atomic Power Station (NAPS) consists of twin units of pressurized heavy water reactors (PHWR) using natural uranium as fuel and heavy water as moderator and coolant. On-power bi-directional refueling is employed at NAPS. En Masse Coolant Channel Replacement (EMCCR) necessitates the low burn-up bundles present in core to be utilized. The different schemes of In-core fuel management viz. internal, total internal and external recycling were worked out to utilize these low burn-up bundles. This led to saving of: (a) 2011 natural uranium bundles at NAPS and (b) 4 and half months in NAPS-1 and 3 and half months in case of NAPS-2 in core de-fueling time. (author)

A comprehensive, consistent and systematic mathematical model of PEM fuel cells

International Nuclear Information System (INIS)

Baschuk, J.J.; Li Xianguo

2009-01-01

This paper presents a comprehensive, consistent and systematic mathematical model for PEM fuel cells that can be used as the general formulation for the simulation and analysis of PEM fuel cells. As an illustration, the model is applied to an isothermal, steady state, two-dimensional PEM fuel cell. Water is assumed to be in either the gas phase or as a liquid phase in the pores of the polymer electrolyte. The model includes the transport of gas in the gas flow channels, electrode backing and catalyst layers; the transport of water and hydronium in the polymer electrolyte of the catalyst and polymer electrolyte layers; and the transport of electrical current in the solid phase. Water and ion transport in the polymer electrolyte was modeled using the generalized Stefan-Maxwell equations, based on non-equilibrium thermodynamics. Model simulations show that the bulk, convective gas velocity facilitates hydrogen transport from the gas flow channels to the anode catalyst layers, but inhibits oxygen transport. While some of the water required by the anode is supplied by the water produced in the cathode, the majority of water must be supplied by the anode gas phase, making operation with fully humidified reactants necessary. The length of the gas flow channel has a significant effect on the current production of the PEM fuel cell, with a longer channel length having a lower performance relative to a shorter channel length. This lower performance is caused by a greater variation in water content within the longer channel length

THESEE-3, Orgel Reactor Performance and Statistic Hot Channel Factors

International Nuclear Information System (INIS)

Chambaud, B.

1974-01-01

1 - Nature of physical problem solved: The code applies to a heavy-water moderated organic-cooled reactor channel. Different fuel cluster models can be used (circular or hexagonal patterns). The code gives coolant temperatures and velocities and cladding temperatures throughout the channel and also channel performances, such as power, outlet temperature, boiling and burn-out safety margins (see THESEE-1). In a further step, calculations are performed with statistical values obtained by random retrieval of geometrical in- put data and taking into account construction tolerances, vibrations, etc. The code evaluates the mean value and standard deviation for the more important thermal and hydraulic parameters. 2 - Method of solution: First step calculations are performed for nominal values of parameters by solving iteratively the non-linear system of equations which give the pressure drops in subchannels of the current zone (see THESEE-1). Then a Gaussian probability distribution of possible statistical values of the geometrical input data is assumed. A random number generation routine determines the statistical case. Calculations are performed in the same way as for the nominal case. In the case of several channels, statistical performances must be adjusted to equalize the normal pressure drop. A special subroutine (AVERAGE) then determines the mean value and standard deviation, and thus probability functions of the most significant thermal and hydraulic results. 3 - Restrictions on the complexity of the problem: Maximum 7 fuel clusters, each divided into 10 axial zones. Fuel bundle geometries are restricted to the following models - circular pattern 6/7, 18/19, 36/67 rods, with or without fillers. The fuel temperature distribution is not studied. The probability distribution of the statistical input is assumed to be a Gaussian function. The principle of random retrieval of statistical values is correct, but some additional correlations could be found from a more

Fuel assemblies for BWR type reactors

International Nuclear Information System (INIS)

Ishizuka, Takao.

1981-01-01

Purpose: To enable effective failed fuel detection by the provision of water rod formed with a connecting section connected to a warmed water feed pipe of a sipping device at the lower portion and with a warmed water jetting port in the lower portion in a fuel assembly of a BWR type reactor to thereby carry out rapid sipping. Constitution: Fuel rods and water rods are contained in the channel box of a fuel assembly, and the water rod is provided at its upper portion with a connecting section connected to the warmed water feed pipe of the sipping device and formed at its lower portion with a warmed water jetting port for jetting warmed water fed from the warmed water feed pipe. Upon detection of failed fuels, the reactor operation is shut down and the reactor core is immersed in water. The cover for the reactor container is removed and the cap of the sipping device is inserted to connect the warmed water feed pipe to the connecting section of the water rod. Then, warmed water is fed to the water rod and jetted out from the warmed water jetting port to cause convection and unify the water of the channel box in a short time. Thereafter, specimen is sampled and analyzed for the detection of failed fuels. (Moriyama, K.)

Comparison of fuel assemblies in lead cooled fast reactors

Energy Technology Data Exchange (ETDEWEB)

Perez, A.; Sanchez, H.; Aguilar, L.; Espinosa P, G., E-mail: alejandria.peval@gmail.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico)

2016-09-15

This paper presents a comparison of the thermal-fluid processes in the core, fuel heat transfer, and thermal power between two fuel assemblies: square and hexagonal, in a lead-cooled fast reactor (Lfr). A multi-physics reduced order model for the analysis of Lfr single channel is developed in this work. The work focused on a coupling between process of neutron kinetic, fuel heat transfer process and thermal-fluid, in a single channel. The thermal power is obtained from neutron point kinetics model, considering a non-uniform power distribution. The analysis of the processes of thermal-fluid considers thermal expansion effects. The transient heat transfer in fuel is carried out in an annular geometry, and one-dimensional in radial direction for each axial node. The results presented in comparing these assemblies consider the temperature field in the fuel, in the thermal fluid and under steady state, and transient conditions. Transients consider flow of coolant and inlet temperature of coolant. The mathematical model of Lfr considers three main modules: the heat transfer in the annular fuel, the power generation with feedback effects on neutronic, and the thermal-fluid in the single channel. The modeling of nuclear reactors in general, the coupling is crucial by the feedback between the neutron processes with fuel heat transfer, and thermo-fluid, where is very common the numerical instabilities, after all it has to refine the model to achieve the design data. In this work is considered as a reference the ELSY reactor for the heat transfer analysis in the fuel and pure lead properties for analyzing the thermal-fluid. The results found shows that the hexagonal array has highest temperature in the fuel, respect to square array. (Author)

Comparison of fuel assemblies in lead cooled fast reactors

International Nuclear Information System (INIS)

Perez, A.; Sanchez, H.; Aguilar, L.; Espinosa P, G.

2016-09-01

This paper presents a comparison of the thermal-fluid processes in the core, fuel heat transfer, and thermal power between two fuel assemblies: square and hexagonal, in a lead-cooled fast reactor (Lfr). A multi-physics reduced order model for the analysis of Lfr single channel is developed in this work. The work focused on a coupling between process of neutron kinetic, fuel heat transfer process and thermal-fluid, in a single channel. The thermal power is obtained from neutron point kinetics model, considering a non-uniform power distribution. The analysis of the processes of thermal-fluid considers thermal expansion effects. The transient heat transfer in fuel is carried out in an annular geometry, and one-dimensional in radial direction for each axial node. The results presented in comparing these assemblies consider the temperature field in the fuel, in the thermal fluid and under steady state, and transient conditions. Transients consider flow of coolant and inlet temperature of coolant. The mathematical model of Lfr considers three main modules: the heat transfer in the annular fuel, the power generation with feedback effects on neutronic, and the thermal-fluid in the single channel. The modeling of nuclear reactors in general, the coupling is crucial by the feedback between the neutron processes with fuel heat transfer, and thermo-fluid, where is very common the numerical instabilities, after all it has to refine the model to achieve the design data. In this work is considered as a reference the ELSY reactor for the heat transfer analysis in the fuel and pure lead properties for analyzing the thermal-fluid. The results found shows that the hexagonal array has highest temperature in the fuel, respect to square array. (Author)

Nuclear reactor fuel assembly

International Nuclear Information System (INIS)

Vikhorev, Yu.V.; Biryukov, G.I.; Kirilyuk, N.A.; Lobanov, V.N.

1977-01-01

A fuel assembly is proposed for nuclear reactors allowing remote replacement of control rod bundles or their shifting from one assembly to another, i.e., their multipurpose use. This leads to a significant increase in fuel assembly usability. In the fuel assembly the control rod bundle is placed in guide tube channels to which baffles are attached for fuel element spacing. The remote handling of control rods is provided by a hollow cylinder with openings in its lower bottom through which the control rods pass. All control rods in a bundle are mounted to a cross beam which in turn is mounted in the cylinder and is designed for grasping the whole rod bundle by a remotely controlled telescopic mechanism in bundle replacement or shifting. (Z.M.)

On the application of the PFEM to droplet dynamics modeling in fuel cells

Science.gov (United States)

Ryzhakov, Pavel B.; Jarauta, Alex; Secanell, Marc; Pons-Prats, Jordi

2017-07-01

The Particle Finite Element Method (PFEM) is used to develop a model to study two-phase flow in fuel cell gas channels. First, the PFEM is used to develop the model of free and sessile droplets. The droplet model is then coupled to an Eulerian, fixed-grid, model for the airflow. The resulting coupled PFEM-Eulerian algorithm is used to study droplet oscillations in an air flow and droplet growth in a low-temperature fuel cell gas channel. Numerical results show good agreement with predicted frequencies of oscillation, contact angle, and deformation of injected droplets in gas channels. The PFEM-based approach provides a novel strategy to study droplet dynamics in fuel cells.

Thermal performance of a meso-scale liquid-fuel combustor

International Nuclear Information System (INIS)

Vijayan, V.; Gupta, A.K.

2011-01-01

Research highlights: → Demonstrated successful combustion of liquid fuel-air mixtures in a novel meso-scale combustor. → Flame quenching was eliminated using heat recirculation in a swiss roll type combustor that also extended the flammability limits. → Liquid fuel was rapidly vaporized with the use of hot narrow channel walls that eliminated the need of a fuel atomizer. → Maximum power density of the combustor was estimated to be about 8.5 GW/m3 and heat load in the range of 50-280W. → Overall efficiency of the combustor was estimated in the range of 12 to 20%. - Abstract: Combustion in small scale devices poses significant challenges due to the quenching of reactions from wall heat losses as well as the significantly reduced time available for mixing and combustion. In the case of liquid fuels there are additional challenges related to atomization, vaporization and mixing with the oxidant in the very short time-scale liquid-fuel combustor. The liquid fuel employed here is methanol with air as the oxidizer. The combustor was designed based on the heat recirculating concept wherein the incoming reactants are preheated by the combustion products through heat exchange occurring via combustor walls. The combustor was fabricated from Zirconium phosphate, a ceramic with very low thermal conductivity (0.8 W m -1 K -1 ). The combustor had rectangular shaped double spiral geometry with combustion chamber in the center of the spiral formed by inlet and exhaust channels. Methanol and air were introduced immediately upstream at inlet of the combustor. The preheated walls of the inlet channel also act as a pre-vaporizer for liquid fuel which vaporizes the liquid fuel and then mixes with air prior to the fuel-air mixture reaching the combustion chamber. Rapid pre-vaporization of the liquid fuel by the hot narrow channel walls eliminated the necessity for a fuel atomizer. Self-sustained combustion of methanol-air was achieved in a chamber volume as small as 32.6 mm 3

Two-phase flow instabilities in a vertical annular channel

Energy Technology Data Exchange (ETDEWEB)

Babelli, I.; Nair, S.; Ishii, M. [Purdue Univ., West Lafayette, IN (United States)

1995-09-01

An experimental test facility was built to study two-phase flow instabilities in vertical annular channel with emphasis on downward flow under low pressure and low flow conditions. The specific geometry of the test section is similar to the fuel-target sub-channel of the Savannah River Site (SRS) Mark 22 fuel assembly. Critical Heat Flux (CHF) was observed following flow excursion and flow reversal in the test section. Density wave instability was not recorded in this series of experimental runs. The results of this experimental study show that flow excursion is the dominant instability mode under low flow, low pressure, and down flow conditions. The onset of instability data are plotted on the subcooling-Zuber (phase change) numbers stability plane.

Contact behavior modelling and its size effect on proton exchange membrane fuel cell

Science.gov (United States)

Qiu, Diankai; Peng, Linfa; Yi, Peiyun; Lai, Xinmin; Janßen, Holger; Lehnert, Werner

2017-10-01

Contact behavior between the gas diffusion layer (GDL) and bipolar plate (BPP) is of significant importance for proton exchange membrane fuel cells. Most current studies on contact behavior utilize experiments and finite element modelling and focus on fuel cells with graphite BPPs, which lead to high costs and huge computational requirements. The objective of this work is to build a more effective analytical method for contact behavior in fuel cells and investigate the size effect resulting from configuration alteration of channel and rib (channel/rib). Firstly, a mathematical description of channel/rib geometry is outlined in accordance with the fabrication of metallic BPP. Based on the interface deformation characteristic and Winkler surface model, contact pressure between BPP and GDL is then calculated to predict contact resistance and GDL porosity as evaluative parameters of contact behavior. Then, experiments on BPP fabrication and contact resistance measurement are conducted to validate the model. The measured results demonstrate an obvious dependence on channel/rib size. Feasibility of the model used in graphite fuel cells is also discussed. Finally, size factor is proposed for evaluating the rule of size effect. Significant increase occurs in contact resistance and porosity for higher size factor, in which channel/rib width decrease.

Core design and fuel management studies

International Nuclear Information System (INIS)

Min, Byung Joo; Chan, P.

1997-06-01

The design target for the CANDU 9 requires a 20% increase in electrical power output from an existing 480-channel CANDU core. Assuming a net electrical output of 861 MW(e) for a natural uranium fuelled Bruce-B/Darlington reactor in a warm water site, the net electrical output of the reference CANDU 9 reactor would be 1033 MW(e). This report documents the result of the physics studies for the design of the CANDU 9 480/SEU core. The results of the core design and fuel management studies of the CANDU 9 480/SEU reactor indicated that up to 1033 MW(e) output can be achieved in a 480-channel CANDU core by using SEU core can easily be maintained indefinitely using an automated refuelling program. Fuel performance evaluation based on the data of the 500 FPDs refuelling simulation concluded that SEU fuel failure is not expected. (author). 2 tabs., 38 figs., 5 refs

Detector for failed fuel elements

International Nuclear Information System (INIS)

Ito, Masaru.

1979-01-01

Purpose: To provide automatic monitor for the separation or reactor water and sampling water, in a failed fuel element detector using a sipping chamber. Constitution: A positional detector for the exact mounting of a sipping chamber on a channel box and a level detector for the detection of complete discharge of cooling water in the sipping chamber are provided in the sipping chamber. The positional detector is contacted to the upper end of the channel box and operated when the sipping chamber is correctly mounted to the fuel assemblies. The level detector comprises a float and a limit switch and it is operated when the water in the sipping chamber is discharged by a predetermined amount. Isolation of reactor water and sampling water are automatically monitored by the signal from these two detectors. (Ikeda, J.)

Heat transfer study of a submerged reactor channel under boil-off condition

Energy Technology Data Exchange (ETDEWEB)

Mukhopadhyay, Deb [Bhabha Atomic Research Centre, Mumbai (India). Reactor Safety Div.; Sahoo, P.K. [Indian Institute of Technology, Roorkee (India). Dept. of Mechanical and Industrial Engineering; Ghosh, A.K. [Bhabha Atomic Research Centre, Mumbai (India). Health, Safety and Environment Group

2012-12-15

Experiments have been carried out to study the heatup behavior of a single segmented reactor channel for Pressurized Heavy Water Reactor under submerged, partially submerged and exposed conditions. This situation may arise from a severe accident scenario of Pressurised Heavy Water Reactors where full or segmented reactor channels are likely to be disassembled and form a submerged debris bed. An assembly of electrical heater rod, simulating fuel bundle and channel components like Pressure Tube and Calandria Tube constitutes the segmented reactor channel. Heatup of this assembly is observed with respect to different water levels ranging from full submergence to totally exposed and power levels of 6-8 kW, typical to decay power level. It has been observed from the set of experiment that fuel bundle local dry out followed by heatup does not happen till the bundle is partially submerged. Temperature excursion of the bundle is evident when the bundle is exposed to steam-air environment. (orig.)

Analysis Of The Effect Of Flow Channel Width On The Performance Of PEMFC

Directory of Open Access Journals (Sweden)

Elif Eker

2013-08-01

Full Text Available In this work, it was analysed the effect of different channel width on performance of PEM fuel cell. Current density were measured on the single cells of parallel flow fields that has 25 cm² active layer, using three different kinds of channel width. The cell width and the channel height remain constant.The results show that increasing the channel width while the cell width remains constant decreases the current density.

Subchannel analysis of sodium-cooled reactor fuel assemblies with annular fuel pins

International Nuclear Information System (INIS)

Memmott, Matthew; Buongiorno, Jacopo; Hejzlar, Pavel

2009-01-01

Using a RELAP5-3D subchannel analysis model, the thermal-hydraulic behavior of sodium-cooled fuel assemblies with internally and externally cooled annular fuel rods was investigated, in an effort to enhance the economic performance of sodium-fast reactors by increasing the core power density, decreasing the core pressure drop, and extending the fuel discharge burnup. Both metal and oxide fuels at high and low conversion ratios (CR=0.25 and CR=1.00) were investigated. The externally and internally cooled annular fuel design is most beneficial when applied to the low CR core, as clad temperatures are reduced by up to 62.3degC for the oxide fuel, and up to 18.5degC for the metal fuel. This could result in a power uprates of up to ∼44% for the oxide fuel, and up to ∼43% for the metal fuel. The use of duct ribs was explored to flatten the temperature distribution at the core outlet. Subchannel analyses revealed that no fuel melting would occur in the case of complete blockage of the hot interior-annular channel for both metal and oxide fuels. Also, clad damage would not occur for the metal fuel if the power uprate is 38% or less, but would indeed occur for the oxide fuel. (author)

Optimization of nonthermal fusion power consistent with energy channeling

International Nuclear Information System (INIS)

Snyder, P.B.; Herrmann, M.C.; Fisch, N.J.

1995-02-01

If the energy of charged fusion products can be diverted directly to fuel ions, non-Maxwellian fuel ion distributions and temperature differences between species will result. To determine the importance of these nonthermal effects, the fusion power density is optimized at constant-β for nonthermal distributions that are self-consistently maintained by channeling of energy from charged fusion products. For D-T and D- 3 He reactors, with 75% of charged fusion product power diverted to fuel ions, temperature differences between electrons and ions increase the reactivity by 40-70%, while non- Maxwellian fuel ion distributions and temperature differences between ionic species increase the reactivity by an additional 3-15%

Single-channel model for steady thermal-hydraulic analysis in nuclear reactor

International Nuclear Information System (INIS)

Zhang Xiaoying; Huang Yuanyuan

2010-01-01

This article established a single-channel model for steady analysis in the reactor and an example of thermal-hydraulic analysis was made by using this model, including the Maximum heat flux density of fuel element, enthalpy, Coolant flow, various kinds of pressure drop, enthalpy increase in average tube and thermal tube. I also got the Coolant temperature distribution and the fuel element temperature distribution and analysis of the final result. The results show that some relevant parameters which we got in this paper are well coincide with the actual operating parameters. It is also show that the single-channel model can be used to the steady thermal-hydraulic analysis. (authors)

Evaluation of microstructure of irradiated fuel channel components of PHWR

International Nuclear Information System (INIS)

Ramadasan, E.

2005-01-01

Performance evaluation and failure analysis of irradiated reactor structural components such as those in-core and PHT circuit components necessitate metallographic evaluation using special metallographic specimen preparation techniques due to the radiation dose and contamination levels involved in handling the specimens. The metallographic specimen preparation techniques that are resorted to involve use of fully automatic and semi automatic machines, shielded metallographic microscope and specialised equipment developed for lead-cell metallography. The techniques used and the results obtained in the metallographic studies on irradiated fuel channel components such as pressure tubes and garter springs of various Indian PHWRs at RAPS, NAPS and MAPS are presented as case studies in the paper. The evaluation of oxidation and hydriding behaviour of zircaloy-2 pressure tubes and garter springs are presented. The paper also gives in detail the microstructural evaluation of hydride blistering seen at the PT-CT contact location of the pressure tubes of RAPS-2. The evaluation revealed that the hydride blisters was small compared to their length, unlike the hydride blisters seen in the CANDU pressure tube G-16 of Pickering-2. This could be attributed to be due to the difference in the annulus conditions between the two types of reactors. The hydride blisters in J-07 pressure tube of RAPS-2 had ductile material adjacent to them. The paper also gives the hydride blistering observation on irradiated Zr-2.5% Nb-0.5% Cu garter springs of RAPS-2. It was seen that there was only negligible hydriding of the garter springs during service through they showed presence of benign hydride blisters in them. The general hydriding observations made on the pressure tubes of Indian PHWRs under different conditions are also presented. (author)

Thermo hydraulic analysis of narrow channel effect in supercritical-pressure light water reactor

International Nuclear Information System (INIS)

Zhou Tao; Chen Juan; Cheng Wanxu

2012-01-01

Highlights: ► Detailed thermal analysis with different narrow gaps between fuel rods is given. ► Special characteristics of narrow channels effect on heat transfer in supercritical pressure are shown. ► Reasonable size selection of gaps between fuel rods is proposed for SCWR. - Abstract: The size of the gap between fuel rods has important effects on flow and heat transfer in a supercritical-pressure light water reactor. Based on thermal analysis at different coolant flow rates, the reasonable value range of gap size between fuel rods is obtained, for which the maximum cladding temperature safety limits and installation technology are comprehensively considered. Firstly, for a given design flow rate of coolant, thermal hydraulic analysis of supercritical pressure light water reactor with different gap sizes is provided by changing the fuel rod pitch only. The results show that, by means of reducing the gap size between fuel rods, the heat transfer coefficients between coolant and fuel rod, as well as the heat transfer coefficient between coolant and water rod, would both increase noticeably. Furthermore, the maximum cladding temperature will significantly decrease when the moderator temperature is decreased but coolant temperature remains essentially constant. Meanwhile, the reduction in the maximum cladding temperature in the inner assemblies is much larger than that in the outer assemblies. In addition, the maximum cladding temperature could be further reduced by means of increasing coolant flow rate for each gap size. Finally, the characteristics of narrow channels effect are proposed, and the maximum allowable gap between fuel rods is obtained by making full use of the enhancing narrow channels effect on heat transfer, and concurrently considering installation. This could provide a theoretical reference for supercritical-pressure light water reactor design optimization, in which the effects of gap size and flow rate on heat transfer are both considered.

Channel-Mediated Lactate Release by K+-Stimulated Astrocytes

KAUST Repository

Sotelo-Hitschfeld, T.; Niemeyer, M. I.; Machler, P.; Ruminot, I.; Lerchundi, R.; Wyss, M. T.; Stobart, J.; Fernandez-Moncada, I.; Valdebenito, R.; Garrido-Gerter, P.; Contreras-Baeza, Y.; Schneider, B. L.; Aebischer, P.; Lengacher, S.; San Martin, A.; Le Douce, J.; Bonvento, G.; Magistretti, Pierre J.; Sepulveda, F. V.; Weber, B.; Barros, L. F.

2015-01-01

in response to local field stimulation. The existence of an astrocytic lactate reservoir and its quick mobilization via an ion channel in response to a neuronal cue provides fresh support to lactate roles in neuronal fueling and in gliotransmission.

High conversion pressurized water reactor with boiling channels