Effective new chemicals to prevent corrosion due to chlorine in power plant superheaters

Energy Technology Data Exchange (ETDEWEB)

Martti Aho; Pasi Vainikka; Raili Taipale; Patrik Yrjas [VTT, Jyvaeskylae (Finland)

2008-05-15

Firing or co-firing of biomass in efficient power plants can lead to high-temperature corrosion of superheaters due to condensation of alkali chlorides into superheater deposits. Corrosion can be prevented if a significant portion of the alkali chlorides present in the flue gases is destroyed before reaching the superheaters. The alkali capturing power of aluminium and ferric sulphates was determined in a pilot-scale fluidised bed (FB) reactor. The reagents were added in solution, through a spraying nozzle, to the upper part of the freeboard. Both reagents, at economical dosages, fast and effectively destroyed the alkali chlorides by producing sufficient SO{sub 3} for the sulphation. Both the mass flow rate and type of sulphate affected the sulphation ability. Thus, the cation, too, plays a role in the reaction. The required chemical dosage is not directly proportional to the S{sub reagent}/Cl{sub 2fuel} ratio because alkali chlorides must compete with calcium and magnesium oxides and probably also with alkali oxides for the available SO{sub 3}. 17 refs., 16 figs., 1 tab.

FARO tests corium-melt cooling in water pool: Roles of melt superheat and sintering in sediment

Energy Technology Data Exchange (ETDEWEB)

Hwang, Gisuk [Department of Mechanical Engineering, Wichita State University, Wichita, KS 67260 (United States); Kaviany, Massoud [Department of Mechanical Engineering, University of Michigan, Ann Arbor, MI 48109 (United States); Division of Advance Nuclear Engineering, POSTECH, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Moriyama, Kiyofumi [Division of Advance Nuclear Engineering, POSTECH, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Park, Hyun Sun, E-mail: hejsunny@postech.ac.kr [Division of Advance Nuclear Engineering, POSTECH, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Hwang, Byoungcheol; Lee, Mooneon; Kim, Eunho; Park, Jin Ho [Division of Advance Nuclear Engineering, POSTECH, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Nasersharifi, Yahya [Department of Mechanical Engineering, Wichita State University, Wichita, KS 67260 (United States)

2016-08-15

Highlights: • The numerical approach for FARO experimental data is suggested. • The cooling mechanism of ex-vessel corium is suggested. • The predicted minimum pool depth for no cake formation is suggested. - Abstract: The FARO tests have aimed at understanding an important severe accident mitigation action in a light water reactor when the accident progresses from the reactor pressure vessel boundary. These tests have aimed to measure the coolability of a molten core material (corium) gravity dispersed as jet into a water pool, quantifying the loose particle diameter distribution and fraction converted to cake under range of initial melt superheat and pool temperature and depth. Under complete hydrodynamic breakup of corium and consequent sedimentation in the pool, the initially superheated corium can result in debris bed consisting of discrete solid particles (loose debris) and/or a solid cake at the bottom of the pool. The success of the debris bed coolability requires cooling of the cake, and this is controlled by the large internal resistance. We postulate that the corium cake forms when there is a remelting part in the sediment. We show that even though a solid shell forms around the melt particles transiting in the water pool due to film-boiling heat transfer, the superheated melt allows remelting of the large particles in the sediment (depending on the water temperature and the transit time) using the COOLAP (Coolability Analysis with Parametric fuel-cooant interaction models) code. With this remelting and its liquid-phase sintering of the non-remelted particles, we predict the fraction of the melt particles converting to a cake through liquid sintering. Our predictions are in good agreement with the existing results of the FARO experiments. We address only those experiments with pool depths sufficient/exceeding the length required for complete breakup of the molten jet. Our analysis of the fate of molten corium aimed at devising the effective

Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings -- Phase 2 field testing

Energy Technology Data Exchange (ETDEWEB)

Blough, J.L. [Foster Wheeler Development Corp., Livingston, NJ (United States)

1996-08-01

In Phase 1 of this project, a variety of developmental and commercial tubing alloys and claddings was exposed to laboratory fireside corrosion testing simulating a superheater or reheater in a coal-fired boiler. Phase 2 (in situ testing) has exposed samples of 347, RA85H, HR3C, 253MA, Fe{sub 3}Al + 5Cr, 310 modified, NF 709, 690 clad, and 671 clad for over 10,000 hours to the actual operating conditions of a 250-MW coal-fired boiler. The samples were installed on air-cooled, retractable corrosion probes, installed in the reheater cavity, controlled to the operating metal temperatures of an existing and advanced-cycle, coal-fired boiler. Samples of each alloy are being exposed for 4,000, 12,000, and 16,000 hours of operation. The present results are for the metallurgical examination of the corrosion probe samples after approximately 4,400 hours of exposure.

Effects of Different Fuel Specifications and Operation Conditions on the Performance of Coated and Uncoated Superheater Tubes in Two Different Biomass-Fired Boilers

DEFF Research Database (Denmark)

Wu, Duoli; Dahl, Kristian V.; Madsen, Jesper L.

2018-01-01

Fireside corrosionis a serious concern in biomass firing powerplants such that the efficiency of boilers is limited by high temperature corrosion. Application of protective coatings on superheater tubes is a possible solution to combat fireside corrosion. The current study investigates the corros......Fireside corrosionis a serious concern in biomass firing powerplants such that the efficiency of boilers is limited by high temperature corrosion. Application of protective coatings on superheater tubes is a possible solution to combat fireside corrosion. The current study investigates...... the corrosion performance of coated tubes compared to uncoated Esshete 1250 and TP347H tubes, which were exposed in two different biomass-fired boilers for one year. Data on the fuel used, temperature of the boilers, and temperature fluctuations are compared for the two boilers, and how these factors influence...... deposit formation, corrosion, and the stability of the coatings is discussed. The coatings (Ni and Ni2Al3) showed protective behavior ina wood-fired plant where the outlet steam temperature was 520 °C. However, at the plant that fired straw with an outlet steam temperature of 540 °C and where severe...

Adaptive Superheat Control of a Refrigeration Plant using Backstepping

DEFF Research Database (Denmark)

Rasmussen, Henrik

2008-01-01

This paper proposes a novel method for superheat and capacity control of refrigeration systems. The new idea is to control the superheat by the compressor speed and capacity by the refrigerant flow. This gives a highly nonlinear transfer operator from compressor speed input to the superheat output....... A new low order nonlinear model of the evaporator is developed and used in a backstepping design of an adaptive nonlinear controller.  The stability of the proposed method is validated theoretically by Lyapunov analysis and experimental results shows the performance of the system for a wide range...

Superheat in magma oceans

Science.gov (United States)

Jakes, Petr

1992-01-01

The existence of 'totally molten' planets implies the existence of a superheat (excess of heat) in the magma reservoirs since the heat buffer (i.e., presence of crystals having high latent heat of fusion) does not exist in a large, completely molten reservoir. Any addition of impacting material results in increase of the temperature of the melt and under favorable circumstances heat is stored. The behavior of superheat melts is little understood; therefore, we experimentally examined properties and behavior of excess heat melts at atmospheric pressures and inert gas atmosphere. Highly siliceous melts (70 percent SiO2) were chosen for the experiments because of the possibility of quenching such melts into glasses, the slow rate of reaction in highly siliceous composition, and the fact that such melts are present in terrestrial impact craters and impact-generated glasses. Results from the investigation are presented.

Superheater corrosion in biomass-fired power plants: Investigation of Welds

DEFF Research Database (Denmark)

Montgomery, Melanie; Carlsen, B; Biede, O

2002-01-01

-fired Masnedø combined heat and power (CHP) plant to investigate corrosion at temperatures higher than that of the actual plant. The highest steam temperature investigated was 570°C. Various alloys of 12-22% chromium content were welded into this test loop. Their corrosion rates were similar and increased...... condense on superheater components. This gives rise to specific corrosion problems not previously encountered in coal-fired power plants. The type of corrosion attack can be directly ascribed to the composition of the deposit and the metal surface temperature. A test superheater was built into the straw...... with temperature. The mechanism of attack was grain boundary attack as a precursor to selective chromium depletion of the alloy. In addition welds coupling various tubes sections were also investigated. It was seen that there was preferential attack around those welds that had a high nickel content. The welds...

Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings -- Phase 2 field testing

Energy Technology Data Exchange (ETDEWEB)

Blough, J.L.; Seitz, W.W.; Girshik, A. [Foster Wheeler Development Corp., Livingston, NJ (United States)

1998-06-01

In Phase 1 of this project, laboratory experiments were performed on a variety of developmental and commercial tubing alloys and claddings by exposing them to fireside corrosion tests which simulated a superheater or reheater in a coal-fired boiler. Phase 2 (in situ testing) has exposed samples of 347, RA85H, HR3C, RA253MA, Fe{sub 3}Al + 5Cr, Ta-modified 310, NF 709, 690 clad, 671 clad, and 800HT for up to approximately 16,000 hours to the actual operating conditions of a 250-MW, coal-fired boiler. The samples were installed on air-cooled, retractable corrosion probes, installed in the reheater cavity, and controlled to the operating metal temperatures of an existing and advanced-cycle, coal-fired boiler. Samples of each alloy were exposed for 4,483, 11,348, and 15,883 hours of operation. The present results are for the metallurgical examination of the corrosion probe samples after the full 15,883 hours of exposure. A previous topical report has been issued for the 4,483 hours of exposure.

Non-contact Measurement of Remaining Thickness of Corroding Superheater Tubes. Phase 2

Energy Technology Data Exchange (ETDEWEB)

Borggreen, Kjeld; Storesund, Jan

2007-12-15

The Electro Magnetic Acoustic Transducer (EMAT) is a contactless thickness gauge for detection of corrosion on superheater tubes; it candidates as substitute for conventional manually operated contact UT transducers. It is the aim of the project to demonstrate the usefulness of two simple EMAT systems, Panametrics and Sonatest, for fast and reliable tube thickness inspections in difficult-to-access superheater sections. The present Phase 2 of the project involves testing of the systems on real compact superheaters in remote operation with the help of a mechanical manipulator designed and built for the purpose. The results are the following: - Both EMAT systems work well when tested in the field during handheld operation on tubes with a moderate thick layer of corrosion products and ash. The practical obtainable speed of cross-panel inspection of easily accessible panels is approximately 6 tubes per minute (6 thickness readings per minute). - The Sonatest system works well when tested in the field during remote operation on heavily corroded superheater tubes with thick ash layer. The Panametrics system was not found suitable for this type of field work. - The mechanical manipulator works well for cross-panel inspection of difficult-to-access superheater sections independent of the tube dimensions and the free space between the panels. In its present design it needs few improvements. - The practical obtainable speed of cross-panel inspection is 3 tubes per minute (3 thickness readings per minute). This speed is limited by the detection rate of the EMAT system and not by the travelling speed of the probe. - Scanning of tubes along their axis was never attempted, because the EMAT instruments were not capable of collecting data coming as a continuous stream. - It cannot be judged from visual alone and hardly from the service data, if a tube or a panel can be inspected by the magnetostrictive EMAT method or not. - The main contribution to failure of the EMAT inspection

Sodium-fuel interaction: dropping experiments and subassembly test

International Nuclear Information System (INIS)

Holtbecker, H.; Schins, H.; Jorzik, E.; Klein, K.

1978-01-01

Nine dropping tests, which bring together 2 to 4 kg of molten UO 2 with 150 l sodium, showed the incoherency and non-violence of these thermal interactions. The pressures can be described by sodium incipient boiling and bubble collapse; the UO 2 fragmentation by thermal stress and bubble collapse impact forces. The mildness of the interaction is principally due to the slowness and incoherency of UO 2 fragmentation. This means that parametric models which assume instantaneous mixing and fragmentation are of no use for the interpretation of dropping experiments. One parametric model, the Caldarola Fuel Coolant Interaction Variable Mass model, is being coupled to the two dimensional time dependent hydrodynamic REXCO-H code. In a first step the coupling is applicated to a monodimensional geometry. A subassembly test is proposed to validate the model. In this test rapid mixing between UO 2 and sodium has to be obtained. Dispersed molten UO 2 fuel is obtained by flashing injected sodium drops inside a UO 2 melt. This flashing is theoretically explained and modelled as a superheat limited explosion. The measured sodium drop dwell times of two experiments are compared to results obtained from the mentioned theory, which is the basis of the Press 2 Code

Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings -- Phase 2

Energy Technology Data Exchange (ETDEWEB)

Blough, J.L.; Seitz, W.W. [Foster Wheeler Development Corp., Livingston, NJ (United States)

1997-12-01

In Phase 1 a variety of developmental and commercial tubing alloys and claddings were exposed to laboratory fireside corrosion testing simulating a superheater or reheater in a coal-fired boiler. Phase 2 (in situ testing) has exposed samples of 347 RA-85H, HR3C, 253MA, Fe{sub 3}Al + 5Cr, 310 Ta modified, NF 709, 690 clad, and 671 clad for approximately 4,000, 12,000, and 16,000 hours to the actual operating conditions of a 250-MW coal-fired boiler. The samples were assembled on an air-cooled, retractable corrosion probe, the probe was installed in the reheater activity of the boiler and controlled to the operating metal temperatures of an existing and advanced-cycle coal-fired boiler. The results will be presented for the preliminary metallurgical examination of the corrosion probe samples after 16,000 hours of exposure. Continued metallurgical and interpretive analysis is still on going.

Simulasi Thermal Stress Pada Tube Superheater Package Boiler

OpenAIRE

Hamdani

2013-01-01

This project investigates the thermal stress behavior and the mechanisms of superheater tube failure with experimental method and numerical analysis. First of all the procedures for failure analysis were applied to determine the root cause of them. A visual assessment of boiler critical pressure parts was carried out, and then the failed tube is examined by nondestructive evaluation. For the numerical domain, initially the elastic solution for a superheater tube subjected to an internal press...

A steam separator-superheater apparatus

International Nuclear Information System (INIS)

Androw, Jean; Bessouat, Roger; Peyrelongue, J.-P.

1973-01-01

Description is given of a separator-superheater apparatus comprising an outer enclosure containing a separating-unit and a steam superheating unit according to the main patent. The present addition relates to an improvement in that apparatus, characterized in that the separating unit and the superheating unit, mounted in two distinct portions of the outer enclosure, are divided into the same number of sub-units of each unit being identical and operating in parallel, and in that to each separator sub-unit is associated a superheater sub-unit, said sub-units being mounted in series and located in one in the other of the enclosure two portions, respectively. This can be applied to the treatment of the exhaust steam of a turbine high pressure body, prior to re-injecting said steam into the low pressure body [fr

Nonlinear Superheat Control of a Refrigeration Plant using Backstepping

DEFF Research Database (Denmark)

Rasmussen, Henrik

2008-01-01

This paper proposes a novel method for superheat and capacity control of refrigeration systems. The main idea is to control the superheat by the compressor speed and capacity by the refrigerant flow. A new low order nonlinear model of the evaporator is developed and used in a backstepping design...... of a nonlinear controller. The proposed method is validated by experimental results....

Automatic Tuning of the Superheat Controller in a Refrigeration Plant

DEFF Research Database (Denmark)

Rasmussen, Henrik; Thybo, Claus; Larsen, Lars F. S.

2006-01-01

This paper proposes an automatic tuning of the superheat control in a refrigeration system using a relay method. By means of a simple evaporator model that captures the important dynamics and non-linearities of the superheat a gain-scheduling that compensates for the variation of the process gain...

Analysis of temperature and stress distribution of superheater tubes after attemperation or sootblower activation

International Nuclear Information System (INIS)

Madejski, Paweł; Taler, Dawid

2013-01-01

Highlights: • The CFD simulation was used to calculate 3D steam and tube wall temperature distributions in the platen superheater. • The CFD results can be used in design of superheaters made of tubes with complex cross-section. • The CFD analysis enables the proper selection of the steel grade. • The transient temperature and stress distributions were calculated using Finite Volume Method. • The detailed analysis prevents superheater tubes from excessive stresses during sootblower or attemperator activation. - Abstract: Superheaters are characterized by high metal temperatures due to higher steam temperature and low heat transfer coefficients on the tube inner surfaces. Superheaters have especially difficult operating conditions, particularly during attemperator and sootblower activations, when temperature and steam flow rate as well as tube wall temperature change with time. A detailed thermo-mechanical analysis of the superheater tubes makes it possible to identify the cause of premature high-temperature failures and aids greatly in the changes in tubing arrangement and improving start-up technology. This paper presents a thermal and strength analysis of a tube “double omega”, used in the steam superheaters in CFB boilers

Non-contact Measurement of Remaining Thickness of Corroding Superheater Tubes. Phase 1

Energy Technology Data Exchange (ETDEWEB)

Borggreen, Kjeld; Storesund, Jan

2006-10-15

Corrosion of superheaters has become a severe problem in many biomass boilers and incineration plants. This new situation calls for frequent tube wall thickness testing of the superheaters during very short shut-downs. To meet this demand Electro Magnetic Acoustic Transducer (EMAT) candidates as substitute for conventional manually operated contact UT-transducers. The EMAT can contactlessly generate an ultrasonic wave in the interphase between the external oxide and the metal. This means that measurements can be undertaken much quicker and with a much higher coverage simultaneously, without preceding blast operations. It is the aim of the project to demonstrate the usefulness of two simple EMAT systems, Panametrics and Sonatest, for fast and reliable tube thickness inspections in difficult-to-access superheater sections. The present Phase 1 of the project involves testing of the performance of the two systems in laboratory with the following results: 1. Both instruments work well on plate, tube, and pipe samples assuming the presence of an external oxide layer formed at a temperature above approximately 400 deg C. 2. Both instruments work well on all types of ferritic and martensitic steels but not on austenitic steels. 3. Both instruments work well independent of the thickness of the active oxide layer. 4. Both instruments work well independent of tube diameter, wall thickness, and sample width. 5. Both instruments work well over a very large range of wall thicknesses. Minimum tube wall thickness is less than 1.8 mm. 6. The tolerable lift-off (free distance between transducer and tube surface) is 2.4 - 3.0 mm for Panametrics system and 3.6 - 4.8 mm for Sonatest's system. The tolerable lift-off is a measure of the thickness of ash deposits, which can be tolerated on the tube surface as well as the misplacement, which can be tolerated in case of remote tube testing. 7. The tolerable off-set between tube axis and probe axis is very large for both instruments (10

Single Temperature Sensor Superheat Control Using a Novel Maximum Slope-seeking Method

DEFF Research Database (Denmark)

Vinther, Kasper; Rasmussen, Henrik; Izadi-Zamanabadi, Roozbeh

2013-01-01

Superheating of refrigerant in the evaporator is an important aspect of safe operation of refrigeration systems. The level of superheat is typically controlled by adjusting the flow of refrigerant using an electronic expansion valve, where the superheat is calculated using measurements from...

Nonlinear Superheat and Evaporation Temperature Control of a Refrigeration Plant

DEFF Research Database (Denmark)

Rasmussen, Henrik; Thybo, Claus; Larsen, Lars F. S.

2006-01-01

This paper proposes novel control of the superheat of the evaporator in a refrigeration system. A new model of the evaporator is developed and based on this model the superheat is transferred to a referred variable. It is shown that control of this variable leads to a linear system independent...... of the working point. The model also gives a method for control of the evaporation temperature. The proposed method is validated by experimental results....

Steam separator-superheater with drawing of a fraction of the dried steam

International Nuclear Information System (INIS)

Bessouat, Roger; Marjollet, Jacques.

1976-01-01

This invention concerns a vertical separator-superheater of the steam from a high pressure expansion turbine before it is admitted to an expansion turbine at a lower pressure, by heat exchange with steam under a greater pressure, and drawing of a fraction of the dried steam before it is superheated. Such drawing off is necessary in the heat exchange systems of light water nuclear reactors. Its purpose is to provide a separator-superheater that provides an even flow of non superheated steam and a regular distribution of the steam to be superheated to the various superheating bundles, with a significantly uniform temperature of the casing, thereby preventing thermal stresses and ensuring a minimal pressure drop. The vertical separator-superheater of the invention is divided into several vertical sections comprising as from the central area, a separation area of the steam entrained water and a superheater area and at least one other vertical section with only a separation area of the steam entrained water [fr

Remaining Life Estimation Of Secondary Superheater Outlet On Industrial Electrical Boiler

International Nuclear Information System (INIS)

Soedardjo; Andryansyah; Arhatari, B.D.; Natsir, Muhammad; Triyadi, Ari; Farokhi

2001-01-01

Remaining life estimation of secondary superheater header outlet (SSHO) on industrial electrical boiler has been carried out. Estimation conducted by the observation of microstructure cavitation development based on Neubauer and Wedel theory. The result is available for isolated cavitation development present yet. That Secondary Superheater Outlet component is in good condition after 14 years operated and predicted could be operated for 36 years again

Oxidation rate in ferritic superheater materials

International Nuclear Information System (INIS)

Falk, I.

1992-05-01

On the steam side of superheater tubes, compact oxide layers are formed which have a tendency to crack and flake off (exfoliate). Oxide particles then travel with the steam and can give rise to erosion damage in valves and on turbine blades. In an evaluation of conditions in superheater tubes from Swedish power boilers, it was found that the exfoliation frequency for one material quality (SS 2218) was greater than for other qualities. Against this background, a literature study has been carried out in order to determine which mechanisms govern the build-up of oxide and the exfoliation phenomenon. The study reveals that the oxide morphology is similar on all ferritic steels with Cr contents up to 5%. and that the oxide properties can therefore be expected to be similar. The reason why the exfoliation frequency is greater for tubes of SS 2218 is probably that the tubes have been exposed to higher temperatures. SS 2218 (2.25 Cr) is normally used in a higher temperature range which is accompanied by improved strength data as compared with SS 2216 (1 Cr). The principal cause of the exfoliation is said to be stresses which arise in the oxide during the cooling-down process associated with shutdowns. The stresses give rise to longitudinal cracks in the oxide, and are formed as a result of differences in thermal expansion between the oxide and the tube material. In addition, accounts are presented of oxidation constants and growth velocities, and thickness and running time. These data constitute a valuable basis for practical estimates of the operating temperature in routine checks and investigations into damage in superheater tubes. (au)

DYNAM, Once Through Boiling Flow with Steam Superheat, Laplace Transformation

International Nuclear Information System (INIS)

Schlueter, G.; Efferding, L.E.

1973-01-01

1 - Description of problem or function: DYNAM performs a dynamic analysis of once-through boiling flow oscillations with steam superheat. The model describing the superheat regime (single- phase, variable density fluid) for subcritical pressure operation is also applicable to the study of once-through operation using supercritical pressure water. 2 - Method of solution: Linearized partial differential conservation equations are solved using Laplace transformation of the temporal terms and integration of the spatial variations. DYNAM is written in complex variable notation. 3 - Restrictions on the complexity of the problem - Maxima of: 30 intervals used to describe the power distribution in the non-boiling and boiling regions, 29 boiling nodes, 7 intervals and corresponding friction multipliers read in per case, 14 exit qualities read in per case, 40 superheat nodes, 10 coefficients read in for the phi 2 vs, x-polynomial fit, 48 frequencies at which open-loop frequency response is desired, 48 frequencies at which signal output is desired

Thermomechanical CSM analysis of a superheater tube in transient state

Science.gov (United States)

Taler, Dawid; Madejski, Paweł

2011-12-01

The paper presents a thermomechanical computational solid mechanics analysis (CSM) of a pipe "double omega", used in the steam superheaters in circulating fluidized bed (CFB) boilers. The complex cross-section shape of the "double omega" tubes requires more precise analysis in order to prevent from failure as a result of the excessive temperature and thermal stresses. The results have been obtained using the finite volume method for transient state of superheater. The calculation was carried out for the section of pipe made of low-alloy steel.

Review of Development Status of Nuclear Superheat; Expose sur l'etat actuel des travaux concernant la surchauffe nucleaire; Obzor razrabotki voprosa o yadernykh peregrevatelyakh; Estudio de los progresos realizados en niateria de sobrecalentamiento nuclear

Energy Technology Data Exchange (ETDEWEB)

Imhoff, D. H.; Pennington, R. T. [General Electric Company, San Jose, CA (United States)

1963-10-15

fuel irradiation performance, in-pile and out-of-pile uniform and localized corrosion evaluations, results from thermal superheat critical experiments, results from experimental heat-transfer testing and a brief appraisal of the economic incentives of the separate superheat reactor, integral superheat reactor, and mixed-spectrum superheat reactor design studies. (b) A brief description is given of the USAEC-Sponsored ESADA-VESR nuclear superheat fuel development programme. The development tasks, the initial core superheat fuel-element design, the range of experimental variables, and the expected results for the three-year fuel development programme are discussed. (author [French] Depuis 1959, la Societe General Electric s'occupe activement de mettre au point la surchauffe nucleaire dans les reacteurs ralentis a l'eau ordinaire. A cette epoque, le ''Superheat Advance Demonstration Experiment'' (SADE), finance par la societe, a permis d'obtenir, pour la premiere fois aux Etats-Unis, de la vapeur surchauffee par un dispositif nucleaire. Les auteurs du memoire font le point de la situation actuelle concernant la surchauffe nucleaire. Ils decrivent, dans une premiere partie, les trois principales installations de surchauffe nucleaire utilisees par la General Electric et, dans la seconde, les deux principaux programmes de recherches sur la surchauffe ainsi que les resultats les plus importants obtenus jusqu'ici dans ce domaine. 1. Principales installations pour les experiences de surchauffe: a) Breve description du SADE utilise dans le reacteur a eau bouillante de Vallecitos (VBWR), avec tableaux des conditions dans lesquelles se deroulent les experiences, et des elements combustibles irradies pendant la periode comprise entre mai 1959 et juin 1962; puis examen critique des resultats les plus importants ainsi obtenus. b) Breve description de l' ''Expanded Superheat Advance Demonstration Experiment'' (E-SADE) installe dans le reacteur de Vallecitos, avec tableaux des

Influence of Superheated Steam Temperature Regulation Quality on Service Life of Boiler Steam Super-Heater Metal

Directory of Open Access Journals (Sweden)

G. T. Kulakov

2009-01-01

Full Text Available The paper investigates influence of change in quality of superheated steam temperature regulations on service life of super-heater metal. А dependence between metal service life and dispersion value for different steel grades has been determined in the paper. Numerical values pertaining to increase of super-heater metal service life in case of transferring from manual regulation to standard system of automatic regulation (SAR have been determined and in case of transferring from standard SAR to improved SAR. The analysis of tabular data and plotted dependencies makes it possible to conclude that any change in conditions of convection super-heater metal work due to better quality of the regulation leads to essential increase of time period which is left till the completion of the service life of a super-heater heating surface.

Contamination prevention of superheaters and reheaters during initial startup and operation

International Nuclear Information System (INIS)

Gabrielli, F.; Sylvester, W.R.; Thimot, G.W.

1976-01-01

The general precautions that should be taken to minimize the potential for harmful contamination or oxygen corrosion of power plant superheaters and reheaters during the period from field storage through operation are discussed and summarized. Present boiler industry start-up and operating practices intended to minimize the introduction of solids to the superheater are, as proven by experience, adequate to avoid contamination-related problems. No basic changes to general industry practice are necessary. What is needed, however, is a continuing awareness of the potential for contamination-related problems so that in the specific application of these practices all likely sources of contamination will be considered

Phase Identification and Internal Stress Analysis of Steamside Oxides on Plant Exposed Superheater Tubes

DEFF Research Database (Denmark)

Pantleon, Karen; Montgomery, Melanie

2012-01-01

During long-term, high-temperature exposure of superheater tubes in thermal power plants, various oxides are formed on the inner side (steamside) of the tubes, and oxide spallation is a serious problem for the power plant industry. Most often, oxidation in a steam atmosphere is investigated...... in laboratory experiments just mimicking the actual conditions in the power plant for simplified samples. On real plant-exposed superheater tubes, the steamside oxides are solely investigated microscopically. The feasibility of X-ray diffraction for the characterization of steamside oxidation on real plant......-exposed superheater tubes was proven in the current work; the challenges for depth-resolved phase analysis and phase-specific residual stress analysis at the inner side of the tubes with concave surface curvature are discussed. Essential differences between the steamside oxides formed on two different steels...

Survey of the current state of knowledge of incipient boiling superheat in sodium

International Nuclear Information System (INIS)

Greer, B.

1979-01-01

Superheat data obtained by various investigators indicate that many parameters affect this phenomenon. Controlling parameters appear to be inert gas concentration, oxide concentration, system pressure, pressure-temperature history, rate of temperature rise, heat flux, flow rate, operating time on the system, surface conditions, and radiation. Of these, the two believed most influential in controlling incipient boiling superheat are the inert gas concentration and oxide concentration. Experimental results for the heat flux and rate of temperature rise appear to be the most inconsistent

Overheating failure of superheater suspension tubes of a captive thermal power plant boiler

International Nuclear Information System (INIS)

Bhattacharya, Sova; Amir, Q.M.; Kannan, C.; Mahapatra, S.B.

2000-01-01

Failure of boiler tubes is the foremost cause of forced boiler outages. One of the predominant failure mechanism of boiler tubes is the stress rupture failure in the form of either short term overheating or long term overheating which are normally encountered in superheater and reheater sections working in the creep range. The strength of the boiler tube depends on the stress level as well on the temperature of exposure in the creep range. An increase in either can reduce the time to rupture. Time at the exposure temperature is an important factor based on which the failures are categorised as either short term or long term. Though there is no established time duration criteria demarcating the short or long term stress rupture failures, depending on the various manifestations on the failed samples, one can categorise the failure. This paper addresses one such stress rupture failure in the superheater section of a captive thermal power plant of a refinery. Multiple failures on the suspension coil of a superheater section was investigated for the cause of failure. Laboratory investigation of the failed sample involved visual inspection, dimensional measurements, chemical analysis of internal deposits and microstructural study. On the basis of these, the failure was attributed to deposition of trisodium phosphate carried over by the feed water into the superheater section resulting in chokage and increase in local operating hoop stresses of the tube. The ultimate failure was thus categorised as long term overheating failure. (author)

Nonlinear superheat and capacity control of a refrigeration plant

DEFF Research Database (Denmark)

Rasmussen, Henrik; Larsen, Lars F. S.

2009-01-01

This paper proposes a novel method for superheat and capacity control of refrigeration systems. A new low order nonlinear model of the evaporator is developed and used in a backstepping design of a nonlinear controller. The stability of the proposed method is validated theoretically by Lyapunov...

Analysis of the heat and mass transfer processes of a UO2 bubble in sodium for the Fuel Aerosol Simulant Test (FAST)

International Nuclear Information System (INIS)

Tobias, M.L.

1979-01-01

The anticipated behavior of uranium oxide vapor bubbles produced by the capacitor discharge vaporization (CDV) method in the Fuel Aerosol Simulant Test (FAST) Facility is discussed on the basis of relatively simple physical models. Results of calculations for the rate of bubble rise and for heat and mass transfer rates are presented. Parametric studies indicate that future analysis efforts should emphasize the diffusion condensation process and the loss of heat from the bubble by radiation. Transfer of heat in the surrounding sodium is rapid enough that simplified models should be adequate. No important effects were noted in connection with bubble depth, initial quantity of UO 2 , or initial superheat

Research on the Superheater Material Properties for USC Boiler with 700°C Steam Parameter

Science.gov (United States)

Chongbin, Wang; Xueyuan, Xu; Yufeng, Zhu; Yongqiang, Jin; Hui, Tong; Yu, Wang; Xiaoli, Lu

This paper discusses the materials' properties of superheater for 700°C USC boiler, including Sanicro25, HR6W, 617mod and 740H, and analyzes the range of applicable temperature of superheater made of different tubes, such as T91, T92, Super304H, TP310HCbN, Sanicro25, HR6W, 617Mod and 740H. In addition, some suggestions on the material selection have been proposed.

Integrated boiler, superheater, and decomposer for sulfuric acid decomposition

Science.gov (United States)

Moore, Robert [Edgewood, NM; Pickard, Paul S [Albuquerque, NM; Parma, Jr., Edward J.; Vernon, Milton E [Albuquerque, NM; Gelbard, Fred [Albuquerque, NM; Lenard, Roger X [Edgewood, NM

2010-01-12

A method and apparatus, constructed of ceramics and other corrosion resistant materials, for decomposing sulfuric acid into sulfur dioxide, oxygen and water using an integrated boiler, superheater, and decomposer unit comprising a bayonet-type, dual-tube, counter-flow heat exchanger with a catalytic insert and a central baffle to increase recuperation efficiency.

A fault tolerant superheat control strategy for supermarket refrigeration systems

DEFF Research Database (Denmark)

Vinther, Kasper; Izadi-Zamanabadi, Roozbeh; Rasmussen, Henrik

2013-01-01

, based on a maximum slope-seeking control method and only a single temperature sensor, is developed to drive the evaporator outlet temperature to a level that gives a suitable superheat of the refrigerant. The FTC strategy requires no a priori system knowledge or additional hardware and functions...

Tuning and performance evaluation of PID controller for superheater steam temperature control of 200 MW boiler using gain phase assignment algorithm

Science.gov (United States)

Begum, A. Yasmine; Gireesh, N.

2018-04-01

In superheater, steam temperature is controlled in a cascade control loop. The cascade control loop consists of PI and PID controllers. To improve the superheater steam temperature control the controller's gains in a cascade control loop has to be tuned efficiently. The mathematical model of the superheater is derived by sets of nonlinear partial differential equations. The tuning methods taken for study here are designed for delay plus first order transfer function model. Hence from the dynamical model of the superheater, a FOPTD model is derived using frequency response method. Then by using Chien-Hrones-Reswick Tuning Algorithm and Gain-Phase Assignment Algorithm optimum controller gains has been found out based on the least value of integral time weighted absolute error.

Effects of Degree of Superheat on the Running Performance of an Organic Rankine Cycle (ORC Waste Heat Recovery System for Diesel Engines under Various Operating Conditions

Directory of Open Access Journals (Sweden)

Kai Yang

2014-04-01

Full Text Available This study analyzed the variation law of engine exhaust energy under various operating conditions to improve the thermal efficiency and fuel economy of diesel engines. An organic Rankine cycle (ORC waste heat recovery system with internal heat exchanger (IHE was designed to recover waste heat from the diesel engine exhaust. The zeotropic mixture R416A was used as the working fluid for the ORC. Three evaluation indexes were presented as follows: waste heat recovery efficiency (WHRE, engine thermal efficiency increasing ratio (ETEIR, and output energy density of working fluid (OEDWF. In terms of various operating conditions of the diesel engine, this study investigated the variation tendencies of the running performances of the ORC waste heat recovery system and the effects of the degree of superheat on the running performance of the ORC waste heat recovery system through theoretical calculations. The research findings showed that the net power output, WHRE, and ETEIR of the ORC waste heat recovery system reach their maxima when the degree of superheat is 40 K, engine speed is 2200 r/min, and engine torque is 1200 N·m. OEDWF gradually increases with the increase in the degree of superheat, which indicates that the required mass flow rate of R416A decreases for a certain net power output, thereby significantly decreasing the risk of environmental pollution.

The Phenomenon of Superheat of Liquids: In Memory of Vladimir P. Skripov

Science.gov (United States)

Skripov, P. V.; Skripov, A. P.

2010-05-01

This article is devoted to the memory of Vladimir P. Skripov (1927-2006). He has received worldwide recognition for his monograph on metastable liquids published in 1972 (the English edition was published in 1974). We briefly discuss some studies deal with the phenomenon of attainable superheat of liquids and with measurements of thermophysical properties of liquids under conditions of a moderate degree of superheat. Main attention is paid to the studies performed by V.P. Skripov and his research group in the 1960s and 1970s. Experimental methods which provided break-throughs in research on both spontaneous boiling-up kinetics and substance properties (the specific volume, isobaric heat capacity, ultrasound speed, and viscosity) in super-heated states are presented.

A Stochastic mesoscopic model for predicting the globular grain structure and solute redistribution in cast alloys at low superheat

International Nuclear Information System (INIS)

Nastac, Laurentiu; El Kaddah, Nagy

2012-01-01

It is well known that casting at low superheat has a strong influence on the solidification morphology and macro- and microstructures of the cast alloy. This paper describes a stochastic mesoscopic solidification model for predicting the grain structure and segregation in cast alloy at low superheat. This model was applied to predict the globular solidification morphology and size as well as solute redistribution of Al in cast Mg AZ31B alloy at superheat of 5°C produced by the Magnetic Suspension Melting (MSM) process, which is an integrated containerless induction melting and casting process. The castings produced at this low superheat have fine globular grain structure, with an average grain size of 80 μm, which is about 3 times smaller than that obtained by conventional casting techniques. The stochastic model was found to reasonably predict the observed grain structure and Al microsegregation. This makes the model a useful tool for controlling the structure of cast magnesium alloys.

Diagnosis of Heat Exchanger Tube Failure in Fossil Fuel Boilers Through Estimation of Steady State Operating Conditions

International Nuclear Information System (INIS)

Herszage, A.; Toren, M.

1998-01-01

Estimation of operating conditions for fossil fuel boiler heat exchangers is often required due to changes in working conditions, design modifications and especially for monitoring performance and failure diagnosis. Regular heat exchangers in fossil fuel boilers are composed of tube banks through which water or steam flow, while hot combustion (flue) gases flow outside the tubes. This work presents a top-down approach to operating conditions estimation based on field measurements. An example for a 350 MW unit superheater is thoroughly discussed. Integral calculations based on measurements for all unit heat exchangers (reheaters, superheaters) were performed first. Based on these calculations a scheme of integral conservation equations (lumped parameter) was then formulated at the single tube level. Steady state temperatures of superheater tube walls were obtained as a main output, and were compared to the maximum allowable operating temperatures of the tubes material. A combined lumped parameter - CFD (Computational Fluid Dynamics, FLUENT code) approach constitutes an efficient tool in certain cases. A brief report of such a case is given for another unit superheater. We conclude that steady state evaluations based on both integral and detailed simulations are a valuable monitoring and diagnosis tool for the power generation industry

Review about corrosion of superheaters tubes in biomass plants

International Nuclear Information System (INIS)

Berlanga-Labari, C.; Fernandez-Carrasquilla, J.

2006-01-01

The design of new biomass-fired power plants with increased steam temperature raises concerns of high-temperature corrosion. The high potassium and chlorine contents in many biomass, specially in wheat straw, are potentially harmful elements with regard to corrosion. Chlorine may cause accelerated corrosion resulting in increased oxidation, metal wastage, internal attack, void formations and loose non-adherent scales. The most severe corrosion problems in biomass-fired systems are expected to occur due to Cl-rich deposits formed on superheater tubes. In the first part of this revision the corrosion mechanism proposed are described in function of the conditions and compounds involved. The second part is focused on the behaviour of the materials tested so far in the boiler and in the laboratory. First the traditional commercial alloys are studied and secondly the new alloys and the coasting. (Author). 102 refs

Preventing superheater corrosion by additives; Tulistimien kloorikorroosion estaeminen lisaeainein

Energy Technology Data Exchange (ETDEWEB)

Aho, M.; Vainikka, P. [VTT, Espoo (Finland); Skrifvars, B.J.; Yrjas, P. [Aabo Akademi, Process Chemistry, Turku (Finland)

2006-12-19

The new superheater protection methods enable combustion of demanding biomass with higher portions than at present. This benefit reduces CO{sub 2} emissions from energy production and the use of demanding biomass in energy production will extend replacing biowaste landfilling with strong CH{sub 4} formation. The results assist also to meet the goals of the use of logging residues in energy production in Finland. (orig.)

Long term testing of materials for tube shielding, stage 2; Laangtidsprovning av tubskyddsmaterial, etapp 2

Energy Technology Data Exchange (ETDEWEB)

Norling, Rikard; Hjoernhede, Anders; Mattsson, Mattias

2012-02-15

Circulating Fluidized Bed (CFB) boilers are commonly used for combustion of biomass and are used to some extent for Waste-to-Energy (WtE) plants as well. The superheaters of the latter are for obvious reasons more prone to suffer from high temperature corrosion caused by the corrosive species in the fuel, mainly chlorides. Frequently the final (hottest) superheater is positioned in the loop seal, where the circulating bed material is returned to the furnace after being separated from the flue gas by a cyclone. The environment in the loop seal is relatively free of chlorides, since these primarily follow the flue gas into the convection pass. Hence, higher steam temperature can be allowed without excessive damage to the final superheater. On the other hand the superheaters, which are located in the convection pass, are more exposed to the corrosive species of the flue gas. Further, they are eroded by particles entrained in the gas flow. Particles and condensing gaseous species are to a large extent deposited on the superheaters, which limits the heat transfer and promotes corrosion. The deposits are regularly removed e.g. by soot blowers. The pressurized steam from soot blowers causes additional erosion damage to that caused by entrained particles. It shall be noted that the actual damage is caused by a combined mechanism of erosion and corrosion denoted erosion-corrosion, which usually results in dramatically accelerated wear. To avoid excessive erosion damage on the superheater tubes the first tube row of each bundle is protected by tube shielding. In its simplest form the shields are made from a steel sheet that has been bent into a semi-circular half-cylinder shell. These shields are attached onto the wind-side of the tubes by hangers. A typical material for tube shielding is the austenitic high temperature resistant stainless steel 253MA. Life of tube shielding depends on numerous factors such as boiler design, superheater location, fuel and operating

Damage distribution and remnant life assessment of a super-heater outlet header used for long time

Energy Technology Data Exchange (ETDEWEB)

Hiroyuki, Okamura [Science Univ. of Tokyo (Japan); Ryuichi, Ohotani [Kyoto Univ. (Japan); Kazuya, Fujii [Japan Power Engineering and Inspection Corp., Tokyo (Japan); Masashi, Nakashiro; Fumio, Takemasa; Hideo, Umaki; Tomiyasu, Masumura [Ishikawajima-Harima Heavy Industries Co. Ltd., Tokyo (Japan)

1998-11-01

This paper presents the results of investigation on evaluating damage distribution to base metals and welded joints in the thickness direction and evaluate damage on ligaments. Thick wall tested sample was the superheater outlet header component long term serviced in high pressure and temperature condition in thermal power plant. The simulate unused steel of component material was made from sample by suitable heat treatment, and the extent of damage was assessed based on a comparison of nondestructive and destructive test results between simulate unused and aged samples. Damage evaluation was also made by FEM structural stress analysis. (orig./MM)

Dynamic performances of wet turbine and steam-separator-superheater and their mathematical simulation as objects of temperature control

International Nuclear Information System (INIS)

Golovach, E.A.

1985-01-01

A mathematical model of a turbine and steam-separator-superheater (SSS) as applied to solution of the tasks of steam temperature regulaton after SSS has been developed. SSS as objects of steam temperature control are considerably less inertial, than intermediate superheaters (IS) of power units in thermal power plants, since for typical SSS and IS considered the duration of transition process according to steam temperature after SSS is 5-10 times loweA than for IS

Soft Sensor for Oxide Scales on the Steam Side of Superheater Tubes under Uneven Circumferential Load

Directory of Open Access Journals (Sweden)

Qing Wei Li

2015-01-01

Full Text Available A soft sensor for oxide scales on the steam side of superheater tubes of utility boiler under uneven circumferential loading is proposed for the first time. First finite volume method is employed to simulate oxide scales growth temperature on the steam side of superheater tube. Then appropriate time and spatial intervals are selected to calculate oxide scales thickness along the circumferential direction. On the basis of the oxide scale thickness, the stress of oxide scales is calculated by the finite element method. At last, the oxide scale thickness and stress sensors are established on support vector machine (SMV optimized by particle swarm optimization (PSO with time and circumferential angles as inputs and oxide scale thickness and stress as outputs. Temperature and stress calculation methods are validated by the operation data and experimental data, respectively. The soft sensor is applied to the superheater tubes of some power plant. Results show that the soft sensor can give enough accurate results for oxide scale thickness and stress in reasonable time. The forecasting model provides a convenient way for the research of the oxide scale failure.

A steam superheater exchanger provided with two coaxial casings and an horizontal axis

International Nuclear Information System (INIS)

Marjollet, Jacques; Palacio, Gerard; Tondeur, Gerard.

1976-01-01

This invention concerns the general lay-out of an horizontal axis separator-superheater for supplying steam to a high power turbine, particularly for a nuclear power station. The invention significantly reduces the length of the pipework connecting the superheated steam outlet and its inlet to the turbine. For this, the outer casing is provided with a coaxial internal annular sleeve in which are housed, one above the other, the separator and the bundle of superheater tubes through which circulates the water emulsion to be separated and steam to be superheated. At the end of its treatment, the superheated steam spreads out in the space between the sleeve and the outer casing from whence it can be drawn off at any point of its periphery, thus making it possible to choose an extraction point as near as possible to the inlet of the turbine to be fed [fr

Fluidized bed combustion of refuse-derived fuel in presence of protective coal ash

Energy Technology Data Exchange (ETDEWEB)

Ferrer, Eduardo [CIRCE, Universidad de Zaragoza, Maria de Luna, 3, Zaragoza (Spain); Aho, Martti [VTT Processes, P.O. Box 1603, 40101 Jyvaeskylae (Finland); Silvennoinen, Jaani; Nurminen, Riku-Ville [Kvaerner Power, P.O.Box 109, FIN-33101 Tampere (Finland)

2005-12-15

Combustion of refuse-derived fuel (RDF) alone or together with other biomass leads to superheater fouling and corrosion in efficient power plants (with high steam values) due to vaporization and condensation of alkali chlorides. In this study, means were found to raise the portion of RDF to 40% enb without risk to boilers. This was done by co-firing RDF with coal and optimizing coal quality. Free aluminum silicate in coal captured alkalies from vaporized alkali chlorides preventing Cl condensation to superheaters. Strong fouling and corrosion were simultaneously averted. Results from 100 kW and 4 MW CFB reactors are reported. (author)

Corrosion behaviour of boiler tube materials during combustion of fuels containing Zn and Pb

Energy Technology Data Exchange (ETDEWEB)

Bankiewicz, D.

2012-11-01

Many power plants burning challenging fuels such as waste-derived fuels experience failures of the superheaters and/or increased waterwall corrosion due to aggressive fuel components already at low temperatures. To minimize corrosion problems in waste-fired boilers, the steam temperature is currently kept at a relatively low level which drastically limits power production efficiency. The elements found in deposits of waste and waste-derived fuels burning boilers that are most frequently associated with high-temperature corrosion are: Cl, S, and there are also indications of Br; alkali metals, mainly K and Na, and heavy metals such as Pb and Zn. The low steam pressure and temperature in waste-fired boilers also influence the temperature of the waterwall steel which is nowadays kept in the range of 300 deg C - 400 deg C. Alkali chloride (KCl, NaCl) induced high-temperature corrosion has not been reported to be particularly relevant at such low material temperatures, but the presence of Zn and Pb compounds in the deposits have been found to induce corrosion already in the 300 deg C - 400 deg C temperature range. Upon combustion, Zn and Pb may react with Cl and S to form chlorides and sulphates in the flue gases. These specific heavy metal compounds are of special concern due to the formation of low melting salt mixtures. These low melting, gaseous or solid compounds are entrained in the flue gases and may stick or condense on colder surfaces of furnace walls and superheaters when passing the convective parts of the boiler, thereby forming an aggressive deposit. A deposit rich in heavy metal (Zn, Pb) chlorides and sulphates increases the risk for corrosion which can be additionally enhanced by the presence of a molten phase. The objective of this study was to obtain better insight into high-temperature corrosion induced by Zn and Pb and to estimate the behaviour and resistance of some boiler superheater and waterwall materials in environments rich in those heavy metals

Degradation of superheater tubes made of austenitic T321H steel after long term service

Energy Technology Data Exchange (ETDEWEB)

Hernas, Adam [Silesian Technical Univ., Katowice (Poland). Faculty of Material Science; Augustyniak, Boleslaw; Chmielewski, Marek [Gdansk Univ. of Technology (Poland). Mechanical Dept.; Sablik, M.J. [Applied Magnetic and Physical Modeling, LLC, San Antonio, TX (United States)

2010-07-01

There are presented results of complementary tests performed for the evaluation of creep damage in austenitic steel grade T321H exploited over 200,000 hours in the secondary superheater part of a power plant boiler. The following techniques have been applied: SEM microscopy, X-ray diffraction, tensile tests, hardness measurements and novel eddy current inspection. The novel eddy current inspection is proposed as a non-destructive method of estimating the creep damage stage of austenite steel boiler tubes after long-term service in power plants. We compare the results provided by the different techniques and discuss the correlations and also point out the problems which need to be addressed in order to elaborate the remaining life assessment of austenitic boiler tubes. (orig.)

Out-pile Test of Double Cladding Fuel Rod Mockups for a Nuclear Fuel Irradiation Test

Energy Technology Data Exchange (ETDEWEB)

Sohn, Jaemin; Park, Sungjae; Kang, Younghwan; Kim, Harkrho; Kim, Bonggoo; Kim, Youngki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2008-05-15

An instrumented capsule for a nuclear fuel irradiation test has been developed to measure fuel characteristics, such as a fuel temperature, internal pressure of a fuel rod, a fuel pellet elongation and a neutron flux during an irradiation test at HANARO. In the future, nuclear fuel irradiation tests under a high temperature condition are expected from users. To prepare for this request, we have continued developing the technology for a high temperature nuclear fuel irradiation test at HANARO. The purpose of this paper is to verify the possibility that the temperature of a nuclear fuel can be controlled at a high temperature during an irradiation test. Therefore we designed and fabricated double cladding fuel rod mockups. And we performed out-pile tests using these mockups. The purposes of a out-pile test is to analyze an effect of a gap size, which is between an outer cladding and an inner cladding, on the temperature and the effect of a mixture ratio of helium gas and neon gas on the temperature. This paper presents the design and fabrication of double cladding fuel rod mockups and the results of the out-pile test.

Probabilistic approach to determining the optimum replacement of a superheater stage in 680 MW coal-fired boiler

Energy Technology Data Exchange (ETDEWEB)

Bos, Robert; Star, Ruud van der [Nuon Power Generation, Amsterdam (Netherlands)

2009-07-01

The boiler of the NUON power plant HW08 that went into operation in 1993 is designed as Benson boiler and mainly fired with hard coal. A creep-related tube failure occurred in the tertiary superheater that had been due to increased wall temperature caused by steam side formation of oxide layers. The theoretical lifetime of the components was calculated with the aid of the results of steam side oxide measurements and condition evaluation of the tertiary superheater with the aid of tube samples. The objective is to establish an operation and maintenance schedule for the desired operating lifetime of 300,000 hours. (orig.)

Improved superheater tubing material - Ti and Nb bearing austenitic steel

International Nuclear Information System (INIS)

Kinoshita, K.; Mimino, T.; Minegishi, I.

1975-01-01

A newly developed 18 Cr-8 Ni stainless steel modified with small amounts of Ti and Nb has considerably high stress-rupture strength and is considered to be suitable for superheater material for power boilers. Data for stress-rupture and creep for long times, the strength of welded joints, the changes of characteristics due to exposure to high temperatures, etc., are presented and discussed. Some investigations after trial services indicate that the experimental data are applicable to actual applications. (author)

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.

Leak testing fuel stored in the ICPP fuel storage basin

International Nuclear Information System (INIS)

Lee, J.L.; Rhodes, D.W.

1977-06-01

Irradiated fuel to be processed at the Idaho Chemical Processing Plant is stored under water at the CPP-603 Fuel Storage Facility. Leakage of radionuclides through breaks in the cladding of some of the stored fuels contaminates the water with radionuclides resulting in radiation exposure to personnel during fuel handling operations and contamination of the shipping casks. A leak test vessel was fabricated to test individual fuel assemblies which were suspected to be leaking. The test equipment and procedures are described. Test results demonstrated that a leaking fuel element could be identified by this method; of the eleven fuel assemblies tested, six were estimated to be releasing greater than 0.5 Ci total radionuclides/day to the basin water

Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

International Nuclear Information System (INIS)

Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho

2014-01-01

A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests

Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

Energy Technology Data Exchange (ETDEWEB)

Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests.

40 CFR 94.108 - Test fuels.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 20 2010-07-01 2010-07-01 false Test fuels. 94.108 Section 94.108... EMISSIONS FROM MARINE COMPRESSION-IGNITION ENGINES Test Procedures § 94.108 Test fuels. (a) Distillate diesel test fuel. (1) The diesel fuels for testing Category 1 and Category 2 marine engines designed to...

The creep life of superheater and reheater tubes under varying pressure conditions in operational boilers

International Nuclear Information System (INIS)

Mizen, D.C.; Plastow, B.

1975-01-01

The first of each manufacturer's 500 MW boilers supplied to the CEGB (Central Electricity Generating Board) have been subjected to an extensive programme of tests for performance optimization and safe operation. Around 250 thermocouples on superheater and reheater tubes have in each case been monitored as part of the exercise. The readings are corrected and used to compute creep rupture damage based on internationally agreed stress rupture data and a simple cumulative damage concept. Comparison of the design creep rupture life and the cumulative life consumed has in several applications been invaluable in influencing operating procedures and arranging tube modifications or replacements, so that loss of generation by creep rupture failure is minimized. (author)

Fuel temperature prediction during high burnup HTGR fuel irradiation test. US-JAERI irradiation test for HTGR fuel

International Nuclear Information System (INIS)

Sawa, Kazuhiro; Fukuda, Kousaku; Acharya, R.

1995-01-01

This report describes the preirradiation thermal analysis of the HRB-22 capsule designed for an irradiation test in a removable beryllium position of the High Flux Isotope Reactor(HFIR) at Oak Ridge National Laboratory. This test is being carried out under Annex 2 of the Arrangement between the U.S. Department of Energy and the Japan Atomic Energy Research Institute on Cooperation in Research and Development regarding High-Temperature Gas-cooled Reactors. The fuel used in the test is an advanced type. The advanced fuel was designed aiming at burnup of about 10%FIMA(% fissions per initial metallic atom) which was higher than that of the first charge fuel for the High Temperature Engineering Test Reactor(HTTR) and was produced in Japan. CACA-2, a heavy isotope and fission product concentration calculational code for experimental irradiation capsules, was used to determine time-dependent fission power for the fuel compacts. The Heat Engineering and Transfer in Nine Geometries(HEATING) code was used to solve the steady-state heat conduction problem. The diameters of the graphite fuel body, which contains the fuel compacts, and of the primary pressure vessel were determined such that the requirements of running the fuel compacts at an average temperature less than 1250degC and of not exceeding a maximum fuel temperature of 1350degC were met throughout the four cycles of irradiation. The detail design of the capsule was carried out based on this analysis. (author)

Thermal load non-uniformity estimation for superheater tube bundle damage evaluation

Directory of Open Access Journals (Sweden)

Naď Martin

2018-01-01

Full Text Available Industrial boiler damage is a common phenomenon encountered in boiler operation which usually lasts several decades. Since boiler shutdown may be required because of localized failures, it is crucial to predict the most vulnerable parts. If damage occurs, it is necessary to perform root cause analysis and devise corrective measures (repairs, design modifications, etc.. Boiler tube bundles, such as those in superheaters, preheaters and reheaters, are the most exposed and often the most damaged boiler parts. Both short-term and long-term overheating are common causes of tube failures. In these cases, the design temperatures are exceeded, which often results in decrease of remaining creep life. Advanced models for damage evaluation require temperature history, which is available only in rare cases when it has been measured and recorded for the whole service life. However, in most cases it is necessary to estimate the temperature history from available operation history data (inlet and outlet pressures and temperatures etc.. The task may be very challenging because of the combination of complex flow behaviour in the flue gas domain and heat transfer phenomena. This paper focuses on estimating thermal load non-uniformity on superheater tubes via Computational Fluid Dynamics (CFD simulation of flue gas flow including heat transfer within the domain consisting of a furnace and a part of the first stage of the boiler.

Fuel motion in overpower tests of metallic integral fast reactor fuel

International Nuclear Information System (INIS)

Rhodes, E.A.; Bauer, T.H.; Stanford, G.S.; Regis, J.P.; Dickerman, C.E.

1992-01-01

In this paper results from hodoscope data analyses are presented for transient overpower (TOP) tests M5, M6, and M7 at the Transient Reactor Test Facility, with emphasis on transient feedback mechanisms, including prefailure expansion at the tops of the fuel pins, subsequent dispersive axial fuel motion, and losses in relative worth of the fuel pins during the tests. Tests M5 and M6 were the first TOP tests of margin to cladding breach and prefailure elongation of D9-clad ternary (U-Pu-Zr) integral fast reactor-type fuel. Test M7 extended these results to high-burnup fuel and also initiated transient testing of HT-9-clad binary (U-Zr) Fast Flux Test Facility driver fuel. Results show significant prefailure negative reactivity feedback and strongly negative feedback from fuel driven to failure

Evaporator Superheat Control With One Temperature Sensor Using Qualitative System Knowledge

DEFF Research Database (Denmark)

Vinther, Kasper; Hillerup Lyhne, Casper; Baasch Sørensen, Erik

2012-01-01

This paper proposes a novel method for superheat control using only a single temperature sensor at the outlet of the evaporator, while eliminating the need for a pressure sensor. An inner loop controls the outlet temperature and an outer control loop provides a reference set point, which is based...... filling of the evaporator, with only one temperature sensor. No a priori model knowledge was used and it is anticipated that the method is applicable on a wide variety of refrigeration systems....

Interim report spent nuclear fuel retrieval system fuel handling development testing

Energy Technology Data Exchange (ETDEWEB)

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

1997-06-01

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

High temperature corrosion in biomass- and waste fired boilers. A status report; Kunskapslaeget betraeffande hoegtemperaturkorrosion i aangpannor foer biobraensle och avfall

Energy Technology Data Exchange (ETDEWEB)

Henderson, P; Ifwer, K; Staalenheim, A; Montgomery, M; Hoegberg, J; Hjoernhede, A

2006-12-15

Many biomass- or waste-fired plants have problems with high temperature corrosion on the furnace walls or at the superheaters, especially if the steam temperature is greater than 500 deg C. An increase in the combustion of waste fuels means that an increasing number of boilers have had problems. Therefore, there is great interest from plant owners to reduce the costs associated with high temperature corrosion. At the same time there exists a considerable driving force towards improving the electrical efficiency of a plant by the use of more advanced steam data. The purpose of the work presented here was to answer three main questions: What can be done to reduce high temperature corrosion with current fuel blends and steam temperatures? How can more waste fuels be burnt without an increased risk for corrosion? What needs to be done to reach higher steam temperatures in the future? The level of knowledge of high temperature corrosion in biomass- and waste-fired boilers has been described and summarised. The following measures are recommended to reduce corrosion in existing plant: Make sure that the fuel is well mixed and improve fuel feeding to obtain a more even spread of the fuel over the cross-section of the boiler. Use combustion technology methods to stabilize the oxygen content of the flue gases near the membrane walls and other heat transfer surfaces. Experiment with additives and/or supplementary fuels which contain sulphur in some form, for example peat. Reduce the flue gas temperature at the superheaters. Review soot-blowing procedures or protect heat transfer surfaces from soot blowers. Evaluate coated membrane wall panels in parts of the furnace that experience the worst corrosion. Test more highly alloyed steels suitable for superheaters and when replacing a superheater change to a more highly alloyed steel. For the future, the following should be considered: The role of sulphur needs to be investigated more and other additives should be investigated

On-line Auto-Tuning of PI Control of the Superheat for a Supermarket Refrigeration System

DEFF Research Database (Denmark)

Yang, Zhenyu; Andersen, Casper; Izadi-Zamanabadi, Roozbeh

2011-01-01

An online PI auto-tuning method is proposed for superheat control for a type of supermarket refrigeration systems. The proposed procedure consists of three serial steps: Step-One uses one of the two proposed empirical methods, namely multi-step method and relay method, for modeling initialization...

CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

International Nuclear Information System (INIS)

Chung, Chang Hwan; Chang, S.K.; Hong, S.D.

1999-04-01

A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs

CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

Energy Technology Data Exchange (ETDEWEB)

Chung, Chang Hwan; Chang, S.K.; Hong, S.D. [and others

1999-04-01

A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs.

Complementary Methods for the Characterization of Corrosion Products on a Plant-Exposed Superheater Tube

DEFF Research Database (Denmark)

Okoro, Sunday Chukwudi; Nießen, Frank; Villa, Matteo

2017-01-01

In this work, complex corrosion products on a superheater tube exposed to biomass firing were characterized by the complementary use of energy-dispersive synchrotron diffraction, electron microscopy, and energy-dispersive X-ray spectroscopy. Non-destructive synchrotron diffraction in transmission......-rich austenite phase to selective removal of Fe and Cr from the alloy, via a KCl-induced corrosion mechanism. Compositional variations were related to diffraction results and revealed a qualitative influence of the spinel cation concentration on the observed diffraction lines.......In this work, complex corrosion products on a superheater tube exposed to biomass firing were characterized by the complementary use of energy-dispersive synchrotron diffraction, electron microscopy, and energy-dispersive X-ray spectroscopy. Non-destructive synchrotron diffraction in transmission...... geometry measuring with a small gauge volume from the sample surface through the corrosion product allowed depth-resolved phase identification and revealed the presence of (Fe,Cr)2O3 and FeCr2O4. This was supplemented by microstructural and elemental analysis correlating the additional presence of a Ni...

Premature failure of dissimilar metal weld joint at intermediate temperature superheater tube

OpenAIRE

Al Hajri, Mohammed; Malik, Anees U.; Meroufel, Abdelkader; Al-Muaili, Fahd

2015-01-01

Dissimilar metal weld (DMW) joint between alloyed steel (AS) and stainless steel (SS) failed at one of intermediate temperature superheater (ITSH) tube in steam/power generation plant boiler. The premature failure was detected after a relatively short time of operation (8 years) where the crack propagated circumferentially from AS side through the ITSH tube. Apart from physical examination, microstructural studies based on optical microscopy, SEM and EDX analysis were performed. The results o...

Spent nuclear fuel retrieval system fuel handling development testing. Final report

International Nuclear Information System (INIS)

Jackson, D.R.; Meeuwsen, P.V.

1997-09-01

Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin, clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge), remove the contents from the canisters and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. This report describes fuel handling development testing performed from May 1, 1997 through the end of August 1997. Testing during this period was mainly focused on performance of a Schilling Robotic Systems' Conan manipulator used to simulate a custom designed version, labeled Konan, being fabricated for K-Basin deployment. In addition to the manipulator, the camera viewing system, process table layout, and fuel handling processes were evaluated. The Conan test manipulator was installed and fully functional for testing in early 1997. Formal testing began May 1. The purposes of fuel handling development testing were to provide proof of concept and criteria, optimize equipment layout, initialize the process definition, and identify special needs/tools and required design changes to support development of the performance specification. The test program was set up to accomplish these objectives through cold (non-radiological) development testing using simulated and prototype equipment

Validation of MCNP: SPERT-D and BORAX-V fuel

International Nuclear Information System (INIS)

Crawford, C.; Palmer, B.

1992-11-01

This report discusses critical experiments involving SPERT-D 1,2 fuel elements and BORAX-V 3-8 fuel which have been modeled and calculations performed with MCNP. MCNP is a Monte Carlo based transport code. For this study continuous-energy nuclear data from the ENDF/B-V cross section library was used. The SPERT-D experiments consisted of various arrays of fuel elements moderated and reflected with either water or a uranyl nitrate solution. Some SPERT-D experiments used cadmium as a fixed neutron poison, while others were poisoned with various concentrations of boron in the moderating/reflecting solution. ne BORAX-V experiments were arrays of either boiling fuel rod assemblies or superheater assemblies, both types of arrays were moderated and reflected with water. In one boiling fuel experiment, two fuel rods were replaced with borated stainless steel poison rods

Verification tests for CANDU advanced fuel

International Nuclear Information System (INIS)

Chung, Chang Hwan; Chang, S.K.; Hong, S.D.

1997-07-01

For the development of a CANDU advanced fuel, the CANFLEX-NU fuel bundles were tested under reactor operating conditions at the CANDU-Hot test loop. This report describes test results and test methods in the performance verification tests for the CANFLEX-NU bundle design. The main items described in the report are as follows. - Fuel bundle cross-flow test - Endurance fretting/vibration test - Freon CHF test - Production of technical document. (author). 25 refs., 45 tabs., 46 figs

Evaluation report on CCTF Core-I reflood tests Cl-2 (Run 11) and Cl-3 (Run 12)

International Nuclear Information System (INIS)

Akimoto, Hajime; Murao, Yoshio

1983-06-01

In order to clarify the effect of the initial superheat of the downcomer wall on the system and the core cooling behaviors during the reflood phase of a PWR-LOCA, two tests were performed with the Cylindrical Core Test Facility (CCTF). One is the superheated wall test (test Cl-2) with the initial superheat of 79K, as in the actual PWR, and the other is the saturated wall test (test Cl-3) without any initial superheat. Through the comparisons of the test results from these two tests, the following conclusions were obtained. (1) The initial superheat of the downcomer wall resulted in the lower downcomer water head as observed in our separate-effect tests for the downcomer water head. (2) The superheat also caused the core inlet subcooling to be decreased, and led to the lower core water head. (3) The mass flow rate through the intact loop was reduced only by 4% by the initial superheat of the downcomer wall because the core water head was reduced as well as the downcomer water head. Whereas the mass flow rate through the broken loop was increased because of the increased pressure drop through the broken cold leg. (4) The difference of the core inlet mass flow rate was small between the superheated and the saturated wall tests. It can be considered that small difference of the core inlet mass flow rate results from the compensation of the decreased mass flow rate through the intact loops by the increased mass flow rate through the broken loop. (5) The main discrepancies of the core cooling and the carry-over behaviors between two CCTF tests, were consistent with those observed in the parametric tests for the core inlet subcooling of the FLECHT LOW FLOODING TEST series. (author)

Future Transient Testing of Advanced Fuels

Energy Technology Data Exchange (ETDEWEB)

Jon Carmack

2009-09-01

The transient in-reactor fuels testing workshop was held on May 4–5, 2009 at Idaho National Laboratory. The purpose of this meeting was to provide a forum where technical experts in transient testing of nuclear fuels could meet directly with technical instrumentation experts and nuclear fuel modeling and simulation experts to discuss needed advancements in transient testing to support a basic understanding of nuclear fuel behavior under off-normal conditions. The workshop was attended by representatives from Commissariat à l'Énergie Atomique CEA, Japanese Atomic Energy Agency (JAEA), Department of Energy (DOE), AREVA, General Electric – Global Nuclear Fuels (GE-GNF), Westinghouse, Electric Power Research Institute (EPRI), universities, and several DOE national laboratories. Transient testing of fuels and materials generates information required for advanced fuels in future nuclear power plants. Future nuclear power plants will rely heavily on advanced computer modeling and simulation that describes fuel behavior under off-normal conditions. TREAT is an ideal facility for this testing because of its flexibility, proven operation and material condition. The opportunity exists to develop advanced instrumentation and data collection that can support modeling and simulation needs much better than was possible in the past. In order to take advantage of these opportunities, test programs must be carefully designed to yield basic information to support modeling before conducting integral performance tests. An early start of TREAT and operation at low power would provide significant dividends in training, development of instrumentation, and checkout of reactor systems. Early start of TREAT (2015) is needed to support the requirements of potential users of TREAT and include the testing of full length fuel irradiated in the FFTF reactor. The capabilities provided by TREAT are needed for the development of nuclear power and the following benefits will be realized by

Future Transient Testing of Advanced Fuels

International Nuclear Information System (INIS)

Carmack, Jon

2009-01-01

The transient in-reactor fuels testing workshop was held on May 4-5, 2009 at Idaho National Laboratory. The purpose of this meeting was to provide a forum where technical experts in transient testing of nuclear fuels could meet directly with technical instrumentation experts and nuclear fuel modeling and simulation experts to discuss needed advancements in transient testing to support a basic understanding of nuclear fuel behavior under off-normal conditions. The workshop was attended by representatives from Commissariat energie Atomique CEA, Japanese Atomic Energy Agency (JAEA), Department of Energy (DOE), AREVA, General Electric - Global Nuclear Fuels (GE-GNF), Westinghouse, Electric Power Research Institute (EPRI), universities, and several DOE national laboratories. Transient testing of fuels and materials generates information required for advanced fuels in future nuclear power plants. Future nuclear power plants will rely heavily on advanced computer modeling and simulation that describes fuel behavior under off-normal conditions. TREAT is an ideal facility for this testing because of its flexibility, proven operation and material condition. The opportunity exists to develop advanced instrumentation and data collection that can support modeling and simulation needs much better than was possible in the past. In order to take advantage of these opportunities, test programs must be carefully designed to yield basic information to support modeling before conducting integral performance tests. An early start of TREAT and operation at low power would provide significant dividends in training, development of instrumentation, and checkout of reactor systems. Early start of TREAT (2015) is needed to support the requirements of potential users of TREAT and include the testing of full length fuel irradiated in the FFTF reactor. The capabilities provided by TREAT are needed for the development of nuclear power and the following benefits will be realized by the

High temperature (salt melt) corrosion tests with ceramic-coated steel

Energy Technology Data Exchange (ETDEWEB)

Schütz, Adelheid [University Bayreuth, Metals and Alloys, Ludwig-Thoma-Str. 36b, D-95447 Bayreuth (Germany); Günthner, Martin; Motz, Günter [University Bayreuth, Ceramic Materials Engineering, L.-Thoma-Str. 36b, D-95447 Bayreuth (Germany); Greißl, Oliver [EnBW Kraftwerke AG, Schelmenwasenstraße 13-15, D-70567 Stuttgart (Germany); Glatzel, Uwe, E-mail: uwe.glatzel@uni-bayreuth.de [University Bayreuth, Metals and Alloys, Ludwig-Thoma-Str. 36b, D-95447 Bayreuth (Germany)

2015-06-01

Thermal recycling of refuse in waste-to-energy plants reduces the problems connected to waste disposal, and is an alternative source of electric energy. However, the combustion process in waste incinerators results in a fast degradation of the steam-carrying superheater steel tubes by corrosive attack and abrasive wear. Higher firing temperatures are used to increase their efficiency but lead to higher corrosion rates. It is more economical to apply protective coatings on the superheater steel tubes than to replace the base material. In-situ tests were conducted in a waste-to-energy plant first in order to identify and quantify all involved corrosive elements. Laboratory scale experiments with salt melts were developed accordingly. The unprotected low-alloyed steel displayed substantial local corrosion. Corrosion was predominant along the grain boundaries of α-ferrite. The corrosion rate was further increased by FeCl{sub 3} and a mixture of HCL and FeCl{sub 3}. Coatings based on pre-ceramic polymers with specific filler particles were engineered to protect superheater tubes. Tests proved their suitability to protect low-alloYed steel tubes from corrosive attack under conditions typical for superheaterS in waste incinerators, rendering higher firing temperatures in waste-to-energy plants possible. - Highlights: • Corrosion wall thickness losses of 400 μm/2 weeks occurred in a waste incinerator. • Abrasion is a major problem on superheater tubes in waste incinerators. • Laboratory salt melt tests can simulate metal corrosion in waste incinerators. • Corrosion protection coatings for steel (temperature: max. 530 °C) were developed. • Higher steam temperatures are possible in WIs with the developed coatings.

HANARO fuel irradiation test(II)

Energy Technology Data Exchange (ETDEWEB)

Sohn, D. S.; Kim, H. R.; Chae, H. T.; Lee, B. C.; Lee, C. S.; Kim, B. G.; Lee, C. B.; Hwang, W

2001-04-01

In order to fulfill the requirement to prove HANARO fuel integrity when irradiated at a power greater than 112.8 kW/m, which was imposed during HANARO licensing, and to verify the irradiation performance of HANARO fuel, the in-pile irradiation test of HANARO fuel has been performed. Two types of test fuel, the un-instrumented Type A fuel for higher burnup irradiation in shorter period than the driver fuel and the instrumented Type B fuel for higher linear heat rate and precise measurement of irradiation conditions, have been designed and fabricated. The test fuel assemblies were irradiated in HANARO. The two Type A fuel assemblies were intended to be irradiated to medium and high burnup and have been discharged after 69.9 at% and 85.5 at% peak burnup, respectively. Type B fuel assembly was intended to be irradiatied at high power with different instrumentations and achieved a maximum power higher than 120 kW/m without losing its integrity and without showing any irregular behavior. The Type A fuel assemblies were cooled for about 6 months and transported to the IMEF(Irradiated Material Examination Facility) for consequent evaluation. Detailed non-destructive and destructive PIE (Post-Irradiation Examination), such as the measurement of burnup distribution, fuel swelling, clad corrosion, dimensional changes, fuel rod bending strength, micro-structure, etc., has been performed. The measured results have been analysed/compared with the predicted performance values and the design criteria. It has been verified that HANARO fuel maintains proper in-pile performance and integrity even at the high power of 120 kw/m up to the high burnup of 85 at%.

Preliminary nuclear design for test MOX Fuel rods

Energy Technology Data Exchange (ETDEWEB)

Joo, Hyung Kook; Kim, Taek Kyum; Jeong, Hyung Guk; Noh, Jae Man; Cho, Jin Young; Kim, Young Il; Kim, Young Jin; Sohn, Dong Seong

1997-10-01

As a part of activity for future fuel development project, test MOX fuel rods are going to be loaded and irradiated in Halden reactor core as a KAERI`s joint international program with Paul Scherrer Institute (PSI). PSI will fabricate test MOX rods with attrition mill device which was developed by KAERI. The test fuel assembly rig contains three MOX rods and three inert matrix rods. One of three MOX rods will be fabricated by BNFL, the other two MOX fuel rods will be manufacturing jointly by KAERI and PSI. Three inert matrix fuel rods will be fabricated with Zr-Y-Er-Pu oxide. Neutronic evaluation was preliminarily performed for test fuel assembly suggested by PSI. The power distribution of test fuel rod in test fuel assembly was analyzed for various fuel rods position in assembly and the depletion characteristic curve for test fuel was also determined. The fuel rods position in test fuel assembly does not effect the rod power distribution, and the proposal for test fuel rods suggested by PSI is proved to be feasible. (author). 2 refs., 13 tabs., 16 figs.

Corrosion evaluation of heat recovery steam generator superheater tube in two methods of testing: Tafel polarization and electrochemical impedance spectroscopy (EIS)

Science.gov (United States)

Santoso, Rio Pudjidarma; Riastuti, Rini

2018-05-01

The purpose of this research is to evaluate the corrosion process which occurs on the water side of Heat Recovery Steam Generator (HRSG) superheater tube. The tube was 13CrMo44 and divided into 3 types of specimen: new tube, used tube (with oxide layer on surface), cleaned-used tube (without oxide layer on surface). The evaluation of corrosion parameters wasperformed using deaerated ultra-high purity water (boiler feed water) in two methods of testing: Tafel polarization and Electrochemical Impedance Spectroscopy (EIS). Tafel polarization was excellent as its capability to show the value of corrosion current and the corrosion rate explicitly, on the other hand, EIS was excellent as its capability to explain for corrosion mechanism on metal interface in detail. Both methods showed that the increase of electrolyte temperature from 25°C to 55°C would increase the corrosion rate with the mechanism of decreasing polarization resistance due to thinning out the passive film thickness and enlarge the area of reduction reaction of cathode. Magnetite oxide scale which is laid on the surface of used tube specimen shows protective nature to reduce the corrosion rate, and clear up this oxide would increase the corrosion rate back as new tube.

Determination of the concentration profile of chemical elements in superheater pipes

International Nuclear Information System (INIS)

Aldape U, F.; Aspiazu F, J.

1986-05-01

This work has for object to determine the profile of concentration of chemical elements at trace level in a superheater pipe of Thermoelectric Plants using the X-ray emission spectroscopy technique induced by protons coming from the Accelerator of the Nuclear Center. In the X-ray detection, a Si Li detector was used. The technique was chosen because it allows a multielemental analysis, of high sensitivity and precision. The results can help to understand the problems that are had in the change of flexibility or of corrosion. This will be from utility to the Federal Electricity Commission (CFE). (Author)

Pre-irradiation testing of experimental fuel elements

International Nuclear Information System (INIS)

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

1979-01-01

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

Accident analysis of HANARO fuel test loop

Energy Technology Data Exchange (ETDEWEB)

Kim, J. Y.; Chi, D. Y

1998-03-01

Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. The HANARO fuel test loop was designed to match the CANDU and PWR fuel operating conditions. The accident analysis was performed by RELAP5/MOD3 code based on FTL system designs and determined the detail engineering specification of in-pile test section and out-pile systems. The accident analysis results of FTL system could be used for the fuel and materials designer to plan the irradiation testing programs. (author). 23 refs., 20 tabs., 178 figs.

Locomotive fuel tank structural safety testing program : passenger locomotive fuel tank jackknife derailment load test.

Science.gov (United States)

2010-08-01

This report presents the results of a passenger locomotive fuel tank load test simulating jackknife derailment (JD) load. The test is based on FRA requirements for locomotive fuel tanks in the Title 49, Code of Federal Regulations (CFR), Part 238, Ap...

HANARO fuel irradiation test (II): revision

Energy Technology Data Exchange (ETDEWEB)

Sohn, D. S.; Kim, H.; Chae, H. T.; Lee, C. S.; Kim, B. G.; Lee, C. B

2001-04-01

In order to fulfill the requirement to prove HANARO fuel integrity when irradiated at a power greater than 112.8 kW/m, which was imposed during HANARO licensing, and to verify the irradiation performance of HANARO fuel, the in-pile irradiation test of HANARO fuel has been performed. Two types of test fuel, the un-instrumented Type A fuel for higher burnup irradiation in shorter period than the driver fuel and the instrumented Type B fuel for higher linear heat rate and precise measurement of irradiation conditions, have been designed and fabricated. The test fuel assemblies were irradiated in HANARO. The two Type A fuel assemblies were intended to be irradiated to medium and high burnup and have been discharged after 69.9 at% and 85.5 at% peak burnup, respectively. Type B fuel assembly was intended to be irradiated at high power with different instrumentations and achieved a maximum power higher than 120 kW/m without losing its integrity and without showing any irregular behavior. The Type A fuel assemblies were cooled for about 6 months and transported to the IMEF(Irradiated Material Examination Facility) for consequent evaluation. Detailed non-destructive and destructive PIE (Post-Irradiation Examination), such as the measurement of burnup distribution, fuel swelling, clad corrosion, dimensional changes, fuel rod bending strength, micro-structure, etc., has been performed. The measured results have been analysed/compared with the predicted performance values and the design criteria. It has been verified that HANARO fuel maintains proper in-pile performance and integrity even at the high power of 120 kw/m up to the high burnup of 85 at%. This report is the revision of KAERI/TR-1816/2001 on the irradiation test for HANARO fuel.

Fuel Retrieval Sub-Project (FRS) Stuck Fuel Station Performance Test Data Report

International Nuclear Information System (INIS)

THIELGES, J.R.

2000-01-01

This document provides the test data report for Stuck Fuel Station Performance Testing in support of the Fuel Retrieval Sub-Project. The stuck fuel station was designed to provide a means of cutting open a canister barrel to release fuel elements, etc

Results of recent KROTOS FCI tests. Alumina vs. corium melts

Energy Technology Data Exchange (ETDEWEB)

Huhtiniemi, I.; Magallon, D.; Hohmann, H. [Commission of the European Communities, Ispra (Italy). Joint Research Center

1998-01-01

Recent results from KROTOS fuel-coolant interaction experiments are discussed. Five tests with alumina were performed under highly subcooled conditions, all of these tests resulted in spontaneous steam explosions. Additionally, four tests were performed at low subcooling to confirm, on one hand, the suppression of spontaneous steam explosions under such conditions and, on the other hand, that such a system is still triggerable using an external initiator. The other test parameters in these alumina tests included the melt superheat and the initial pressure. All the tests in the investigated superheat range (150 K - 750 K) produced a steam explosion and no evidence of the explosion suppression by the elevated initial pressure (in the limited range of 0.1 - 0.375 MPa) was observed in the alumina tests. The corium test series include a test with 3 kg of melt under both subcooled and near saturated conditions at ambient pressure. Two additional tests were performed with subcooled water; one test was performed at an elevated pressure of 0.2 MPa with 2.4 kg of melt and another test with 5.1 kg of melt at ambient pressure. None of these tests with corium produced a propagating energetic steam explosion. However, propagating low energy (about twice the energy of the trigger pulse) events were observed. All corium tests produced significantly higher water level swells during the mixing phase than the corresponding alumina tests. Present experimental evidence suggests that the water depletion in the mixing zone suppresses energetic steam explosions with corium melts at ambient pressure and in the present pour geometry. Processes that could produce such a difference in void generation are discussed. (author)

Recent metal fuel safety tests in TREAT

International Nuclear Information System (INIS)

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

1986-01-01

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

Fuels and materials testing capabilities in Fast Flux Test Facility

International Nuclear Information System (INIS)

Baker, R.B.; Chastain, S.A.; Culley, G.E.; Ethridge, J.L.; Lovell, A.J.; Newland, D.J.; Pember, L.A.; Puigh, R.J.; Waltar, A.E.

1989-01-01

The Fast Flux Test Facility (FFTF) reactor, which started operating in 1982, is a 400 MWt sodium-cooled fast neutron reactor located in Hanford, Washington State, and operated by Westinghouse Hanford Co. under contract with U.S. Department of Energy. The reactor has a wide variety of functions for irradiation tests and special tests, and its major purpose is the irradiation of fuel and material for liquid metal reactor, nuclear reactor and space reactor projects. The review first describes major technical specifications and current conditions of the FFTF reactor. Then the plan for irradiation testing is outlined focusing on general features, fuel pin/assembly irradiation tests, and absorber irradiation tests. Assemblies for special tests include the material open test assembly (MOTA), fuel open test assembly (FOTA), closed loop in-reactor assembly (CLIRA), and other special fuel assemblies. An interim examination and maintenance cell (FFTF/IEM cell) and other hot cells are used for nondestructive/destructive tests and physical/mechanical properties test of material after irradiation. (N.K.)

Failure Investigation of Radiant Platen Superheater Tube of Thermal Power Plant Boiler

Science.gov (United States)

Ghosh, D.; Ray, S.; Mandal, A.; Roy, H.

2015-04-01

This paper highlights a case study of typical premature failure of a radiant platen superheater tube of 210 MW thermal power plant boiler. Visual examination, dimensional measurement and chemical analysis, are conducted as part of the investigations. Apart from these, metallographic analysis and fractography are also conducted to ascertain the probable cause of failure. Finally it has been concluded that the premature failure of the super heater tube can be attributed to localized creep at high temperature. The corrective actions has also been suggested to avoid this type of failure in near future.

Fuel fragmentation data review and separate effects testing

International Nuclear Information System (INIS)

Yueh, Ken. H.; Snis, N.; Mitchell, D.; Munoz-Reja, C.

2014-01-01

A simple alternative test has been developed to study the fuel fragmentation process at loss of coolant accident (LOCA) temperatures. The new test heats a short section of fuel, approximately two pellets worth of material, in a tube furnace open to air. An axial slit is cut in the test sample cladding to reduce radial restraint and to simulate ballooned condition. The tube furnace allows the fuel fragmentation process be observed during the experiment. The test was developed as a simple alternative so large number of tests could be conducted quickly and efficiently to identify key variables that influence fuel fragmentation and to zeroing on the fuel fragmentation burn-up threshold. Several tests were conducted, using fuel materials from fuel rods that were used in earlier integral tests to benchmark and validate the test technique. High burn-up fuel materials known to be above the fragmentation threshold was used to evaluate the fragmentation process as a function of temperature. Even with an axial slit and both ends open, no significant fuel detachment/release was detected until above 750°C. Additional tests were conducted with fuel materials at burn-ups closer to the fuel fragmentation burn-up threshold. Results from these tests indicate a minor power history effect on the fuel fragmentation burn-up threshold. An evaluation of available literature and data generated from this work suggest a fuel fragmentation burn-up threshold between 70 and 75 GWd/MTU. (author)

In-pile tests of HTGR fuel particles and fuel elements

International Nuclear Information System (INIS)

Chernikov, A.S.; Kolesov, V.S.; Deryugin, A.I.

1985-01-01

Main types of in-pile tests for specimen tightness control at the initial step, research of fuel particle radiation stability and also study of fission product release from fuel elements during irradiation are described in this paper. Schemes and main characteristics of devices used for these tests are also given. Principal results of fission gas product release measurements satisfying HTGR demands are illustrated on the example of fuel elements, manufactured by powder metallurgy methods and having TRISO fuel particles on high temperature pyrocarbon and silicon carbide base. (author)

Fast Flux Test Facility fuel and test management: The first 10 years

International Nuclear Information System (INIS)

Bennett, R.A.; Bennett, C.L.; Campbell, L.R.; Dobbin, K.D.; Tang, E.L.

1991-07-01

Core design and fuel and test management have been performed efficiently at the Fast Flux Test Facility. No outages have been extended to adjust core loadings. Development of mixed oxide fuels for advanced liquid metal breeder reactors has been carried out successfully. In fact, the fuel performance is extraordinary. Failures have been so infrequent that further development and refinement of fuel requirements seem appropriate and could lead to a significant reduction in projected electrical busbar costs. The Fast Flux Test Facility is also involved in early metal fuel development tests and appears to be an ideal test bed for any further fuel development or refinement testing. 3 refs., 4 figs., 2 tabs

Fire-Side Corrosion: A Case Study of Failed Tubes of a Fossil Fuel Boiler

OpenAIRE

Asnavandi, Majid; Kahram, Mohaddeseh; Rezaei, Milad; Rezakhani, Davar

2017-01-01

The failures of superheater and reheater boiler tubes operating in a power plant utilizing natural gas or mazut as a fuel have been analysed and the fire-side corrosion has been suggested as the main reason for the failure in boiler tubes. The tubes have been provided by a fossil fuel power plant in Iran and optical and electron microscopy investigations have been performed on the tubes as well as the corrosion products on their surfaces. The results showed that the thickness of the failed tu...

14 CFR 29.965 - Fuel tank tests.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel tank tests. 29.965 Section 29.965 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Powerplant Fuel System § 29.965 Fuel tank tests. (a) Each fuel tank...

14 CFR 27.965 - Fuel tank tests.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel tank tests. 27.965 Section 27.965 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: NORMAL CATEGORY ROTORCRAFT Powerplant Fuel System § 27.965 Fuel tank tests. (a) Each fuel tank...

Improvement of Reactor Fuel Element Heat Transfer by Surface Roughness

Energy Technology Data Exchange (ETDEWEB)

Kjellstroem, B; Larsson, A E

1967-04-15

In heat exchangers with a limited surface temperature such as reactor fuel elements, rough heat transfer surfaces may give lower pumping power than smooth. To obtain data for choice of the most advantageous roughness for the superheater elements in the Marviken reactor, measurements were made of heat transfer and pressure drop in an annular channel with a smooth or rough test rod in a smooth adiabatic shroud. 24 different roughness geometries were tested. The results were transformed to rod cluster geometry by the method of W B Hall, and correlated by the friction and heat transfer similarity laws as suggested by D F Dipprey and R H Sabersky with RMS errors of 12.5 % in the friction factor and 8.1 % in the Stanton number. The relation between the Stanton number and the friction factor could be described by a relation of the type suggested by W Nunner, with a mean error of 3.1 % and an RMS error of 11.6 %. Application of the results to fuel element calculations is discussed, and the great gains in economy which can be obtained with rough surfaces are demonstrated by two examples.

Improvement of Reactor Fuel Element Heat Transfer by Surface Roughness

International Nuclear Information System (INIS)

Kjellstroem, B.; Larsson, A.E.

1967-04-01

In heat exchangers with a limited surface temperature such as reactor fuel elements, rough heat transfer surfaces may give lower pumping power than smooth. To obtain data for choice of the most advantageous roughness for the superheater elements in the Marviken reactor, measurements were made of heat transfer and pressure drop in an annular channel with a smooth or rough test rod in a smooth adiabatic shroud. 24 different roughness geometries were tested. The results were transformed to rod cluster geometry by the method of W B Hall, and correlated by the friction and heat transfer similarity laws as suggested by D F Dipprey and R H Sabersky with RMS errors of 12.5 % in the friction factor and 8.1 % in the Stanton number. The relation between the Stanton number and the friction factor could be described by a relation of the type suggested by W Nunner, with a mean error of 3.1 % and an RMS error of 11.6 %. Application of the results to fuel element calculations is discussed, and the great gains in economy which can be obtained with rough surfaces are demonstrated by two examples

Legacy Vehicle Fuel System Testing with Intermediate Ethanol Blends

Energy Technology Data Exchange (ETDEWEB)

Davis, G. W.; Hoff, C. J.; Borton, Z.; Ratcliff, M. A.

2012-03-01

The effects of E10 and E17 on legacy fuel system components from three common mid-1990s vintage vehicle models (Ford, GM, and Toyota) were studied. The fuel systems comprised a fuel sending unit with pump, a fuel rail and integrated pressure regulator, and the fuel injectors. The fuel system components were characterized and then installed and tested in sample aging test rigs to simulate the exposure and operation of the fuel system components in an operating vehicle. The fuel injectors were cycled with varying pulse widths during pump operation. Operational performance, such as fuel flow and pressure, was monitored during the aging tests. Both of the Toyota fuel pumps demonstrated some degradation in performance during testing. Six injectors were tested in each aging rig. The Ford and GM injectors showed little change over the aging tests. Overall, based on the results of both the fuel pump testing and the fuel injector testing, no major failures were observed that could be attributed to E17 exposure. The unknown fuel component histories add a large uncertainty to the aging tests. Acquiring fuel system components from operational legacy vehicles would reduce the uncertainty.

Advanced steam cycles for light water reactors. Final report

International Nuclear Information System (INIS)

Mitchell, R.C.

1975-07-01

An appraisal of the potential of adding superheat to improve the overall LWR plant cycle performance is presented. The study assesses the economic and technical problems associated with the addition of approximately 500 0 F of superheat to raise the steam temperature to 1000 0 F. The practicality of adding either nuclear or fossil superheat to LWR's is reviewed. The General Electric Company Boiling Water Reactor (BWR) model 238-732 (BWR/6) is chosen as the LWR starting point for this evaluation. The steam conditions of BWR/6 are representative of LWR's. The results of the fossil superheat portion of the evaluation are considered directly applicable to all LWR's. In spite of the potential of a nuclear superheater to provide a substantial boost to the LWR cycle efficiency, nuclear superheat offers little promise of development at this time. There are difficult technical problems to resolve in the areas of superheat fuel design and emergency core cooling. The absence of a developed high integrity, high temperature fuel for operation in the steam/water environment is fundamental to this conclusion. Fossil superheat offers the potential opportunity to utilize fossil fuel supplies more efficiently than in any other mode of central station power generation presently available. Fossil superheat topping cycles evaluated included atmospheric fluidized beds (AFB), pressurized fluidized beds, pressurized furnaces, conventional furnaces, and combined gas/steam turbine cycles. The use of an AFB is proposed as the preferred superheat furnace. Fossil superheat provides a cycle efficiency improvement for the LWR of two percentage points, reduces heat rejection by 15 percent per kWe generated, increases plant electrical output by 54 percent, and burns coal with an incremental net efficiency of approximately 40 percent. This compares with a net efficiency of 36--37 percent which might be achieved with an all-fluidized bed fossil superheat plant design

Design criteria and fabrication in-pile test section of HANARO fuel test loop

Energy Technology Data Exchange (ETDEWEB)

Kim, J. Y.

1997-10-01

Safety state fuel test loop will be equipped in HANARO to obtain the development and betterments of advanced fuel and materials through the irradiation tests. The objective of this study is to determine the design criteria and technical specification of in-pile test section and to specify the manufacturing requirements of in-pile test section. HANARO fuel test loop was designed to meet the CANDU and PWR fuel testing and in-pile section will be manufactured and installed in HANARO. The design criteria and technical specification of in-pile test section could be used the fuel and materials design with for irradiation testing IPS of HANARO fuel test loop. This results will become guidances for the planning and programming of irradiation testing. (author). 12 refs., tabs., figs.

Fuel Cell Development and Test Laboratory | Energy Systems Integration

Science.gov (United States)

Facility | NREL Fuel Cell Development and Test Laboratory Fuel Cell Development and Test Laboratory The Energy System Integration Facility's Fuel Cell Development and Test Laboratory supports fuel cell research and development projects through in-situ fuel cell testing. Photo of a researcher running

Particle fuel bed tests

International Nuclear Information System (INIS)

Horn, F.L.; Powell, J.R.; Savino, J.M.

1985-01-01

Gas-cooled reactors, using packed beds of small diameter coated fuel particles have been proposed for compact, high-power systems. The particulate fuel used in the tests was 800 microns in diameter, consisting of a thoria kernel coated with 200 microns of pyrocarbon. Typically, the bed of fuel particles was contained in a ceramic cylinder with porous metallic frits at each end. A dc voltage was applied to the metallic frits and the resulting electric current heated the bed. Heat was removed by passing coolant (helium or hydrogen) through the bed. Candidate frit materials, rhenium, nickel, zirconium carbide, and zirconium oxide were unaffected, while tungsten and tungsten-rhenium lost weight and strength. Zirconium-carbide particles were tested at 2000 K in H 2 for 12 hours with no visible reaction or weight loss

Technical specification of HANARO fuel test loop

Energy Technology Data Exchange (ETDEWEB)

Kim, J. Y

1998-03-01

The design and installation of the irradiation test facility for verification test of the fuel performance are very important in connection with maximization of the utilization of HANARO. HANARO fuel test loop was designed in accordance with the same code and standards of nuclear power plant because HANARO FTL will be operated the high pressure and temperature same as nuclear power plant operation conditions. The objective of this study is to confirm the operation limit, safety limit, operation condition and checking points of HANARO fuel test loop. This results will become guidances for the planning of irradiation testing and operation of HANARO fuel test loop. (author). 13 refs., 13 tabs., 8 figs.

The role of spent fuel test facilities in the fuel cycle strategy

International Nuclear Information System (INIS)

Huang, S. T.; Gross, D. L.; Snyder, N. W.; Woods, W. D.

1988-01-01

Disposal of commercial spent nuclear fuels in the major industrialized countries may be categorized into two broad approaches: a once-through policy which will dispose of spent fuels and recycle fissile materials. Within reprocess spent fuels and recycle fissile materials. Within each policy, various technical, licensing, institutional and public issues exist. These issues tend to complicate the formulation of an effective and acceptable fuel cycle strategy which will meet various cost, schedule, and legislative constraints. This paper examines overall fuel cycle strategies from the viewpoint of these underlying technical issues and assesses the roles of spent fuel test facilities in the overall fuel cycles steps. Basic functions of such test facilities are also discussed. The main emphasis is placed on the once-through policy although the reprocessing / recycle policy is also discussed. Benefits of utilizing test facilities in the fuel cycle strategies are explored. The results indicate that substantial benefits may be obtained in terms of minimizing programmatic risks, increasing public confidence, and more effective utilization of overall budgetary resources by structuring and highlighting the test facilities as an important element in the overall strategy

Production of A356 aluminum alloy wheels by thixo-forging combined with a low superheat casting process

Directory of Open Access Journals (Sweden)

Wang Shuncheng

2013-09-01

Full Text Available The A356 aluminum alloy wheels were produced by thixo-forging combined with a low superheat casting process. The as-cast microstructure, microstructure evolution during reheating and the mechanical properties of thixo-forged wheels made from the A356 aluminum alloy were studied. The results show that the A356 aluminum alloy round billet with fine, uniform and non-dendritic grains can be obtained when the melt is cast at 635 篊. When the round billet is reheated at 600 篊 for 60 min, the non-dendritic grains are changed into spherical ones and the round billet can be easily thixo-forged into wheels. The tensile strength, yield strength and elongation of the thixo-forged wheels with T6 heat treatment are 327.6 MPa, 228.3 MPa and 7.8%, respectively, which are higher than those of a cast wheel. It is suggested that the thixo-forging combined with the low superheat casting process is an effective technique to produce aluminum alloy wheels with high mechanical properties.

Fabrication of the instrumented fuel rods for the 3-Pin Fuel Test Loop at HANARO

International Nuclear Information System (INIS)

Sohn, Jae Min; Park, Sung Jae; Shin, Yoon Tag; Lee, Jong Min; Ahn, Sung Ho; Kim, Soo Sung; Kim, Bong Goo; Kim, Young Ki; Lee, Ki Hong; Kim, Kwan Hyun

2008-09-01

The 3-Pin Fuel Test Loop(hereinafter referred to as the '3-Pin FTL') facility has been installed at HANARO(High-flux Advanced Neutron Application Reactor) and the 3-Pin FTL is under a test operation. The purpose of this report is to fabricate the instrumented fuel rods for the 3-Pin FTL. The fabrication of these fuel rods was based on experiences and technologies of the instrumented fuel rods for an irradiation fuel capsule. The three instrumented fuel rods of the 3-Pin FTL have been designed. The one fuel rod(180 .deg. ) was designed to measure the centerline temperature of the nuclear fuels and the internal pressure of the fuel rod, and others(60 .deg. and 300 .deg. ) were designed to measure the centerline temperature of the fuel pellets. The claddings were made of the reference material 1 and 2 and new material 1 and 2. And nuclear fuel was used UO 2 (2.0w/o) pellet type with large grain and standard grain. The major procedures of fabrication are followings: (1) the assembling and weld of fuel rods with the pellet mockups and the sensor mockups for the qualification tests, (2) the qualification tests(dimension measurements, tensile tests, metallography examinations and helium leak tests) of weld, (3) the assembling and weld of instrumented fuel rods with the nuclear pellets and the sensors for the irradiation test, and (4) the qualification tests(the helium leak test, the dimensional measurement, electric resistance measurements of sensors) of test fuel rods. Satisfactory results were obtained for all the qualification tests of the instrumented fuel rods for the 3-Pin FTL. Therefore the three instrumented fuel rods for the 3-Pin FTL have been fabricated successfully. These will be installed in the In-Pile Section of 3-Pin FTL. And the irradiation test of these fuel rods is planned from the early next year for about 3 years at HANARO

Energy deposition in NSRR test fuels

International Nuclear Information System (INIS)

Ohnishi, Nobuaki; Tanzawa, Sadamitsu; Tanzawa, Tomio; Kitano, Teruaki; Okazaki, Shuji

1978-02-01

Interpretation of fuel performance data collected during inpile testing in the NSRR requires a knowledge of the energy deposition or enthalpy increase in each sample tested. The report describes the results of absolute measurement of fission products and contents of uranium in irradiated test fuels which were performed to determine the energy deposition. (auth.)

F2 phenomenological test on fuel motion (Interim report)

International Nuclear Information System (INIS)

Palm, R.G.; Fink, C.L.; Stewart, R.R.; Gehl, S.M.; Rothman, A.B.

1976-09-01

TREAT F-series tests are being conducted to provide data on fuel motion at accident power levels from one to about ten times design for use in development of fuel motion models. Test F2 was conducted to evaluate motion of high power fuel in a hypothetical LMFBR unprotected TUC (transient undercooling) accident. Fuel and fuel-boundary conditions following coolant boiling and dryout under TUC conditions are achieved in each F-series test with a single fuel element surrounded by a nuclear heated wall in a dry test capsule. Test F2 was conducted with a low burnup but restructured fuel element to investigate the effect of fuel vapor pressure on fuel motion. Results are presented and discussed

Study on superheat of TiAl melt during cold crucible levitation melting. TiAl no cold crucible levitation yokai ni okeru yoto kanetsudo no kenkyu

Energy Technology Data Exchange (ETDEWEB)

Miwa, K.; Kobayashi, K.; Ninomiya, M. (Government Industrial Research Institute, Nagoya, Nagoya (Japan))

1992-06-20

Investigations were given on effects of test sample weights and sample positions in cold crucibles on superheat of melts when the intermetallic compound TiAl is melted using cold crucible levitation melting process, one of noncontaminated melting processes. The cold crucibles used in the experiment are a water-cooled copper crucible with an inner diameter of 42 mm and a length of 140 mm, into which a column-like ingot sample with an outer diameter of 32 mm (Al containing Ti at 33.5% by mass) was put and melted using the levitation melting. Comparisons and discussions were given on the relationship between sample weights and melt temperatures, the relationship between positions of the inserted samples and melt temperatures, and the state of contamination at melting of casts obtained from the melts resulted from the levitation melting and high-frequency melting poured into respective ceramic dies. Elevating the superheat temperature of the melts requires optimizing the sample weights and positions. Melt temperatures were measured using a radiation thermometer and a thermocouple, and the respective measured values were compared. 7 refs., 4 figs., 1 tab.

30 CFR 36.50 - Tests of fuel tank.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Tests of fuel tank. 36.50 Section 36.50 Mineral... Requirements § 36.50 Tests of fuel tank. The fuel tank shall be inspected and tested to determine whether: (a) It is fuel-tight, (b) the vent maintains atmospheric pressure within the tank, and (c) the vent and...

Fuel Cell Stations Automate Processes, Catalyst Testing

Science.gov (United States)

2010-01-01

Glenn Research Center looks for ways to improve fuel cells, which are an important source of power for space missions, as well as the equipment used to test fuel cells. With Small Business Innovation Research (SBIR) awards from Glenn, Lynntech Inc., of College Station, Texas, addressed a major limitation of fuel cell testing equipment. Five years later, the company obtained a patent and provided the equipment to the commercial world. Now offered through TesSol Inc., of Battle Ground, Washington, the technology is used for fuel cell work, catalyst testing, sensor testing, gas blending, and other applications. It can be found at universities, national laboratories, and businesses around the world.

Performance tests for integral reactor nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Sohn, Dong-Seong; Yim, Jeong-Sik; Lee, Chong-Tak; Kim, Han-Soo; Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Oh, Je-Yong

2006-02-15

An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34{approx}38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc.

Testing Header Component of Electricity Power Industry Boiler

International Nuclear Information System (INIS)

Soedardjo, S.A; Andryansyah, B; Artahari, Dewi; Natsir, Muhammad; Triyadi, Ari; Farokhi

2000-01-01

Testing of header component of Suralaya Unit II electricity power by replication method has been carried out. That header component is cross over pipe which interconnection between Primary and Superheater Outlet Header Secondary Superheater Outlet Header with the operation time over 14 years. The main composition of cross over pipe is 2 1/4 Cr 1 Mo or frequently specified as ferritique steel. The replication testing shown that the damage classification on those cross over pipe in A class based on failure classification from Neubauer and Wedel. Simple calculation in favor of cross over pipe remaining lifetime is about 16.5 years moreover

CANFLEX fuel bundle strength tests (test report)

International Nuclear Information System (INIS)

Chang, Seok Kyu; Chung, C. H.; Kim, B. D.

1997-08-01

This document outlines the test results for the strength tests of the CANFLEX fuel bundle. Strength tests are performed to determine and verify the amount of the bundle shape distortion which is against the side-stops when the bundles are refuelling. There are two cases of strength test; one is the double side-stop test which simulates the normal bundle refuelling and the other is the single side-stop test which simulates the abnormal refuelling. the strength test specification requires that the fuel bundle against the side-stop(s) simulators for this test were fabricated and the flow rates were controlled to provide the required conservative hydraulic forces. The test rig conditions of 120 deg C, 11.2 MPa were retained for 15 minutes after the flow rate was controlled during the test in two cases, respectively. The bundle loading angles of number 13- number 15 among the 15 bundles were 67.5 deg CCW and others were loaded randomly. After the tests, the bundle shapes against the side-stops were measured and inspected carefully. The important test procedures and measurements were discussed as follows. (author). 5 refs., 22 tabs., 5 figs

CANFLEX fuel bundle impact test

International Nuclear Information System (INIS)

Chang, Seok Kyu; Chung, C. H.; Park, J. S.; Hong, S. D.; Kim, B. D.

1997-08-01

This document outlines the test results for the impact test of the CANFLEX fuel bundle. Impact test is performed to determine and verify the amount of general bundle shape distortion and defect of the pressure tube that may occur during refuelling. The test specification requires that the fuel bundles and the pressure tube retain their integrities after the impact test under the conservative conditions (10 stationary bundles with 31kg/s flow rate) considering the pressure tube creep. The refuelling simulator operating with pneumatic force and simulated shield plug were fabricated and the velocity/displacement transducer and the high speed camera were also used in this test. The characteristics of the moving bundle (velocity, displacement, impacting force) were measured and analyzed with the impact sensor and the high speed camera system. The important test procedures and measurement results were discussed as follows. 1) Test bundle measurements and the pressure tube inspections 2) Simulated shield plug, outlet flange installation and bundle loading 3) refuelling simulator, inlet flange installation and sensors, high speed camera installation 4) Perform the impact test with operating the refuelling simulator and measure the dynamic characteristics 5) Inspections of the fuel bundles and the pressure tube. (author). 8 refs., 23 tabs., 13 figs

14 CFR 25.965 - Fuel tank tests.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel tank tests. 25.965 Section 25.965 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System § 25.965 Fuel tank tests. (a) It must be...

RIA tests in CABRI with MOX fuel

International Nuclear Information System (INIS)

Schmitz, F.; Papin, J.; Gonnier, C.

2000-01-01

Three MOX-fuel tests have been successfully performed within the framework of the CABRI REP-Na test program. From the experimental findings which are presently available, no evidence for thermal effects resulting from the heterogeneous nature of the fuel can be given. There are very clear hints however that fission gas effects are enhanced with regard to the behaviour of UO 2 . The clad rupture observed in REP-Na 7 is of different nature than the failures observed in Cabri tests with UO 2 fuel. Failures of UO 2 fuel rods only occurred when the clad mechanical properties were severely affected by the presence of hydride blisters, while in REP-Na 7 a clear indication is made that the loading potential of the MOX fuel pellets was high enough to break a sound cladding. Concerning the transient fuel behaviour after reaching the critical heat-flux under reactor typical conditions (pressure, temperature and flow), no data base could be provided by the tests in the present sodium test loop (as for the UO 2 fuel behaviour). The IPSN project to implement into the Cabri reactor a pressurised water loop which will allow to simulate the complete RIA accident sequence under PWR reactor typical conditions, aims at providing this missing data base. (author)

Test plan for K-Basin fuel handling tools

International Nuclear Information System (INIS)

Bridges, A.E.

1995-01-01

The purpose of this document is to provide the test plan and procedures for the acceptance testing of the handling tools enveloped for the removal of an N-Reactor fuel element from its storage canister in the K-Basins storage pool and insertion into the Single fuel Element Can for subsequent shipment to a Hot Cell for examination. Examination of these N-Reactor fuel elements is part of the overall characterization effort. New hand tools were required since previous fuel movement has involved grasping the fuel in a horizontal position. The 305 Building Cold Test Facility will be used to conduct the acceptance testing of the Fuel Handling Tools. Upon completion of this acceptance testing and any subsequent training of operators, the tools will be transferred to the 105 KW Basin for installation and use

LEU WWR-M2 fuel assemblies burnable test

International Nuclear Information System (INIS)

Kirsanov, G.A.; Konoplev, K.A.; Pikulik, R.G.; Sajkov, Yu. P.; Tchmshkyan, D.V.; Tedoradze, L.V.; Zakharov, A.S.

2000-01-01

The results of in-pile irradiation tests of LEU WWR-M2 fuel assemblies with reduced enrichment of fuel are submitted in the report. The tests are made according to the Russian Program on Reduced Enrichment for Research and Test Reactors (RERTR). United States Department of Energy and the Ministry of Atomic Energy of Russian Federation jointly fund this Program. The irradiation tests of 5 WWR-M2 experimental assemblies are carried out at WWR-M reactor of the Petersburg Nuclear Physics Institute (PNPI). The information on assembly design and technique of irradiation tests is presented. In the irradiation tests the integrity of fuel assemblies is periodically measured. The report presents the data for the integrity maintained during the burnup of 5 fuel assemblies up to 45%. These results demonstrate the high reliability of the experimental fuel assemblies within the guaranteed burnup limits specified by the manufacturer. The tests are still in progress; it is planned to test and analyze the change in integrity for burnup of up to 70% - 75% or more. LEU WWR-M2 fuel assemblies are to be offered for export by their Novosibirsk manufacturer. Currently, HEU WWR-M2 fuel assemblies are used in Hungary, Ukraine and Vietnam. LEU WWR-M2 fuel assemblies were designed as a possible replacement for the HEU WWR-M2 fuel assemblies in those countries, but their use can be extended to other research reactors. (author)

Maximum thermal loading test of BWR fuel assembly

International Nuclear Information System (INIS)

Nakajima, Yoshitaka; Yoshimura, Kunihiro; Nakamura, Satoshi; Ishizuka, Takao.

1987-01-01

Various proving tests on the reliability of nuclear power plants have been conducted at the Nuclear Power Engineering Test Center and at the Japan Power Plant Engineering and Inspection Corporation. The tests were initiated at the request of the Ministry of International Trade and Industry (MITI). Toshiba undertook one of the proving tests on the reliability of nuclear fuel assembly; the maximum thermal loading test of BWR fuel assembly from the Nuclear Power Engineering Test Center. These tests are part of the proving tests mentioned above, and their purpose is to confirm the reliability of the thermal hydraulic engineering techniques. Toshiba has been engaged for the past nine years in the design, fabrication and testing of the equipment. For the project, a test model fuel assembly was used to measure the critical power of the BWR fuel assembly and the void and fluidity of the coolant. From the test results, it has been confirmed that the heat is transferred safely from the fuel assembly to the coolant in the BWR nuclear power plant. In addition, the propriety and reliability of the thermal hydraulic engineering techniques for the fuel assembly have been proved. (author)

Fuel Economy Testing and Data

Science.gov (United States)

EPA’s Fuel Economy pages provide information on current standards and how federal agencies work to enforce those laws, testing for national Corporate Average Fuel Economy or CAFE standards, and what you can do to reduce your own vehicle emissions.

Test plan for spent fuel cladding containment credit tests

International Nuclear Information System (INIS)

Wilson, C.N.

1983-11-01

Lawrence Livermore National Laboratory has chosen Westinghouse Hanford Company as a subcontractor to assist them in determining the requirements for successful disposal of spent fuel rods in the proposed Nevada Test Site repository. An initial scoping test, with the objective of determining whether or not the cladding of a breached fuel rod can be given any credit as an effective barrier to radionuclide release, is described in this test plan. 8 references, 2 figures, 4 tables

Status of fuel irradiation tests in HANARO

International Nuclear Information System (INIS)

Kim, Hark Rho; Lee, Choong Sung; Lee, Kye Hong; Jun, Byung Jin; Lee, Ji Bok

1999-01-01

Since 1996 after finishing the long-term operational test, HANARO (High-Flux Advanced Neutron Application Reactor) has been extensively used for material irradiation tests, beam application research, radioisotope production and neutron activation analysis. This paper presents the fuel irradiation test activities which are now conducted or have been finished in HANARO. KAERI developed LEU fuel using an atomization method for the research reactors. Using this LEU, we have set up and conducted three irradiation programs: (1) medium power irradiation test using a short-length mini-assembly made of 3.15 gU/cc U 3 Si, (2) high power irradiation tests using full-length test assemblies made of 3.15 gU/cc U 3 Si, and (3) irradiation test using a short-length mini-plate made of 4.8 gU/cc U 3 Si 2 . DUPIC (Direct Use of spent PWR fuels in CANDU Reactors) simulation fuel pellets, of which compositions are very similar to DUPIC pellets to keep the similarity in the thermo-mechanical property, were developed. Three mini-elements including 5 pellets each were installed in a capsule. This capsule has been irradiated for 2 months and unloaded from the HANARO core at the end of September 1999. Another very important test is the HANARO fuel qualification program at high power, which is required to resolve the licensing issue. This test is imposed on the HANARO operation license due to insufficient test data under high power environment. To resolve this licensing issue, we have been carrying out the required irradiation tests and PIE (Post-irradiation Examination) tests. Through this program, it is believed that the resolution of the licensing issue is achieved. In addition to these programs, several fuel test plans are under way. Through these vigorous activities of fuel irradiation test programs, HANARO is sure to significantly contribute to the national nuclear R and D programs. (author)

Irradiated test fuel shipment plan for the LWR MOX fuel irradiation test project

International Nuclear Information System (INIS)

Shappert, L.B.; Dickerson, L.S.; Ludwig, S.B.

1998-01-01

This document outlines the responsibilities of DOE, DOE contractors, the commercial carrier, and other organizations participating in a shipping campaign of irradiated test specimen capsules containing mixed-oxide (MOX) fuel from the Idaho National Engineering and Environmental Laboratory (INEEL) to the Oak Ridge National Laboratory (ORNL). The shipments described here will be conducted according to applicable regulations of the US Department of Transportation (DOT), US Nuclear Regulatory Commission (NRC), and all applicable DOE Orders. This Irradiated Test Fuel Shipment Plan for the LWR MOX Fuel Irradiation Test Project addresses the shipments of a small number of irradiated test specimen capsules and has been reviewed and agreed to by INEEL and ORNL (as participants in the shipment campaign). Minor refinements to data entries in this plan, such as actual shipment dates, exact quantities and characteristics of materials to be shipped, and final approved shipment routing, will be communicated between the shipper, receiver, and carrier, as needed, using faxes, e-mail, official shipping papers, or other backup documents (e.g., shipment safety evaluations). Any major changes in responsibilities or data beyond refinements of dates and quantities of material will be prepared as additional revisions to this document and will undergo a full review and approval cycle

MTR fuel testing in BR2

International Nuclear Information System (INIS)

Jacquet, P.; Verwimp, A.; Wirix, S.

2000-01-01

New fuel design for MTR 's requires to be qualified under representative conditions, that is geometry, neutron spectrum, heat flux and thermo hydraulic conditions. An irradiation device for fuel plates has been designed to derive the maximum benefit from the BR2 irradiation capacities. The fuel plates can be easily extracted from their support during a shutdown to undergo additional tests. One of these tests is the measurement of the thickness changes along the fuel plate. To that purpose, a facility in the reactor water pool has been designed to measure the fuel swelling with an accuracy of 5 μm using inductive probes. At SCK-CEN, the full range of destructive and non-destructive PIE can be performed, including γ-scanning, wet sipping, surface examination and other methods. (author)

Test requirements of locomotive fuel tank blunt impact tests

Science.gov (United States)

2013-10-15

The Federal Railroad Administrations Office of Research : and Development is conducting research into passenger : locomotive fuel tank crashworthiness. A series of impact tests : are planned to measure fuel tank deformation under two types : of dy...

Stress corrosion cracking in superheater and reheater austenitic tubing

Energy Technology Data Exchange (ETDEWEB)

Dooley, R. Barry [Structural Integrity Associates, Inc., Charlotte, NC (United States); Bursik, Albert [PowerPlant Chemistry GmbH, Neulussheim (Germany)

2011-02-15

University 101 courses are typically designed to help incoming first-year undergraduate students to adjust to the university, develop a better understanding of the college environment, and acquire essential academic success skills. Why are we offering a special Boiler and HRSG Tube Failures PPChem 101? The answer is simple, yet very conclusive: - There is a lack of knowledge on the identification of tube failure mechanisms and for the implementation of adequate counteractions in many power plants, particularly at industrial power and steam generators. - There is a lack of knowledge to prevent repeat tube failures. The vast majority of BTF/HTF have been, and continue to be, repeat failures. It is hoped that the information about the failure mechanisms of BTF supplied in this course will help to put plant engineers and chemists on the right track. The major goal of this course is the avoidance of repeat BTF. This eights lesson is focused on Stress Corrosion Cracking in Superheater and Reheater Austenitic Tubing. (orig.)

Vibration test and endurance test for HANARO 36-element fuel assembly

International Nuclear Information System (INIS)

Ryu, Jeong Soo; Kim, Heon ll; Chung, Heung June

1998-06-01

Vibration test and endurance test for HANARO DU (depleted uranium) 36-element fuel assembly which was fabricated by KAERI were carried out based on the HANARO operation conditions. The endurance test of 22 days was added to the previous 18 days test. The vibration test was performed at various flow rates. Vibration frequency for the 36-element fuel assembly is between 11 to 14.5 Hz. And the maximum vibration displacement is less than 100 μm. From the endurance test result, it can be concluded that the appreciable fretting wear for the 36-element fuel assembly and the hexagonal flow tube was not observed. (author). 4 refs., 5 tabs., 29 figs

Measuring deformation of Fuel pin in a Nuclear Fuel Test Rig

Energy Technology Data Exchange (ETDEWEB)

Heo, S. H.; Yang, T. H.; Hong, J. T.; Joung, C. Y.; Ahn, S. H.; Jang, S. Y.; Kim, J. H. [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In this study, an LVDT core for measuring the longitudinal displacement of fuel pellets and clad was designed and produced. A signal processing method for the prepared core was investigated. The Nuclear Fuel Test Rig is used to observe changes in the characteristics of the fuel according to the neutron irradiation at HANARO (High-flux Advanced Neutron Application Reactor), which is a research reactor. Which are the strain and internal temperature of the irradiated nuclear fuel and the internal pressure of fuel due to fission gas, the characteristics of the fuel are measured using various sensors such as a thermocouple, SPND and LVDT. In this study, two shaped LVDT (Linear Variable Differential Transformer) cores for displacement measurements were designed and manufactured in order to measure the displacement of a fuel pellet and cladding tube using LVDT sensors for measuring electrical signals by converting the physical variation such as the force and displacement into a linear motion. In addition, signals from the manufactured LVDT sensor were collected and calibrated. Moreover, a method for obtaining the displacement in the core according to the sensing signal was planned. A derived equation can used to predict the change in the position of core. A following study should be conducted to test the output signal and real variation of out-pile system. For further work, a performance verification is required for an in-pile irradiation test.

Thermodynamic and heat transfer analysis of heat recovery from engine test cell by Organic Rankine Cycle

Science.gov (United States)

Shokati, Naser; Mohammadkhani, Farzad; Farrokhi, Navid; Ranjbar, Faramarz

2014-12-01

During manufacture of engines, evaluation of engine performance is essential. This is accomplished in test cells. During the test, a significant portion of heat energy released by the fuel is wasted. In this study, in order to recover these heat losses, Organic Rankine Cycle (ORC) is recommended. The study has been conducted assuming the diesel oil to be composed of a single hydrocarbon such as C12H26. The composition of exhaust gases (products of combustion) have been computed (and not determined experimentally) from the stoichiometric equation representing the combustion reaction. The test cell heat losses are recovered in three separate heat exchangers (preheater, evaporator and superheater). These heat exchangers are separately designed, and the whole system is analyzed from energy and exergy viewpoints. Finally, a parametric study is performed to investigate the effect of different variables on the system performance characteristics such as the ORC net power, heat exchangers effectiveness, the first law efficiency, exergy destruction and heat transfer surfaces. The results of the study show that by utilizing ORC, heat recovery equivalent to 8.85 % of the engine power is possible. The evaporator has the highest exergy destruction rate, while the pump has the lowest among the system components. Heat transfer surfaces are calculated to be 173.6, 58.7, and 11.87 m2 for the preheater, evaporator and superheater, respectively.

Superheater fireside corrosion mechanisms in MSWI plants: Lab-scale study and on-site results

Energy Technology Data Exchange (ETDEWEB)

Brossard, J.M.; Chaucherie, X.; Nicol, F. [Veolia Environnement R and D, Zone Portuaire de Limay, 291 Avenue Dreyfous Ducas, Limay 78520 (France); Diop, I. [Veolia Environnement R and D, Zone Portuaire de Limay, 291 Avenue Dreyfous Ducas, Limay 78520 (France); Institut Jean Lamour, departement Chimie et physique des solides et des surfaces, UMR 7198 CNRS - Universite Henri Poincare Nancy 1, Vandoeuvre-Les-Nancy (France); Rapin, C.; Vilasi, M. [Institut Jean Lamour, departement Chimie et physique des solides et des surfaces, UMR 7198 CNRS - Universite Henri Poincare Nancy 1, Vandoeuvre-Les-Nancy (France)

2011-06-15

Combustion of municipal waste generates highly corrosive gases (HCl, SO{sub 2}, NaCl, KCl, and heavy metals chlorides) and ashes containing alkaline chlorides and sulfates. Currently, corrosion phenomena are particularly observed on superheater's tubes. Corrosion rates depend mainly on installation design, operating conditions i.e., gas and steam temperature and velocity of the flue gas containing ashes. This paper presents the results obtained using an innovative laboratory-scale corrosion unit, which simulates MSWI (Municipal Solid Waste Incineration) boilers conditions characterized by a temperature gradient at the metal tube in the presence of corrosive gases and ashes. The presented corrosion tests were realized on carbon steel at fixed metal temperature (400 C). The influence of the flue gas temperature, synthetic ashes composition, and flue gas flow pattern were investigated. After corrosion test, cross sections of tube samples were characterized to evaluate thickness loss and estimate corrosion rate while the elements present in corrosion layers were analyzed. Corrosion tests were carried out twice in order to validate the accuracy and reproducibility of results. First results highlight the key role of molten phase related to the ash composition and flue gas temperature as well as the deposit morphology, related to the flue gas flow pattern, on the mechanisms and corrosion rates. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

A comprehensive in-pile test of PWR fuel bundle

Energy Technology Data Exchange (ETDEWEB)

Kang Rixin; Zhang Shucheng; Chen Dianshan (Academia Sinica, Beijing (China). Inst. of Atomic Energy)

1991-02-01

An in-pile test of PWR fuel bundle has been conducted in HWRR at IAE of China. This paper describes the structure of the test bundle (3x3-2), fabrication process and quality control of the fuel rod, irradiation conditions and the main Post Irradiation Examination (PIE) results. The test fuel bundle was irradiated under the PWR operation and water chemistry conditions with an average linear power of 381 W/cm and reached an average burnup of 25010 MWd/tU of the fuel bundle. After the test, destructive and non-destructive examination of the fuel rods was conducted at hot laboratories. The fission gas release was 10.4-23%. The ridge height of cladding was 3 to 8 {mu}m. The hydrogen content of the cladding was 80 to 140 ppm. The fuel stack height was increased by 2.9 to 3.3 mm. The relative irradiation growth was about 0.11 to 0.17% of the fuel rod length. During the irradiation test, no fuel rod failure or other abnormal phenomena had been found by the on-line fuel failure monitoring system of the test loop and water sampling analysis. The structure of the test fuel assembly was left undamaged without twist and detectable deformation. (orig.).

Irradiation test and performance evaluation of DUPIC fuel

International Nuclear Information System (INIS)

Yang, Myung Seung; Song, K. C.; Moon, J. S.

2002-05-01

The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase II R and D. In order to fulfil this objectives, irradiation test of DUPIC fuel was carried out in HANARO using the non-instrumented and SPND-instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase II are summarized as follows : - Performance evaluation of DUPIC fuel via irradiation test in HANARO - Post irradiation examination of irradiated fuel and performance analysis - Development of DUPIC fuel performance code (modified ELESTRES) considering material properties of DUPIC fuel - Irradiation behavior and integrity assessment under the design power envelope of DUPIC fuel - Foundamental technology development of thermal/mechanical performance evaluation using ANSYS (FEM package)

Microscopical investigation of steamside oxide on X20CrMoV121 superheater tubes

DEFF Research Database (Denmark)

Montgomery, Melanie; Hansson, Anette Nørgaard; Jensen, Søren A.

2011-01-01

X20CrMoV121 is a 12%Cr martensitic steel which has been used in power plants in Europe for many decades. Superheater tubes exposed for various durations up to 135,000 hours in power plants in Denmark at steam temperatures varying from 450 to 575°C were investigated. Light optical and scanning ele...... electron microscopy was used to investigate steamside oxide morphologies. At all temperatures there is a double layered oxide, however, the inner:outer oxide thickness is not always equal. At the lower steam temperature range of...

CANFLEX fuel bundle cross-flow endurance test (test report)

International Nuclear Information System (INIS)

Hong, Sung Deok; Chung, C. H.; Chang, S. K.; Kim, B. D.

1997-04-01

As part of the normal refuelling sequence of CANDU nuclear reactor, both new and irradiated bundles can be parked in the cross-flow region of the liner tubes. This situation occurs normally for a few minutes. The fuel bundle which is subjected to the cross-flow should be capable of withstanding the consequences of cross flow for normal periods, and maintain its mechanical integrity. The cross-flow endurance test was conducted for CANFLEX bundle, latest developed nuclear fuel, at CANDU-Hot Test Loop. The test was carried out during 4 hours at the inlet cross-flow region. After the test, the bundle successfully met all acceptance criteria after the 4 hours cross-flow test. (author). 2 refs., 3 tabs

CANFLEX fuel bundle cross-flow endurance test (test report)

Energy Technology Data Exchange (ETDEWEB)

Hong, Sung Deok; Chung, C. H.; Chang, S. K.; Kim, B. D.

1997-04-01

As part of the normal refuelling sequence of CANDU nuclear reactor, both new and irradiated bundles can be parked in the cross-flow region of the liner tubes. This situation occurs normally for a few minutes. The fuel bundle which is subjected to the cross-flow should be capable of withstanding the consequences of cross flow for normal periods, and maintain its mechanical integrity. The cross-flow endurance test was conducted for CANFLEX bundle, latest developed nuclear fuel, at CANDU-Hot Test Loop. The test was carried out during 4 hours at the inlet cross-flow region. After the test, the bundle successfully met all acceptance criteria after the 4 hours cross-flow test. (author). 2 refs., 3 tabs.

Fuel integrity project: analysis of light water reactor fuel rods test results

International Nuclear Information System (INIS)

Dallongeville, M.; Werle, J.; McCreesh, G.

2004-01-01

BNFL Nuclear Sciences and Technology Services and COGEMA LOGISTICS started in the year 2000 a joint project known as FIP (Fuel Integrity Project) with the aim of developing realistic methods by which the response of LWR fuel under impact accident conditions could be evaluated. To this end BNFL organised tests on both unirradiated and irradiated fuel pin samples and COGEMA LOGISTICS took responsibility for evaluating the test results. Interpretation of test results included simple mechanical analysis as well as simulation by Finite Element Analysis. The first tests that were available for analysis were an irradiated 3 point bending commissioning trial and a lateral irradiated hull compression test, both simulating the loading during a 9 m lateral regulatory drop. The bending test span corresponded roughly to a fuel pin intergrid distance. The outcome of the test was a failure starting at about 35 mm lateral deflection and a few percent of total deformation. Calculations were carried out using the ANSYS code employing a shell and brick model. The hull lateral compaction test corresponds to a conservative compression by neighbouring pins at the upper end of the fuel pin. In this pin region there are no pellets inside. The cladding broke initially into two and later into four parts, all of which were rather similar. Initial calculations were carried out with LS-DYNA3D models. The models used were optimised in meshing, boundary conditions and material properties. The calculation results compared rather well with the test data, in particular for the detailed ANSYS approach of the 3 point bending test, and allowed good estimations of stresses and deformations under mechanical loading as well as the derivation of material rupture criteria. All this contributed to the development of realistic numerical analysis methods for the evaluation of LWR fuel rod behaviour under both normal and accident transport conditions. This paper describes the results of the 3 point bending

Fuel integrity project: analysis of light water reactor fuel rods test results

Energy Technology Data Exchange (ETDEWEB)

Dallongeville, M.; Werle, J. [COGEMA Logistics (AREVA Group) (France); McCreesh, G. [BNFL Nuclear Sciences and Technology Services (United Kingdom)

2004-07-01

BNFL Nuclear Sciences and Technology Services and COGEMA LOGISTICS started in the year 2000 a joint project known as FIP (Fuel Integrity Project) with the aim of developing realistic methods by which the response of LWR fuel under impact accident conditions could be evaluated. To this end BNFL organised tests on both unirradiated and irradiated fuel pin samples and COGEMA LOGISTICS took responsibility for evaluating the test results. Interpretation of test results included simple mechanical analysis as well as simulation by Finite Element Analysis. The first tests that were available for analysis were an irradiated 3 point bending commissioning trial and a lateral irradiated hull compression test, both simulating the loading during a 9 m lateral regulatory drop. The bending test span corresponded roughly to a fuel pin intergrid distance. The outcome of the test was a failure starting at about 35 mm lateral deflection and a few percent of total deformation. Calculations were carried out using the ANSYS code employing a shell and brick model. The hull lateral compaction test corresponds to a conservative compression by neighbouring pins at the upper end of the fuel pin. In this pin region there are no pellets inside. The cladding broke initially into two and later into four parts, all of which were rather similar. Initial calculations were carried out with LS-DYNA3D models. The models used were optimised in meshing, boundary conditions and material properties. The calculation results compared rather well with the test data, in particular for the detailed ANSYS approach of the 3 point bending test, and allowed good estimations of stresses and deformations under mechanical loading as well as the derivation of material rupture criteria. All this contributed to the development of realistic numerical analysis methods for the evaluation of LWR fuel rod behaviour under both normal and accident transport conditions. This paper describes the results of the 3 point bending

Drop-in capsule testing of plutonium-based fuels in the Advanced Test Reactor

International Nuclear Information System (INIS)

Chang, G.S.; Ryskamp, J.M.; Terry, W.K.; Ambrosek, R.G.; Palmer, A.J.; Roesener, R.A.

1996-09-01

The most attractive way to dispose of weapons-grade plutonium (WGPu) is to use it as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PuO[sub 2]) mixed with urania (UO[sub 2]). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. The proposed weapons-grade MOX fuel is unusual, even relative to ongoing foreign experience with reactor-grade MOX power reactor fuel. Some demonstration of the in- reactor thermal, mechanical, and fission gas release behavior of the prototype fuel will most likely be required in a limited number of test reactor irradiations. The application to license operation with MOX fuel must be amply supported by experimental data. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory (INEL) is capable of playing a key role in the irradiation, development, and licensing of these new fuel types. The ATR is a 250- MW (thermal) LWR designed to study the effects of intense radiation on reactor fuels and materials. For 25 years, the primary role of the ATR has been to serve in experimental investigations for the development of advanced nuclear fuels. Both large- and small-volume test positions in the ATR could be used for MOX fuel irradiation. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. Furthermore, these data can be obtained more quickly by using ATR instead of testing in a commercial LWR. Our previous work in this area has demonstrated that it is technically feasible to perform MOX fuel testing in the ATR. This report documents our analyses of sealed drop-in capsules containing plutonium-based test specimens placed in various ATR positions

Characteristic test technology for PWR fuel and its components

Energy Technology Data Exchange (ETDEWEB)

Kim, Dae Ho; Lee, Chan Bock; Bang, Je Gun; Jung, Yeon Ho; Jeong, Yong Hwan; Park, Sang Yoon; Kim, Kyeng Ho; Nam, Cheol; Baek, Jong Hyuk; Lee, Myung Ho; Choi, Byoung Kwon; Song, Kun Woo; Kang, Ki Won; Kim, Keon Sik; Kim, Jong Hun; Kim, Young Min; Yang, Jae Ho; Song, Kee Nam; Kim, Hyung Kyu; Kang, Heung Seok; Yoon, Kyung Ho; Chun, Tae Hyun; In, Wang Kee; Oh, Dong Seok [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-01-01

Characteristic tests of fuel assembly and its components being developed in the Advanced LWR Fuel Development Project supported by the mid-long term nuclear R and D program are described in this report. Performance verification of fuel and its components by the characteristic tests are essential to their development. Fuel components being developed in the Advanced LWR Fuel Development Project are zirconium alloy cladding, UO{sub 2} and burnable absorber pellets, spacer grid and top and bottom end pieces. Detailed test plans for those fuel components are described in this report, and test procedures of cladding and pellet are also described in the Appendix. Examples of the described tests are in- and out-of- pile corrosion and mechanical tests such as creep and burst tests for the cladding, in-pile capsule and ramp tests for the pellet, mechanical tests such as strength and vibration, and thermal-hydraulic tests such as pressure drop and critical heat flux for the spacer grid and top and bottom end pieces. It is expected that this report could be used as the standard reference for the performance verification tests in the development of LWR fuel and its components. 11 refs., 9 figs., 2 tabs. (Author)

Parametric Sensitivity Tests- European PEM Fuel Cell Stack Test Procedures

DEFF Research Database (Denmark)

Araya, Samuel Simon; Andreasen, Søren Juhl; Kær, Søren Knudsen

2014-01-01

performed based on test procedures proposed by a European project, Stack-Test. The sensitivity of a Nafion-based low temperature PEMFC stack’s performance to parametric changes was the main objective of the tests. Four crucial parameters for fuel cell operation were chosen; relative humidity, temperature......As fuel cells are increasingly commercialized for various applications, harmonized and industry-relevant test procedures are necessary to benchmark tests and to ensure comparability of stack performance results from different parties. This paper reports the results of parametric sensitivity tests......, pressure, and stoichiometry at varying current density. Furthermore, procedures for polarization curve recording were also tested both in ascending and descending current directions....

Effect of reinforcement amount, mold temperature, superheat, and mold thickness on fluidity of in-situ Al-Mg2Si composites

Directory of Open Access Journals (Sweden)

Reza Vatankhah Barenji

2018-01-01

Full Text Available In the present study, the effects of mold temperature, superheat, mold thickness, and Mg2Si amount on the fluidity of the Al-Mg2Si as-cast in-situ composites were investigated using the mathematical models. Composites with different amounts of Mg2Si were fabricated, and the fluidity and microstructure of each were then analyzed. For this purpose, the experiments were designed using a central composite rotatable design, and the relationship between parameters and fluidity were developed using the response surface method. In addition, optical and scanning electron microscopes were used for microstructural observation. The ANOVA shows that the mathematical models can predict the fluidity accurately. The results show that by increasing the mold temperature from 25 °C to 200 °C, superheat from 50 °C to 250 °C, and thickness from 3 mm to 12 mm, the fluidity of the composites decreases, where the mold thickness is more effective than other factors. In addition, the higher amounts of Mg2Si in the range from 15wt.% to 25wt.% lead to the lower fluidity of the composites. For example, when the mold temperature, superheat, and thickness are respectively 100 °C, 150 °C, and 7 mm, the fluidity length is changed in the range of 11.9 cm to 15.3 cm. By increasing the amount of Mg2Si, the morphology of the primary Mg2Si becomes irregular and the size of primary Mg2Si is increased. Moreover, the change of solidification mode from skin to pasty mode is the most noticeable microstructural effect on the fluidity.

Influence of surface roughness and melt superheat on HDA process to form a tritium permeation barrier on RAFM steel

Energy Technology Data Exchange (ETDEWEB)

Purushothaman, J. [B.S. Abdur Rahman University, Chennai 600048 (India); MTD, MMG, IGCAR, Kalpakkam 603102 (India); Ramaseshan, R., E-mail: seshan@igcar.gov.in [TFCS, SND, MSG, IGCAR, Kalpakkam 603102 (India); Albert, S.K. [MTD, MMG, IGCAR, Kalpakkam 603102 (India); Rajendran, R. [B.S. Abdur Rahman University, Chennai 600048 (India); Gowrishankar, N. [IP Rings Ltd., Maraimalainagar, Chennai 603209 (India); Ramasubbu, V. [MTD, MMG, IGCAR, Kalpakkam 603102 (India); Murugesan, S.; Dasgupta, Arup [PMG, MMG, IGCAR, Kalpakkam 603102 (India); Jayakumar, T. [MTD, MMG, IGCAR, Kalpakkam 603102 (India)

2015-12-15

Highlights: • Surface modified RAFMS samples were subjected to HDA and thermal oxidation. • Sample modified by SB process showed better coating and interface morphology. • Aluminized samples at 740 °C for 2 min showed Fe{sub 2}Al{sub 9}Si{sub 2} intermetallic phase. • Oxidized samples showed Fe{sub 2}Al{sub 8}Si, Fe{sub 2}Al{sub 3}Si{sub 3} and Fe{sub 3}Al{sub 2}Si{sub 3} intermetallic phases. • A uniform permeation barrier Al{sub 2}O{sub 3} was formed on the coating of oxidized HDA samples. - Abstract: The most optimal candidate material for fabrication of Test Blanket Module (TBM) in the installation of ITER and future fusion reactors is Reduced Activation Ferritic Martensitic (RAFM) steel, yet one of the major challenges that need to be addressed with RAFM is minimizing the loss of tritium in a reactor environment through the formation of tritium permeation barrier. One of the most promising methods for the tritium permeation barrier is through duplex coating with Al{sub 2}O{sub 3}/Fe–Al which is well known to reduce tritium permeation rate by several orders of magnitude. The present work aims to form an alumina layer on RAFM steel by a two-step method, which consists of (i) Hot Dip Aluminizing (HDA) and (ii) conversion of Al into alumina by a subsequent oxidation process. In addition, the influence of surface roughness of the substrate, superheat condition of the Al alloy melt and its composition on microstructural properties of coating before and after oxidation were investigated using OM, SEM–EDS, XRD, indentation micro hardness and scratch test. The experimental results confirmed the formation of alumina layer on RAFM steel after the HDA and oxidation process. Moreover, the surface roughness of the substrate, melt superheat of Al alloy and its composition are found to have a significant influence on the microstructure, thickness, micro-hardness, nature of intermetallic compounds formed and adhesion strength of the coating.

Post irradiation examination on test fuel pins for PWR

International Nuclear Information System (INIS)

Fogaca Filho, N.; Ambrozio Filho, F.

1981-01-01

Certain aspects of irradiation technology on test fuel pins for PWR, are studied. The results of post irradiation tests, performed on test fuel pins in hot cells, are presented. The results of the tests permit an evaluation of the effects of irradiation on the fuel and cladding of the pin. (Author) [pt

The design of in-pile test section for fuel test loop

Energy Technology Data Exchange (ETDEWEB)

Park, K. N.; Lee, J. M.; Shim, B. S.; Zee, D. Y.; Park, S. H.; Ahn, S. H.; Lee, J. Y.; Kim, Y. J. [KAERI, Taejon (Korea, Republic of)

2004-07-01

As an equipment for nuclear fuel's general performance irradiation test in HANARO, Fuel Test Loop(FTL) has been developed that can irradiate the pin to the maximum number of 3 at the core irradiation hole(IR1 hole) by considering for it's utility and user's irradiation requirement. 3-Pin FTL consists of In-Pile Test Section (IPS) and Out-of-Pile System (OPS). IPS consists for IPS Vessel assembly, In-Pool Piping, IPS Support, In-Pool Piping Support etc. Design that such IPS considers interference item consisted to do not bear in existing facilities by one. IVA that is connected to the OPS are controlled and regulated by means of system pressure, system temperature and the water quality. IPS Vessel assembly is consisted of outer pressure vessel, inner pressure vessel, IPS head, inner assembly and test fuel carrier. After 3-Pin FTL development which is expected to be finished by the 2006, FTL will be used for the irradiation test of the new PWR-type fuel and can maximize the usage of HANARO.

Interim results from UO2 fuel oxidation tests in air

International Nuclear Information System (INIS)

Campbell, T.K.; Gilbert, E.R.; Thornhill, C.K.; White, G.D.; Piepel, G.F.; Griffin, C.W.j.

1987-08-01

An experimental program is being conducted at Pacific Northwest Laboratory (PNL) to extend the characterization of spent fuel oxidation in air. To characterize oxidation behavior of irradiated UO 2 , fuel oxidation tests were performed on declad light-water reactor spent fuel and nonirradited UO 2 pellets in the temperature range of 135 to 250 0 C. These tests were designed to determine the important independent variables that might affect spent fuel oxidation behavior. The data from this program, when combined with the test results from other programs, will be used to develop recommended spent fuel dry-storage temperature limits in air. This report describes interim test results. The initial PNL investigations of nonirradiated and spent fuels identified the important testing variables as temperature, fuel burnup, radiolysis of the air, fuel microstructure, and moisture in the air. Based on these initial results, a more extensive statistically designed test matrix was developed to study the effects of temperature, burnup, and moisture on the oxidation behavior of spent fuel. Oxidation tests were initiated using both boiling-water reactor and pressurized-water reactor fuels from several different reactors with burnups from 8 to 34 GWd/MTU. A 10 5 R/h gamma field was applied to the test ovens to simulate dry storage cask conditions. Nonirradiated fuel was included as a control. This report describes experimental results from the initial tests on both the spent and nonirradiated fuels and results to date on the tests in a 10 5 R/h gamma field. 33 refs., 51 figs., 6 tabs

Interation between a superheated uranium dioxide jet and cold concrete

International Nuclear Information System (INIS)

Howe, L.D.; Denham, M.K.; Turland, B.D.; Dop, L.M.G.; Humphreys, R.J.

1992-01-01

A scoping experiment has been carried out at the Winfrith Technology Centre using its Molten Fuel Test Facilities to examine the initial interaction between a fuel melt and concrete. A molten fuel simulant consisting of 81% UO 2 and 19% Mo with a large superheat (T≅3600 K) was poured onto a basaltic concrete target. Thermocouple data indicate that there was an initial high rate of ablation. The test demonstrated that in the case of such high superheats, a vigorous interaction between the jet and the target takes place, with much of the impinging material ejected within the first few seconds. There was a depression eroded into the target by the jet. The experiment has subsequently been modeled at Culham Laboratory using a version of the CORCON MCCI (molten core-concrete interaction) computer code. The calculations were able to produce a representation of this effect. The results of the experiment and the calculation have been compared with jetting correlations, and reasonable agreement has been found. We conclude by advising caution when applying the results of this isolated test to more prototypic interactions. (orig.)

Conventional fuel tank blunt impact tests : test and analysis results

Science.gov (United States)

2014-04-02

The Federal Railroad Administrations Office of Research : and Development is conducting research into fuel tank : crashworthiness. A series of impact tests are planned to : measure fuel tank deformation under two types of dynamic : loading conditi...

Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

2009-06-15

Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

Test reports for K Basins vertical fuel handling tools

Energy Technology Data Exchange (ETDEWEB)

Meling, T.A.

1995-02-01

The vertical fuel handling tools, for moving N Reactor fuel elements, were tested in the 305 Building Cold Test Facility (CTF) in the 300 Area. After fabrication was complete, the tools were functionally tested in the CTF using simulated N Reactor fuel rods (inner and outer elements). The tools were successful in picking up the simulated N Reactor fuel rods. These tools were also load tested using a 62 pound dummy to test the structural integrity of each assembly. The tools passed each of these tests, based on the performance objectives. Finally, the tools were subjected to an operations acceptance test where K Basins Operations personnel operated the tool to determine its durability and usefulness. Operations personnel were satisfied with the tools. Identified open items included the absence of a float during testing, and documentation required prior to actual use of the tools in the 100 K fuel storage basin.

BWR 9 X 9 Fuel Assembly Thermal-Hydraulic Tests (2): Hydraulic Vibration Test

International Nuclear Information System (INIS)

Yoshiaki Tsukuda; Katsuichiro Kamimura; Toshiitsu Hattori; Akira Tanabe; Noboru Saito; Masahiko Warashina; Yuji Nishino

2002-01-01

Nuclear Power Engineering Corporation (NUPEC) conducted thermal-hydraulic projects for verification of thermal-hydraulic design reliability for BWR high-burnup 8 x 8 and 9 x 9 fuel assemblies, entrusted by the Ministry of Economy, Trade and Industry (METI). As a part of the NUPEC thermal-hydraulic projects, hydraulic vibration tests using full-scale test assemblies simulating 9 x 9 fuel assemblies were carried out to evaluate BWR fuel integrity. The test data were applied to development of a new correlation for the estimation of fuel rod vibration amplitude. (authors)

Irradiation testing of coated particle fuel at Hanaro

International Nuclear Information System (INIS)

Goo Kim, Bong; Sung Cho, Moo; Kim, Yong Wan

2014-01-01

TRISO-coated particle fuel is developing to support development of VHTR in Korea. From August 2013, the first irradiation testing of coated particle fuel was begun to demonstrate and qualify TRISO fuel for use in VHTR in the HANARO at KAERI. This experiment is currently undergoing under the atmosphere of a mixed inert gas without on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The irradiation device contains two test rods, one contains nine fuel compacts and the other five compacts and eight graphite specimens. Each compact has 263 coated particles. After a peak burn-up of about 4 at% and a peak fast neutron fluence of about 1.7 x 10 21 n/cm 2 , PIE will be carried out at KAERI's Irradiated Material Examination Facility. This paper is described characteristics of coated particle fuel, the design of test rod and irradiation device for coated particle fuel, discusses the technical results for irradiation testing at HANARO. (authors)

Premature failure of dissimilar metal weld joint at intermediate temperature superheater tube

Directory of Open Access Journals (Sweden)

Mohammed Al Hajri

2015-04-01

Full Text Available Dissimilar metal weld (DMW joint between alloyed steel (AS and stainless steel (SS failed at one of intermediate temperature superheater (ITSH tube in steam/power generation plant boiler. The premature failure was detected after a relatively short time of operation (8 years where the crack propagated circumferentially from AS side through the ITSH tube. Apart from physical examination, microstructural studies based on optical microscopy, SEM and EDX analysis were performed. The results of the investigation point out the limitation of Carbides precipitation at the alloyed steel/welding interface. This is synonym of creep stage I involvement in the failure of ITSH. Improper post-welding operation and bending moment are considered as root causes of the premature failure.

Review of WWER fuel and material tests in the Halden reactor

International Nuclear Information System (INIS)

Volkov, B.; Kolstad, E.

2006-01-01

A review of the tests with WWER fuels and materials conducted in HBWR over the years of cooperation with Russia is presented. The first test with old generation WWER-440 fuel and PWR specification fuel was carried out from 1995 to 1998. Some differences between these fuels regarding irradiation induced densification and pellet design as well as similar fuel thermal behaviour, swelling and FGR were revealed during the test. The data from this test are reviewed and compared with PIE recently performed to confirm the in-pile measurements. The second test was started in March 1999 with the main objective to study different modified WWER fuels also in comparison with PWR fuel. The results indicated that all these modified WWER fuels exhibit improved densification properties relative to earlier tested fuel. In-pile data on fuel densification have been analysed with respect to as fabricated fuel microstructure and can be used for verification of fuel behaviour models. Corrosion and creep tests in the Halden reactor encompass WWER cladding alloys and some results are given. Prospective WWER fuel and material tests foreseen within the frame of the joint program of OECD HRP are also presented. (authors)

PIE Report on the KOMO-3 Irradiation Test Fuels

International Nuclear Information System (INIS)

Park, Jong Man; Ryu, H. J.; Yang, J. H.

2009-04-01

In the KOMO-3, in-reactor irradiation test had been performed for 12 kinds of dispersed U-Mo fuel rods, a multi wire fuel rod and a tube fuel rod. In this report we described the PIE results on the KOMO-3 irradiation test fuels. The interaction layer thickness between fuel particle and matrix could be reduced by using a large size U-Mo fuel particle or introducing Al-Si matrix or adding the third element in the U-Mo particle. Monolithic fuel rod of multi-wire or tube fuel was also effective in reducing the interaction layer thickness

Test plan for Series 3 NNWSI spent fuel leaching/dissolution tests

International Nuclear Information System (INIS)

Wilson, C.N.

1986-04-01

The Series 3 tests will differ from the Series 2 tests in that the Series 3 tests will be run at 85 0 C (J-13 water) in sealed 304 stainless steel (SS) test vessels. The current NNWSI reference spent fuel container material is 304L SS. The candidate NNWSI repository horizon is above the water table, and 95 0 C (boiling temperature at the repository elevation) is the maximum liquid water temperature expected to contact spent fuel in the repository

14 CFR 23.965 - Fuel tank tests.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel tank tests. 23.965 Section 23.965 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: NORMAL, UTILITY, ACROBATIC, AND COMMUTER CATEGORY AIRPLANES Powerplant Fuel System § 23.965 Fuel...

Situation of test and research reactors' spent fuels

International Nuclear Information System (INIS)

Shimizu, Kenichi; Uchiyama, Junzo; Sato, Hiroshi

1996-01-01

The U.S. DOE decided a renewal Off-Site Fuel Policy for stopping to spread a highly enriched uranium which was originally enriched at the U.S., the policy declared that to receive all HEU spent fuels from Test and Research reactors in all the world. In Japan, under bilateral agreement of cooperation between the government of the United States and the government of Japan concerning peaceful uses of nuclear energy, the highly enriched uranium of Test and Research Reactors' fuels was purchased from the U.S. and the fuels had been manufactured in Japan, America, Germany and France. On the other hand, a former president of the U.S. J. Carter proposed that to convert the fuels from HEU to LEU concerning a nonproliferation of nuclear materials in 1978, and Japan absolutely supported this policy. Under this condition, the U.S. stopped to receive the spent fuels from the other countries concerning legal action to the Off-Site Fuels Policy. As a result, the spent fuels are increasing, and to cross to each reactor's storage capacity, and if this policy start, a faced crisis of Test and Research Reactors will be avoided. (author)

Current and prospective fuel test programmes in the MIR reactor

Energy Technology Data Exchange (ETDEWEB)

Izhutov, A.L.; Burukin, A.V.; Iljenko, S.A.; Ovchinnikov, V.A.; Shulimov, V.N.; Smirnov, V.P. [State Scientific Centre of Russia Research Institute of Atomic Reactors, Ulyanovsk region (Russian Federation)

2007-07-01

MIR reactor is a heterogeneous thermal reactor with a moderator and a reflector made of metal beryllium, it has a channel-type design and is placed in a water pool. MIR reactor is mainly designed for testing fragments of fuel elements and fuel assemblies (FA) of different nuclear power reactor types under normal (stationary and transient) operating conditions as well as emergency situations. At present six test loop facilities are being operated (2 PWR loops, 2 BWR loops and 2 steam coolant loops). The majority of current fuel tests is conducted for improving and upgrading the Russian PWR fuel, these tests involve issues such as: -) long term tests of short-size rods with different modifications of cladding materials and fuel pellets; -) further irradiation of power plant re-fabricated and full-size fuel rods up to achieving 80 MW*d/kg U; -) experiments with leaking fuel rods at different burnups and under transient conditions; -) continuation of the RAMP type experiments at high burnup of fuel; and -) in-pile tests with simulation of LOCA and RIA type accidents. Testing of the LEU (low enrichment uranium) research reactor fuel is conducted within the framework of the RERTR programme. Upgrading of the gas cooled and steam cooled loop facilities is scheduled for testing the HTGR fuel and sub-critical water-cooled reactor, correspondingly. The present paper describes the major programs of the WWER high burn-up fuel behavior study in the MIR reactor, capabilities of the applied techniques and some results of the performed irradiation tests. (authors)

Behavior of metallic fuel in treat transient overpower tests

International Nuclear Information System (INIS)

Bauer, T.H.; Wright, A.E.; Robinson, W.R.; Klickman, A.E.

1988-01-01

Results and analyses are reported for TREAT in-pile transient overpower tests of margin to cladding failure and pre-failure axial expansion of metallic fuel. In all cases the power rise was exponential on an 8 s period until either incipient or actual cladding failure was achieved. Test fuel included EBR-II driver fuel and ternary alloy, the reference fuel of the Intergral Fast Reactor concept. Test pin burnup spanned the widest range available. The nature of the observed cladding failure and resultant fuel dispersals is described. Simple models are presented which describe observed cladding failures and pre-failure axial expansions yet are general enough to apply to all metal fuel types

Remote helium leak test of the DUPIC fuel rod

International Nuclear Information System (INIS)

Kim, W. K; Kim, S. S.; Lim, S. P.; Lee, J. W.; Yang, M. S.

1998-01-01

DUPIC(Direct Use of spent PWR fuel In CANDU reactor) is one of dry reprocessing fuel cycles to reuse irradiated PWR fuel in CANDU power plant. DUPIC fuel is so radioactive that DUPIC fuel is remotely fabricated at hot cell such as IMEF hot cell in which radiation is shielded and remote operation is possible. In this study, Helium leakage has been tested for the simulated DUPIC fuel rod manufactured by Nd:YAG laser end-cap welding at simulated hot cell. The remote inspection technique has been developed to evaluate the soundness of DUPIC fuel fabricated through new processes. Vacuum chamber has been developed to be remotely operated by manipulators at hot cell. As the result of remote test, Helium leakage of DUPIC fuel rod is around background level, CANDU specification has been satisfied. In the result of the study, remote test has been successfully performed at the simulated hot cell, and the soundness of DUPIC fuel rod welded by Nd:YAG laser has been confirmed

HFR irradiation testing of light water reactor (LWR) fuel

International Nuclear Information System (INIS)

Markgraf, J.F.W.

1985-01-01

For the materials testing reactor HFR some characteristic information with emphasis on LWR fuel rod testing capabilities and hot cell investigation is presented. Additionally a summary of LWR fuel irradiation programmes performed and forthcoming programmes are described. Project management information and a list of publications pertaining to LWR fuel rod test programmes is given

Fluid flow test for KMRR fuel assemblies

International Nuclear Information System (INIS)

Chung, Moon Ki; Yang, Sun Kyu; Chung, Chang Hwan; Chun, See Young; Song, Chul Hha; Jun, Hyung Gil; Chung, Heung Joon; Won, Soon Yeun; Cho, Young Rho; Kim, Bok Deuk

1991-01-01

Hydraulic and velocity measurment tests were carried out for the KMRR fuel assembly. Two types of the KMRR fuel assembly are consist of longitudinally finned rods. Experimental data of the pressure drops and friction factors for the KMRR fuel assemlby were produced. The measurement technique for the turbulent flow structure in subchannels using the LDV was obtained. The measurement of the experimental constant of the thermal hydraulic analysis code was investigated. The results in this study are used as the basic data for the development of an analysis code. The measurement technique acquired in this study can be applied to the KMRR thermal hydraulic commissioning test and development of the domestic KMRR fuel fabrication. (Author)

Spent fuel drying system test results (first dry-run)

International Nuclear Information System (INIS)

Klinger, G.S.; Oliver, B.M.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1998-07-01

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site. Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 7.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the first dry-run test, which was conducted without a fuel element. The empty test apparatus was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The data from this dry-run test can serve as a baseline for the first two fuel element tests, 1990 (Run 1) and 3128W (Run 2). The purpose of this dry-run was to establish the background levels of hydrogen in the system, and the hydrogen generation and release characteristics attributable to the test system without a fuel element present. This test also serves to establish the background levels of water in the system and the water release characteristics. The system used for the drying test series was the Whole Element Furnace Testing System, described in Section 2.0, which is located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodology are given in section 3.0, and the experimental

Acoustic emission analysis on tensile failure of steam-side oxide scales formed on T22 alloy superheater tubes

Energy Technology Data Exchange (ETDEWEB)

Huang, Jun-Lin; Zhou, Ke-Yi, E-mail: boiler@seu.edu.cn; Xu, Jian-Qun [Key Laboratory of Energy Thermal Conversion and Control of Ministry of Education, School of Energy and Environment, Southeast University, Nanjing 210096, Jiangsu Province (China); Wang, Xin-Meng; Tu, Yi-You [School of Materials Science and Engineering, Southeast University, Nanjing 210096, Jiangsu Province (China)

2014-07-28

Failure of steam-side oxide scales on boiler tubes can seriously influence the safety of coal-fired power plants. Uniaxial tensile tests employing acoustic emission (AE) monitoring were performed, in this work, to investigate the failure behavior of steam-side oxide scales on T22 alloy boiler superheater tubes. The characteristic frequency spectra of the captured AE signals were obtained by performing fast Fourier transform. Three distinct peak frequency bands, 100-170, 175-250, and 280-390 kHz, encountered in different testing stages were identified in the frequency spectra, which were confirmed to, respectively, correspond to substrate plastic deformation, oxide vertical cracking, and oxide spalling with the aid of scanning electronic microscopy observations, and can thus be used for distinguishing different oxide failure mechanisms. Finally, the critical cracking strain of the oxide scale and the interfacial shear strength of the oxide/substrate interface were estimated, which are the critical parameters urgently desired for modeling the failure behavior of steam-side oxide scales on boiler tubes of coal-fired power plants.

Acoustic emission analysis on tensile failure of steam-side oxide scales formed on T22 alloy superheater tubes

Science.gov (United States)

Huang, Jun-Lin; Zhou, Ke-Yi; Wang, Xin-Meng; Tu, Yi-You; Xu, Jian-Qun

2014-07-01

Failure of steam-side oxide scales on boiler tubes can seriously influence the safety of coal-fired power plants. Uniaxial tensile tests employing acoustic emission (AE) monitoring were performed, in this work, to investigate the failure behavior of steam-side oxide scales on T22 alloy boiler superheater tubes. The characteristic frequency spectra of the captured AE signals were obtained by performing fast Fourier transform. Three distinct peak frequency bands, 100-170, 175-250, and 280-390 kHz, encountered in different testing stages were identified in the frequency spectra, which were confirmed to, respectively, correspond to substrate plastic deformation, oxide vertical cracking, and oxide spalling with the aid of scanning electronic microscopy observations, and can thus be used for distinguishing different oxide failure mechanisms. Finally, the critical cracking strain of the oxide scale and the interfacial shear strength of the oxide/substrate interface were estimated, which are the critical parameters urgently desired for modeling the failure behavior of steam-side oxide scales on boiler tubes of coal-fired power plants.

Acoustic emission analysis on tensile failure of steam-side oxide scales formed on T22 alloy superheater tubes

International Nuclear Information System (INIS)

Huang, Jun-Lin; Zhou, Ke-Yi; Xu, Jian-Qun; Wang, Xin-Meng; Tu, Yi-You

2014-01-01

Failure of steam-side oxide scales on boiler tubes can seriously influence the safety of coal-fired power plants. Uniaxial tensile tests employing acoustic emission (AE) monitoring were performed, in this work, to investigate the failure behavior of steam-side oxide scales on T22 alloy boiler superheater tubes. The characteristic frequency spectra of the captured AE signals were obtained by performing fast Fourier transform. Three distinct peak frequency bands, 100-170, 175-250, and 280-390 kHz, encountered in different testing stages were identified in the frequency spectra, which were confirmed to, respectively, correspond to substrate plastic deformation, oxide vertical cracking, and oxide spalling with the aid of scanning electronic microscopy observations, and can thus be used for distinguishing different oxide failure mechanisms. Finally, the critical cracking strain of the oxide scale and the interfacial shear strength of the oxide/substrate interface were estimated, which are the critical parameters urgently desired for modeling the failure behavior of steam-side oxide scales on boiler tubes of coal-fired power plants.

SP-100 Fuel Pin Performance: Results from Irradiation Testing

Science.gov (United States)

Makenas, Bruce J.; Paxton, Dean M.; Vaidyanathan, Swaminathan; Marietta, Martin; Hoth, Carl W.

1994-07-01

A total of 86 experimental fuel pins with various fuel, liner, and cladding candidate materials have been irradiated in the Experimental Breeder Reactor-II (EBR-II) and the Fast Flux Test Facility (FFTF) reactor as part of the SP-100 fuel pin irradiation testing program. Postirradiation examination results from these fuel pins are key in establishing performance correlations and demonstrating the lifetime and safety of the reactor fuel system. This paper provides a brief description of the in-reactor fuel pin tests and presents the most recent irradiation data on the performance of wrought rhenium (Re) liner material and high density UN fuel at goal burnup of 6 atom percent (at. %). It also provides an overview of the significant variety of other fuel/liner/cladding combinations which were irradiated as part of this program and which may be of interest to more advanced efforts.

Behavior of fission products released from severely damaged fuel during the PBF severe fuel damage tests

International Nuclear Information System (INIS)

Osetek, D.J.; Cronenberg, A.W.; Hagrman, D.L.; Broughton, J.M.; Rest, J.

1984-01-01

The results of fission product release behavior during the first two Power Burst Facility Severe Fuel Damage tests are presented. Measured fission product release is compared with calculated release using temperature dependent release rate correlations and FASTGRASS analysis. The test results indicate that release from fuel of the high volatility fission products (Xe, Kr, I, Cs, and Te) is strongly influenced by parameters other than fuel temperature; namely fuel/fission product morphology, fuel and cladding oxidation state, extent of fuel liquefaction, and quench induced fuel shattering. Fission product transport from the test fuel through the sample system was strongly influenced by chemical effects. Holdup of I and Cs was affected by fission product chemistry, and transport time while Te release was primarily influenced by the extent of zircaloy oxidation. Analysis demonstrates that such integral test data can be used to confirm physical, chemical, and mechanistic models of fission product behavior for severe accident conditions

Power ramp tests of BWR-MOX fuels

International Nuclear Information System (INIS)

Asahi, K.; Oguma, M.; Higuchi, S.; Kamimua, K.; Shirai, Y.; Bodart, S.; Mertens, L.

1996-01-01

Power ramp test of BWR-MOX and UO 2 fuel rods base irradiated up to about 60 GWd/t in Dodewaard reactor have been conducted in BR2 reactor in the framework of the international DOMO programme. The MOX pellets were provided by BN (MIMAS process) and PNC (MH method). The MOX fuel rods with Zr-liner and non-liner cladding and the UO 2 fuel rods with Zr-liner cladding remained intact during the stepwise power ramp tests to about 600 W/cm, even at about 60 GWd/t

Fuel Accident Condition Simulator (FACS) Furnace for Post-Irradiation Heating Tests of VHTR Fuel Compacts

Energy Technology Data Exchange (ETDEWEB)

Paul A Demkowicz; Paul Demkowicz; David V Laug

2010-10-01

Abstract –Fuel irradiation testing and post-irradiation examination are currently in progress as part of the Next Generation Nuclear Plant Fuels Development and Qualification Program. The PIE campaign will include extensive accident testing of irradiated very high temperature reactor fuel compacts to verify fission product retention characteristics at high temperatures. This work will be carried out at both the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory, beginning with accident tests on irradiated fuel from the AGR-1 experiment in 2010. A new furnace system has been designed, built, and tested at INL to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000°C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, Eu, and I) and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator (FACS) furnace system, as well as preliminary system calibration results.

FCTESTNET - Testing fuel cells for transportation

NARCIS (Netherlands)

Winkel, R.G.; Foster, D.L.; Smokers, R.T.M.

2006-01-01

FCTESTNET (Fuel Cell Testing and Standardization Network) is an ongoing European network project within Framework Program 5. It is a three-year project that commenced January 2003, with 55 partners from European research centers, universities, and industry, working in the field of fuel cell R and D.

Power Burst Facility Severe Fuel Damage test series

International Nuclear Information System (INIS)

Buescher, B.J.; Osetek, D.J.; Ploger, S.A.

1982-01-01

The Severe Fuel Damage (SFD) tests planned for the Power Burst Facility (PBF) are described. Bundles containing 32 zircaloy-clad, PWR-type fuel rods will be subjected to severe overheating transients in a high-pressure, superheated-steam environment. Cladding temperatures are expected to reach 2400 0 K, resulting in cladding ballooning and rupture, severe cladding oxidation, cladding melting, fuel dissolution, fuel rod fragmentation, and possibly, rubble bed formation. An experiment effluent collection system is being installed and the PBF fission product monitoring system is being upgraded to meet the special requirements of the SFD tests. Scoping calculations were performed to evaluate performance of the SFD test design and to establish operational requirements for the PBF loop

Posttest examination results of recent treat tests on metal fuel

International Nuclear Information System (INIS)

Holland, J.W.; Wright, A.E.; Bauer, T.H.; Goldman, A.J.; Klickman, A.E.; Sevy, R.H.

1986-01-01

A series of in-reactor transient tests is underway to study the characteristics of metal-alloy fuel during transient-overpower-without-scam conditions. The initial tests focused on determining the margin to cladding breach and the axial fuel motions that would mitigate the power excursion. The tests were conducted in flowing-sodium loops with uranium - 5% fissium EBR-II Mark-II driver fuel elements in the TREAT facility. Posttest examination of the tests evaluated fuel elongation in intact pins and postfailure fuel motion. Microscopic examination of the intact pins studied the nature and extent of fuel/cladding interaction, fuel melt fraction and mass distribution, and distribution of porosity. Eutectic penetration and failure of the cladding were also examined in the failed pins

Nuclear fuels for material test reactors

International Nuclear Information System (INIS)

Ramanathan, L.V.; Durazzo, M.; Freitas, C.T. de

1982-01-01

Experimental results related do the development of nuclear fuels for reactors cooled and moderated by water have been presented cylindrical and plate type fuels have been described in which the core consists of U compouns dispersed in an Al matrix and is clad with aluminium. Fabrication details involving rollmilling, swaging or hot pressing have been described. Corrosion and irradiation test results are also discussed. The performance of the different types of fuels indicates that it is possible to locally fabricate fuel plates with U 3 O 8 +Al cores (20% enriched U) for use in operating Brazilian research reactors. (Author) [pt

Operational reliability testing of FBR fuel in EBR-II

International Nuclear Information System (INIS)

Asaga, Takeo; Ukai, Shigeharu; Nomura, Shigeo; Shikakura, Sakae

1991-01-01

The operational reliability testing of FBR fuel has been conducting in EBR-II as a DOE/PNC collaboration program. This paper reviews the achieved summary of Phase-I test as well as outline of progressing Phase-II test. In Phase-I test, the reliability of FBR fuel pins including 'MONJU' fuel was demonstrated at the event of operational transient. Continued operation of the failed pins was also shown to be feasible without affecting the plant operation. The objectives of the Phase-II test is to extend the data base relating with the operational reliability for long life fuel, and to supply the highly quantitative evaluation. The valuable insight obtained in Phase-II test are considerably expected to be useful toward the achievement of commercial FBR. (author)

Stand for visual ultrasonic testing of spent fuel

International Nuclear Information System (INIS)

Czajkowski, W.; Borek-Kruszewska, E.

2001-01-01

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

Welding of metallic fuel elements for the irradiation test in JOYO. Preliminary tests and welding execution tests (Joint research)

International Nuclear Information System (INIS)

Kikuchi, Hironobu; Nakamura, Kinya; Iwai, Takashi; Arai, Yasuo

2009-10-01

Irradiation tests of metallic fuels elements in fast test reactor JOYO are planned under the joint research of Japan Atomic Energy Agency (JAEA) and Central Research Institute of Electric Power Industry (CRIEPI). Six U-Pu-Zr fuel elements clad with ferritic martensitic steel are fabricated in Plutonium Fuel Research Facility (PFRF) of JAEA-Oarai for the first time in Japan. In PFRF, the procedures of fabrication of the fuel elements were determined and the test runs of the equipments were carried out before the welding execution tests for the fuel elements. Test samples for confirming the welding condition between the cladding tube and top and bottom endplugs were prepared, and various test runs were carried out before the welding execution tests. As a result, the welding conditions were finalized by passing the welding execution tests. (author)

Re-evaluation of superheat conditions postulated in NRC Information Notice 84-90

International Nuclear Information System (INIS)

Alsammarae, A.; Kruger, D.; Beutel, D.; Spisak, M.

1994-01-01

Information Notice 84-90, ''Main Steam Line Break Effect on Environmental Qualification of Equipment,'' describes a potential problem regarding existing plant analyses and Equipment Qualification (EQ) related to a postulated Main Steam Line Break (MSLB) with releases of superheated stream. This notice states that certain methodologies for computing mass and energy releases for a postulated MSLB did not account for heat transfer from the steam generator tube bundles if they were uncovered. Due to this potential change in the original environmental analysis, the EQ of various components may not consider the thermal environment which could result from superheated steam. Subsequent technical assessments may determine that the existing qualification basis for equipment/components does not envelop the postulated superheat condition. Corrective actions need to be taken to demonstrate that the affected equipment is qualified

Evaluation of fuel rods behavior - under irradiation test

International Nuclear Information System (INIS)

Lameiras, F.S.; Terra, J.L.; Pinto, L.C.M.; Dias, M.S.; Pinheiro, R.B.

1981-04-01

By the accompanying of the irradiation of instrumented test fuel rods simulating the operational conditions in reactors, plus the results of post - irradiation exams, tests, evaluation and calibration of analitic modelling of such fuel rods is done. (E.G.) [pt

Severe fuel-damage scoping test performance

International Nuclear Information System (INIS)

Gruen, G.E.; Buescher, B.J.

1983-01-01

As a result of the Three Mile Island Unit-2 (TMI-2) accident, the Nuclear Regulatory Commission has initiated a severe fuel damage test program to evaluate fuel rod and core response during severe accidents similar to TMI-2. The first test of Phase I of this series has been successfully completed in the Power Burst Facility at the Idaho National Engineering Laboratory. Following the first test, calculations were performed using the TRAC-BD1 computer code with actual experimental boundary conditions. This paper discusses the test conduct and performance and presents the calculated and measured test bundle results. The test resulted in a slow heatup to 2000 K over about 4 h, with an accelerated reaction of the zirconium cladding at temperatures above 1600 K in the lower part or the bundle and 2000 K in the upper portion of the bundle

Reusable fuel test assembly for the FFTF

International Nuclear Information System (INIS)

Pitner, A.L.; Dittmer, J.O.

1992-01-01

A fuel test assembly that provides re-irradiation capability after interim discharge and reconstitution of the test pin bundle has been developed for use in the Fast Flux Test Facility (FFTF). This test vehicle permits irradiation test data to be obtained at multiple exposures on a few select test pins without the substantial expense of fabricating individual test assemblies as would otherwise be required. A variety of test pin types can be loaded in the reusable test assembly. A reusable test vehicle for irradiation testing in the FFTF has long been desired, but a number of obstacles previously prevented the implementation of such an experimental rig. The MFF-8A test assembly employs a 169-pin bundle using HT-9 alloy for duct and cladding material. The standard driver pins in the fuel bundle are sodium-bonded metal fuel (U-10 wt% Zr). Thirty-seven positions in the bundle are replaceable pin positions. Standard MFF-8A driver pins can be loaded in any test pin location to fill the bundle if necessary. Application of the MFF-8A reusable test assembly in the FFTF constitutes a considerable cost-saving measure with regard to irradiation testing. Only a few well-characterized test pins need be fabricated to conduct a test program rather than constructing entire test assemblies

Fiscal 2000 achievement report on the followup research on high-efficiency refuse-fueled power generation technology development; 2000 nendo kokoritsu haikibutsu hatsuden gijutsu kaihatsu no follow up kenkyu seika hokokusho

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-03-01

Followup research was conducted using a high-efficiency refuse-fueled power generation pilot plant for interpolating research and development results, improvement on refuse-fueled power generation efficiency, and for introduction and popularization of refuse-fueled power generation. Operation control and safety control were performed in compliance with the pilot plant operation program, when high-temperature high-pressure steam (500 degrees C, 100 atm) was stably generated for approximately 7,600 hours in this fiscal year recording a total of approximately 22,000 hours when counted from the beginning of the verification test. Plant operation data were examined and every part of the furnace was checked and measured while the furnace was at rest, and it was found that the plant satisfied the intended specifications as refuse-fueled power plant facilities. As for its environmental impact, it was confirmed that it fully satisfied the standard values for substances in exhaust gas, which involved soot and dust, sulfur oxides, nitrogen oxides, hydrogen chloride, and dioxins. Dioxins, in particular, were far below the most rigorous legal level of 0.1ng-TEQ/m{sup 3}N. In the test of superheater materials, long-term data covering three years were acquired. (NEDO)

Spent fuel handling system for a geologic storage test at the Nevada Test Site

International Nuclear Information System (INIS)

Duncan, J.E.; House, P.A.; Wright, G.W.

1980-01-01

The Lawrence Livermore Laboratory is conducting a test of the geologic storage of encapsulated spent commercial reactor fuel assemblies in a granitic rock at the Nevada Test Site. The test, known as the Spent Fuel Test-Climax (SFT-C), is sponsored by the US Department of Energy, Nevada Operations Office. Eleven pressurized-water-reactor spent fuel assemblies are stored retrievably for three to five years in a linear array in the Climax stock at a depth of 420 m

Test of fuel handling machine for Monju in sodium

International Nuclear Information System (INIS)

Ishii, Yoichiro; Masuda, Yoichi; Kataoka, Hajime

1980-01-01

Various types of fuel handling machines were studied, and under-the-plug method of fuel exchange and the fuel handling machine of single turning plug, fixed arm type were selected for the prototype reactor ''Monju'', because the turning plug is relatively small, and the rate of operation, safety, operational ability, maintainability and reliability required for the reactor are satisfied, moreover, the extrapolation to the demonstration reactor was considered. Attention must be paid to the points that the fuel handling machine is very long and invisible from outside, and the smooth operation and endurance in sodium are required for it. The full mock-up testing facility of single turning plug, fixed arm type was installed in 1974, and the full mock-up test has been carried out since 1975 in Oarai. Fuel exchange is carried out at about 6 months intervals in Monju, and about 20 to 30% of core and blanket fuels are exchanged for about one month period. The functions required for the fuel handling machine for Monju, the outline of the testing facility, the schedule of the testing, the items of testing and the results, and the matters to be specially written are described. The full mock-up test in sodium has been carried out for 5 years, and the functions and the endurance have been proved sufficiently. (Kako, I.)

Opportunities for mixed oxide fuel testing in the advanced test reactor to support plutonium disposition

International Nuclear Information System (INIS)

Terry, W.K.; Ryskamp, J.M.; Sterbentz, J.W.

1995-08-01

Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification; (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania; (3) The effects of WGPu isotopic composition; (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight; (5) The effects of americium and gallium in WGPu; (6) Fission gas release from MOX fuel pellets made from WGPu; (7) Fuel/cladding gap closure; (8) The effects of power cycling and off-normal events on fuel integrity; (9) Development of radial distributions of burnup and fission products; (10) Power spiking near the interfaces of MOX and urania fuel assemblies; and (11) Fuel performance code validation. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory possesses many advantages for performing tests to resolve most of the issues identified above. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified. The facilities at Argonne National Laboratory-West can meet all potential needs for pre- and post-irradiation examination that might arise in a MOX fuel qualification program

Investigation into Cause of High Temperature Failure of Boiler Superheater Tube

Science.gov (United States)

Ghosh, D.; Ray, S.; Roy, H.; Shukla, A. K.

2015-04-01

The failure of the boiler tubes occur due to various reasons like creep, fatigue, corrosion and erosion. This paper highlights a case study of typical premature failure of a final superheater tube of 210 MW thermal power plant boiler. Visual examination, dimensional measurement, chemical analysis, oxide scale thickness measurement, microstructural examination are conducted as part of the investigations. Apart from these investigations, sulfur print, Energy Dispersive spectroscopy (EDS) and X ray diffraction analysis (XRD) are also conducted to ascertain the probable cause of failure of final super heater tube. Finally it has been concluded that the premature failure of the super heater tube can be attributed to the combination of localized high tube metal temperature and loss of metal from the outer surface due to high temperature corrosion. The corrective actions have also been suggested to avoid this type of failure in near future.

Diesel fuel lubricity testing revisited : Tests von Dieselkraftstoffschmierfähigkeit erneut betrachtet

NARCIS (Netherlands)

van Leeuwen, H.J.

2017-01-01

Fuel is used as a lubricant in several engine components. Diesel fuel is known for its good lubrication properties, better than gasoline. These properties are examined in standard tests, as prescribed by ASTM. Good lubrication properties are designated as a good lubricity. Most commonly, fuel

HRB-22 capsule irradiation test for HTGR fuel. JAERI/USDOE collaborative irradiation test

Energy Technology Data Exchange (ETDEWEB)

Minato, Kazuo; Sawa, Kazuhiro; Fukuda, Kousaku [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; and others

1998-03-01

As a JAERI/USDOE collaborative irradiation test for high-temperature gas-cooled reactor fuel, JAERI fuel compacts were irradiated in the HRB-22 irradiation capsule in the High Flux Isotope Reactor at the Oak Ridge National Laboratory (ORNL). Postirradiation examinations also were performed at ORNL. This report describes 1) the preirradiation characterization of the irradiation samples of annular-shaped fuel compacts containing the Triso-coated fuel particles, 2) the irradiation conditions and fission gas releases during the irradiation to measure the performance of the coated particle fuel, 3) the postirradiation examinations of the disassembled capsule involving visual inspection, metrology, ceramography and gamma-ray spectrometry of the samples, and 4) the accident condition tests on the irradiated fuels at 1600 to 1800degC to obtain information about fuel performance and fission product release behavior under accident conditions. (author)

Feasibility study on the transient fuel test loop installation

International Nuclear Information System (INIS)

Kim, J. Y.; Lee, C. Y.

1997-02-01

The design and installation of the irradiation test facility for verification test of the fuel performance are very important in connection with maximization of the utilization of HANARO. The objective of this study is to investigate and analyze the test capsules and loops in research reactors of the other countries and to design preliminarily the eligible transient fuel test facility to be installed in HANARO. The principle subjects of this study are to analyze the contents, kinds and scopes of the irradiation test facilities for nuclear technology development. The guidances for the basic and detail design of the transient fuel test facility in the future are presented. The investigation and analysis of various kinds of test facilities that are now in operation at the research reactors of nuclear advanced countries are carried out. Based on the design data of HANARO the design materials for an eligible transient fuel test facility comprises two pacts : namely, in pile test fuel in reactor core site, and out of pile system regulates the experimental conditions in the in pile test section. Especially for power ramping and cycling selection of the eligible power variation equipment in HANARO is carried out. (author). 13 refs., 4 tabs., 46 figs

Spent fuel metal storage cask performance testing and future spent fuel concrete module performance testing

International Nuclear Information System (INIS)

McKinnon, M.A.; Creer, J.M.

1988-10-01

REA-2023 Gesellshaft fur Nuklear Service (GNS) CASTOR-V/21, Transnuclear TN-24P, and Westinghouse MC-10 metal storage casks, have been performance tested under the guidance of the Pacific Northwest Laboratory to determine their thermal and shielding performance. The REA-2023 cask was tested under Department of Energy (DOE) sponsorship at General Electric's facilities in Morris, Illinois, using BWR spent fuel from the Cooper Reactor. The other three casks were tested under a cooperative agreement between Virginia Power Company and DOE at the Idaho National Engineering Laboratory (INEL) by EGandG Idaho, Inc., using intact spent PWR fuel from the Surry reactors. The Electric Power Research Institute (EPRI) made contributions to both programs. A summary of the various cask designs and the results of the performance tests is presented. The cask designs include: solid and liquid neutron shields; lead, steel, and nodular cast iron gamma shields; stainless steel, aluminum, and copper baskets; and borated materials for criticality control. 4 refs., 8 figs., 6 tabs

Modular, High-Volume Fuel Cell Leak-Test Suite and Process

Energy Technology Data Exchange (ETDEWEB)

Ru Chen; Ian Kaye

2012-03-12

Fuel cell stacks are typically hand-assembled and tested. As a result the manufacturing process is labor-intensive and time-consuming. The fluid leakage in fuel cell stacks may reduce fuel cell performance, damage fuel cell stack, or even cause fire and become a safety hazard. Leak check is a critical step in the fuel cell stack manufacturing. The fuel cell industry is in need of fuel cell leak-test processes and equipment that is automatic, robust, and high throughput. The equipment should reduce fuel cell manufacturing cost.

Initiation of depleted uranium oxide and spent fuel testing for the spent fuel sabotage aerosol ratio program

Energy Technology Data Exchange (ETDEWEB)

Molecke, M.A.; Gregson, M.W.; Sorenson, K.B. [Sandia National Labs. (United States); Billone, M.C.; Tsai, H. [Argonne National Lab. (United States); Koch, W.; Nolte, O. [Fraunhofer Inst. fuer Toxikologie und Experimentelle Medizin (Germany); Pretzsch, G.; Lange, F. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (Germany); Autrusson, B.; Loiseau, O. [Inst. de Radioprotection et de Surete Nucleaire (France); Thompson, N.S.; Hibbs, R.S. [U.S. Dept. of Energy (United States); Young, F.I.; Mo, T. [U.S. Nuclear Regulatory Commission (United States)

2004-07-01

We provide a detailed overview of an ongoing, multinational test program that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolized materials plus volatilized fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high energy density device, HEDD. The program participants in the U.S. plus Germany, France, and the U.K., part of the international Working Group for Sabotage Concerns of Transport and Storage Casks, WGSTSC have strongly supported and coordinated this research program. Sandia National Laboratories, SNL, has the lead role for conducting this research program; test program support is provided by both the U.S. Department of Energy and Nuclear Regulatory Commission. WGSTSC partners need this research to better understand potential radiological impacts from sabotage of nuclear material shipments and storage casks, and to support subsequent risk assessments, modeling, and preventative measures. We provide a summary of the overall, multi-phase test design and a description of all explosive containment and aerosol collection test components used. We focus on the recently initiated tests on ''surrogate'' spent fuel, unirradiated depleted uranium oxide, and forthcoming actual spent fuel tests. The depleted uranium oxide test rodlets were prepared by the Institut de Radioprotection et de Surete Nucleaire, in France. These surrogate test rodlets closely match the diameter of the test rodlets of actual spent fuel from the H.B. Robinson reactor (high burnup PWR fuel) and the Surry reactor (lower, medium burnup PWR fuel), generated from U.S. reactors. The characterization of the spent fuels and fabrication into short, pressurized rodlets has been performed by Argonne National Laboratory, for testing at SNL. The ratio of the aerosol and respirable particles released from HEDD-impacted spent

Fabrication of Fast Reactor Fuel Pins for Test Irradiations

Energy Technology Data Exchange (ETDEWEB)

Karsten, G. [Institute for Applied Reactor Physics, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany); Dippel, T. [Institute for Radiochemistry, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany); Laue, H. J. [Institute for Applied Reactor Physics, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany)

1967-09-15

An extended irradiation programme is being carried out for the fuel element development of the Karlsruhe fast breeder project. A very important task within the programme is the testing of plutonium-containing fuel pins in a fast-reactor environment. This paper deals with fabrication of such pins by our laboratories at Karlsruhe. For the fast reactor test positions at present envisaged a fuel with 15% plutonium and the uranium fully enriched is appropriate. Hie mixed oxide is both pelletized and vibro-compacted with smeared densities between 80 and 88% theoretical. The pin design is, for example, such that there are two gas plena at the top and bottom, and one blanket above the fuel with the fuel zone fitting to the test reactor core length. The specifications both for fuel and cladding have been adapted to the special purpose of a fast-breeder reactor - the outer dimensions, the choice of cladding and fuel types, the data used and the kind of tests outline the targets of the development. The fuel fabrication is described in detail, and also the powder line used for vibro-compaction. The source materials for the fuel are oxalate PuO{sub 2} and UO{sub 2} from the UF{sub 6} process. The special problems of mechanical mixing and of plutonium homogeneity have been studied. The development of the sintering technique and grain characteristics for vibratory compactive fuel had to overcome serious problems in order to reach 82-83% theoretical. The performance of the pin fabrication needed a major effort in welding, manufacturing of fits and decontamination of the pin surfaces. This was a stimulation for the development of some very subtle control techniques, for example taking clear X-ray photographs and the tube testing. In general the selection of tests was a special task of the production routine. In conclusion the fabrication of the pins resulted in valuable experiences for the further development of fast reactor fuel elements. (author)

Irradiation Testing of TRISO-Coated Particle Fuel in Korea

International Nuclear Information System (INIS)

Kim, Bong Goo; Yeo, Sunghwan; Jeong, Kyung-Chai; Eom, Sung-Ho; Kim, Yeon-Ku; Kim, Woong Ki; Lee, Young Woo; Cho, Moon Sung; Kim, Yong Wan

2014-01-01

In Korea, coated particle fuel is being developed to support development of a VHTR. At the end of March 2014, the first irradiation test in HANARO at KAERI to demonstrate and qualify TRISO-coated particle fuel for use in a VHTR was terminated. This experiment was conducted in an inert gas atmosphere without on-line temperature monitoring and control, or on-line fission product monitoring of the sweep gas. The irradiation device contained two test rods, one has nine fuel compacts and the other five compacts and eight graphite specimens. Each compact contains about 260 TRISO-coated particles. The duration of irradiation testing at HANARO was about 135 full power days from last August 2013. The maximum average power per particle was about 165 mW/particle. The calculated peak burnup of the TRISO-coated fuel was a little less than 4 atom percent. Post-irradiation examination is being carried out at KAERI’s Irradiated Material Examination Facility beginning in September of 2014. This paper describes characteristics of coated particle fuel, the design of the test rod and irradiation device for this coated particle fuel, and discusses the technical results of irradiation testing at HANARO. (author)

Spent fuel sabotage aerosol test program :FY 2005-06 testing and aerosol data summary

International Nuclear Information System (INIS)

Gregson, Michael Warren; Brockmann, John E.; Nolte, O.; Loiseau, O.; Koch, W.; Molecke, Martin Alan; Autrusson, Bruno; Pretzsch, Gunter Guido; Billone, M. C.; Lucero, Daniel A.; Burtseva, T.; Brucher, W; Steyskal, Michele D.

2006-01-01

This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program has been underway for several years. This program provides source-term data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. This document focuses on an updated description of the test program and test components for all work and plans made, or revised, primarily during FY 2005 and about the first two-thirds of FY 2006. It also serves as a program status report as of the end of May 2006. We provide details on the significant findings on aerosol results and observations from the recently completed Phase 2 surrogate material tests using cerium oxide ceramic pellets in test rodlets plus non-radioactive fission product dopants. Results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; status on determination of the spent fuel ratio, SFR (the ratio of respirable particles from real spent fuel/respirables from surrogate spent fuel, measured under closely matched test conditions, in a contained test chamber); and, measurements of enhanced volatile fission product species sorption onto respirable particles. We discuss progress and results for the first three, recently performed Phase 3 tests using depleted uranium oxide, DUO 2 , test rodlets. We will also review the status of preparations and the final Phase 4 tests in this program, using short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. These data plus testing results and design are tailored to support and guide, follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments

Research reactors for power reactor fuel and materials testing - Studsvik's experience

International Nuclear Information System (INIS)

Grounes, M.

1998-01-01

Presently Studsvik's R2 test reactor is used for BWR and PWR fuel irradiations at constant power and under transient power conditions. Furthermore tests are performed with defective LWR fuel rods. Tests are also performed on different types of LWR cladding materials and structural materials including post-irradiation testing of materials irradiated at different temperatures and, in some cases, in different water chemistries and on fusion reactor materials. In the past, tests have also been performed on HTGR fuel and FBR fuel and materials under appropriate coolant, temperature and pressure conditions. Fuel tests under development include extremely fast power ramps simulating some reactivity initiated accidents and stored energy (enthalpy) measurements. Materials tests under development include different types of in-pile tests including tests in the INCA (In-Core Autoclave) facility .The present and future demands on the test reactor fuel in all these cases are discussed. (author)

Spent fuel's behavior under dynamic drip tests

International Nuclear Information System (INIS)

Finn, P.A.; Buck, E.C.; Hoh, J.C.; Bates, J.K.

1995-01-01

In the potential repository at Yucca Mountain, failure of the waste package container and the cladding of the spent nuclear fuel would expose the fuel to water under oxidizing conditions. To simulate the release behavior of radionuclides from spent fuel, dynamic drip and vapor tests with spent nuclear fuel have been ongoing for 2.5 years. Rapid alteration of the spent fuel has been noted with concurrent release of radionuclides. Colloidal species containing americium and plutonium have been found in the leachate. This observation suggests that colloidal transport of radionuclides should be included in the performance assessment of a potential repository

Feasibility study of the Dragon reactor for HTGR fuel testing

International Nuclear Information System (INIS)

Wallroth, C.F.

1975-01-01

The Organization of European Community Development (OECD) Dragon high-temperature reactor project has performed HTGR fuel and fuel element testing for about 10 years. To date, a total of about 250 fuel elements have been irradiated and the test program continues. The feasibility of using this test facility for HTGR fuel testing, giving special consideration to U. S. needs, is evaluated. A detailed description for design, preparation, and data acquisition of a test experiment is given together with all possible options on supporting work, which could be carried out by the experienced Dragon project staff. 11 references. (U.S.)

Behavior of metallic uranium-fissium fuel in TREAT transient overpower tests

International Nuclear Information System (INIS)

Bauer, T.H.; Klickman, A.E.; Lo, R.K.; Rhodes, E.A.; Robinson, W.R.; Stanford, G.S.; Wright, A.E.

1986-01-01

TREAT tests M2, M3, and M4 were performed to obtain information on two key behavior characteristics of fuel under transient overpower accident conditions in metal-fueled fast reactors: the prefailure axial self-extrusion (elongation beyond thermal expansion) of fuel within intact cladding and the margin to cladding breach. Uranium-5 wt% fissium Experimental Breeder Reactor-II driver fuel pins were used for the tests since they were available as suitable stand-ins for the uranium-plutonium-zirconium ternary fuel, which is the reference fuel of the integral fast reactor (IFR) concept. The ternary fuel will be used in subsequent TREAT tests. Preliminary results from tests M2 and M3 were presented earlier. The present report includes significant advances in analysis as well as additional data from test M4. Test results and analysis have led to the development and validation of pin cladding failure and fuel extrusion models for metallic fuel, within reasonable uncertainties for the uranium-fissium alloy. Concepts involved are straightforward and readily extendable to ternary alloys and behavior in full-size reactors

33 CFR 183.580 - Static pressure test for fuel tanks.

Science.gov (United States)

2010-07-01

... pressure test for fuel tanks. A fuel tank is tested by performing the following procedures in the following... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Static pressure test for fuel tanks. 183.580 Section 183.580 Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND...

Power ramp testing method for PWR fuel rod at research reactor

International Nuclear Information System (INIS)

Zhou Yidong; Zhang Peisheng; Zhang Aimin; Gao Yongguang; Wang Huarong

2003-01-01

A tentative power ramp test for short PWR fuel rod has been conducted at the Heavy Water Research Reactor (HWRR) in China Institute of Atomic Energy (CIAE). The test fuel rod was cooled by the circulating water in the test loop. The power ramp was realized by moving solid neutron-absorbing screen around the fuel rod. The linear power of the fuel rod increased from 220 W/cm to 340 W/cm with a power ramp rate of 20 W/cm/min. The power of the fuel rod was monitored by both in-core thermal and nuclear measurement sensors in the test rig. This test provides experiences for further developing the power ramp test methods for PWR fuel rods at research reactor. (author)

Testing plutonium fuel assembly production for fast-neutron reactors

International Nuclear Information System (INIS)

Nougues, B.; Benhamou, A.; Bertothy, G.; Lepetit, H.

1975-01-01

The main characteristics of plutonium fuel elements for fast breeder reactors justify specific test procedures and special techniques. The specific tests relating to the Pu content consist of Pu enrichment and distribution tests, determination of the O/M ratio and external contamination tests. The specific tests performed on fuel configuration are: testing of sintered pellet diameter, testing of pin welding and checking of internal assmbly [fr

Drop testing of the Westinghouse fresh nuclear fuel package

International Nuclear Information System (INIS)

Shappert, L.B.; Sanders, C.F.

1993-01-01

The Westinghouse Columbia Fuel Fabrication Facility has decided to develop and certify a new fresh fuel package design (type A, fissile) that has the capability to transport more highly enriched fuel than was previously possible. A prototype package was tested in support of the Safety Analysis Report of the Packaging (SARP). This paper provides detailed information on the tests and test results. A first prototype test was carried out at the STF, and the design did not give the safety margin that Westinghouse wanted for their containers. The data from the test were used to redesign the connection between the clamping frame and the pressure pad, and the tests were reinitiated. Three packages were then tested at the STF. All packages met the acceptance criteria and acceleration information was obtained that provided an indication of the behavior of the cradle and strongback which holds the fuel assemblies and nuclear poison in place. (J.P.N.)

Spent fuel drying system test results (second dry-run)

International Nuclear Information System (INIS)

Klinger, G.S.; Oliver, B.M.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1998-07-01

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks have been detected in the basins and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 7.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the second dry-run test, which was conducted without a fuel element. With the concurrence of project management, the test protocol for this run, and subsequent drying test runs, was modified. These modifications were made to allow for improved data correlation with drying procedures proposed under the IPS. Details of these modifications are discussed in Section 3.0

Reactivity initiated accident test series Test RIA 1-4 fuel behavior report

International Nuclear Information System (INIS)

Cook, B.A.; Martinson, Z.R.

1984-09-01

This report presents and discusses results from the final test in the Reactivity Initiated Accident (RIA) Test Series, Test RIA 1-4, conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. Nine preirradiated fuel rods in a 3 x 3 bundle configuration were subjected to a power burst while at boiling water reactor hot-startup system conditions. The test resulted in estimated axial peak, radial average fuel enthalpies of 234 cal/g UO 2 on the center rod, 255 cal/g UO 2 on the side rods, and 277 cal/g UO 2 on the corner rods. Test RIA 1-4 was conducted to investigate fuel coolability and channel blockage within a bundle of preirradiated rods near the present enthalpy limit of 280 cal/g UO 2 established by the US Nuclear Regulatory Commission. The test design and conduct are described, and the bundle and individual rod thermal and mechanical responses are evaluated. Conclusions from this final test and the entire PBF RIA Test Series are presented

Results of tests with open fuel in KNK II

International Nuclear Information System (INIS)

Schmitz, G.

1987-03-01

For the operation of Liquid Metal Cooled Fast Breeder Reactors with cladding failures the consequences of increased contamination by fission products and fuel and the possibility of failure propagation to adjacent fuel pins due to fuel swelling have to be envisaged. To clarify some of these problems a KNK II test program involving open fuel was defined with the first experiments of this program being performed between October 1981 and May 1984. After the description of the test equipment and of the test program, the results will be presented on delayed neutron measurements, fission gas measurements and post irradiation examinations. The report will conclude with a discussion of the results [de

Concepts for Small-Scale Testing of Used Nuclear Fuel

Energy Technology Data Exchange (ETDEWEB)

Marschman, Steven Craig [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Philip Lon [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-09-01

This report documents a concept for a small-scale test involving between one and three Boiling Water Rector (BWR) high burnup (HBU) fuel assemblies. This test would be similar to the DOE funded High Burn-Up (HBU) Confirmatory Data Project to confirm the behavior of used high burn-up fuel under prototypic conditions, only on a smaller scale. The test concept proposed would collect data from fuel stored under prototypic dry storage conditions to mimic, as closely as possible, the conditions HBU UNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to an Independent Spent Fuel Storage Installation (ISFSI) for multi-year storage.

Review about corrosion of superheaters tubes in biomass plants; Revision sobre la corrosion de tubos sobrecalentadores en plantas de biomasa

Energy Technology Data Exchange (ETDEWEB)

Berlanga-Labari, C.; Fernandez-Carrasquilla, J.

2006-07-01

The design of new biomass-fired power plants with increased steam temperature raises concerns of high-temperature corrosion. The high potassium and chlorine contents in many biomass, specially in wheat straw, are potentially harmful elements with regard to corrosion. Chlorine may cause accelerated corrosion resulting in increased oxidation, metal wastage, internal attack, void formations and loose non-adherent scales. The most severe corrosion problems in biomass-fired systems are expected to occur due to Cl-rich deposits formed on superheater tubes. In the first part of this revision the corrosion mechanism proposed are described in function of the conditions and compounds involved. The second part is focused on the behaviour of the materials tested so far in the boiler and in the laboratory. First the traditional commercial alloys are studied and secondly the new alloys and the coasting. (Author). 102 refs.

Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

International Nuclear Information System (INIS)

Joung, Chang Young; Ahn, Sung Ho; Heo, Sung Ho; Hong, Jin Tae; Kim, Ka Hye

2013-01-01

It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses

Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

Energy Technology Data Exchange (ETDEWEB)

Joung, Chang Young; Ahn, Sung Ho; Heo, Sung Ho; Hong, Jin Tae; Kim, Ka Hye [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses.

Certification test for safety of new fuel transportation package

International Nuclear Information System (INIS)

Aritomi, Masanori; Sugawa, Osami; Suga, Masao.

1993-01-01

The objective of this certification test is to prove the safety of new fuel transportation package against a fire of actual size caused by traffic accidents. After the fire test, the fuel assemblies were covered with coal-tar like material vaporized from anti-shock material used in the container. Surface color of BWR-type fuel assembly was dark grey that is supposed to be the color of oxide of Zircaloy. As for PWR-type fuel assembly, the condition encountered during fire test caused no change to the outlook of the rod element. Both the BWR and PWR type fuel rod elements showed no deformation and were completely sound. Therefore it may be concluded that the container protected the mimic fuel assemblies against fire of 30 minutes duration and caused no damage. This report is the result of the above experiments and examinations, and we appreciate the cooperation of those who are concerned. (J.P.N.)

Preliminary test results for post irradiation examination on the HTTR fuel

International Nuclear Information System (INIS)

Ueta, Shohei; Umeda, Masayuki; Sawa, Kazuhiro; Sozawa, Shizuo; Shimizu, Michio; Ishigaki, Yoshinobu; Obata, Hiroyuki

2007-01-01

The future post-irradiation program for the first-loading fuel of the HTTR is scheduled using the HTTR fuel handling facilities and the Hot Laboratory in the Japan Materials Testing Reactor (JMTR) to confirm its irradiation resistance and to obtain data on its irradiation characteristics in the core. This report describes the preliminary test results and the future plan for a post-irradiation examination for the HTTR fuel. In the preliminary test, fuel compacts made with the same SiC-coated fuel particle as the first loading fuel were used. In the preliminary test, dimension, weight, fuel failure fraction, and burnup were measured, and X-ray radiograph, SEM, and EPMA observations were carried out. Finally, it was confirmed that the first-loading fuel of the HTTR showed good quality under an irradiation condition. The future plan for the post-irradiation tests was described to confirm its irradiation performance and to obtain data on its irradiation characteristics in the HTTR core. (author)

Iowa Central Quality Fuel Testing Laboratory

Energy Technology Data Exchange (ETDEWEB)

Heach, Don; Bidieman, Julaine

2013-09-30

The objective of this project is to finalize the creation of an independent quality fuel testing laboratory on the campus of Iowa Central Community College in Fort Dodge, Iowa that shall provide the exploding biofuels industry a timely and cost-effective centrally located laboratory to complete all state and federal fuel and related tests that are required. The recipient shall work with various state regulatory agencies, biofuel companies and state and national industry associations to ensure that training and testing needs of their members and American consumers are met. The recipient shall work with the Iowa Department of Ag and Land Stewardship on the development of an Iowa Biofuel Quality Standard along with the Development of a standard that can be used throughout industry.

Four stream breakup of molten IFR [Integral Fast Reactor] metal fuel in sodium

International Nuclear Information System (INIS)

Gabor, J.D.; Purviance, R.T.; Aeschlimann, R.W.; Spencer, B.W.

1988-01-01

Tests have been conducted in which the breakup behavior of kilogram quantities of molten uranium, uranium-zirconium alloy, and uranium-iron alloy pour streams in 600C sodium was studied. A sodium depth of less than 0.3 m was required for hydrodynamic breakup and freezing of 25-mm pour streams of uranium and uranium-zirconium alloy with up to 400C melt superheat. The breakup material was primarily in the form of filaments and sheets with a settled bed voidage on the order of 0.9. The uranium-iron alloy with 800C melt superheat exhibited similar behavior except a sodium depth somewhat greater than 0.3 m was required for breakup and freezing of the particles

14 CFR 25.952 - Fuel system analysis and test.

Science.gov (United States)

2010-01-01

... using the airplane fuel system or a test article that reproduces the operating characteristics of the... AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System § 25.952 Fuel system...

Fuel leak testing performance at NPP Jaslovske Bohunice

International Nuclear Information System (INIS)

Slugen, V.; Krnac, S.; Smiesko, I.

1995-01-01

The NPP Bohunice VVER-440 fuel leak testing experience are relatively extensive in comparison with other VVER-440 users. As the first Europe NPP was adapted Siemens (KWU) in core-sipping equipment to VVER-440 units and since this time were have done these tests also for NPP Paks (Hungary) and NPP Dukovany (Czech Republic). The occurrence of leaking fuel assemblies in NPP is in the last 5 years relatively stabilised and low. A significant difference can be observed between type V-230 (31 leaks) and type V-213 (1 leak). None of of the indicated leaking fuel assemblies has been investigated in the hot cell. Therefore cannot be confirm the effective causes of leak occurrence. Nevertheless, the fuel failure rate and the performance of leak testing in NPP Bohunice are comparable to the world standard at PWR's. 1 tab., 2 figs., 3 refs

Fuel leak testing performance at NPP Jaslovske Bohunice

Energy Technology Data Exchange (ETDEWEB)

Slugen, V; Krnac, S [Slovak Technical Univ., Bratislava (Slovakia); Smiesko, I [Nuclear Powr Plant EBO, Jaslovske Bohuce (Slovakia)

1996-12-31

The NPP Bohunice VVER-440 fuel leak testing experience are relatively extensive in comparison with other VVER-440 users. As the first Europe NPP was adapted Siemens (KWU) in core-sipping equipment to VVER-440 units and since this time were have done these tests also for NPP Paks (Hungary) and NPP Dukovany (Czech Republic). The occurrence of leaking fuel assemblies in NPP is in the last 5 years relatively stabilised and low. A significant difference can be observed between type V-230 (31 leaks) and type V-213 (1 leak). None of of the indicated leaking fuel assemblies has been investigated in the hot cell. Therefore cannot be confirm the effective causes of leak occurrence. Nevertheless, the fuel failure rate and the performance of leak testing in NPP Bohunice are comparable to the world standard at PWR`s. 1 tab., 2 figs., 3 refs.

Irradiation test plan of the simulated DUPIC fuel

Energy Technology Data Exchange (ETDEWEB)

Bae, Ki Kwang; Yang, M. S.; Kim, B. K. [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-11-01

Simulated DUPIC fuel had been irradiated from Aug. 4, 1999 to Oct. 4 1999, in order to produce the data of its in-core behavior, to verify the design of DUPIC non-instrumented capsule developed, and to ensure the irradiation requirements of DUPIC fuel at HANARO. The welding process was certified for manufacturing the mini-element, and simulated DUPIC fuel rods were manufactured with simulated DUPIC pellets through examination and test. The non-instrumented capsule for a irradiation test of DUPIC fuel has been designed and manufactured referring to the design specification of the HANARO fuel. This is to be the design basis of the instrumented capsule under consideration. The verification experiment, whether the capsule loaded in the OR4 hole meet the HANARO requirements under the normal operation condition, as well as the structural analysis was carried out. The items for this experiment were the pressure drop test, vibration test, integrity test, et. al. It was noted that each experimental result meet the HANARO operational requirements. For the safety analysis of the DUPIC non-instrumented capsule loaded in the HANARO core, the nuclear/mechanical compatibility, thermodynamic compatibility, integrity analysis of the irradiation samples according to the reactor condition as well as the safety analysis of the HANARO were performed. Besides, the core reactivity effects were discussed during the irradiation test of the DUPIC capsule. The average power of each fuel rod in the DUPIC capsule was calculated, and maximal linear power reflecting the axial peaking power factor from the MCNP results was evaluated. From these calculation results, the HANARO core safety was evaluated. At the end of this report, similar overseas cases were introduced. 9 refs., 16 figs., 10 tabs. (Author)

Space reactor fuel element testing in upgraded TREAT

International Nuclear Information System (INIS)

Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W.Y.

1993-01-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ∼60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ∼100 MW/L may be achievable

Space reactor fuel element testing in upgraded TREAT

Science.gov (United States)

Todosow, Michael; Bezler, Paul; Ludewig, Hans; Kato, Walter Y.

1993-01-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ˜60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ˜100 MW/L may be achievable.

Tests on CANDU fuel elements sheath samples

International Nuclear Information System (INIS)

Ionescu, S.; Uta, O.; Mincu, M.; Prisecaru, I.

2016-01-01

This work is a study of the behavior of CANDU fuel elements after irradiation. The tests are made on ring samples taken from fuel cladding in INR Pitesti. This paper presents the results of examinations performed in the Post Irradiation Examination Laboratory. By metallographic and ceramographic examination we determinate that the hydride precipitates are orientated parallel to the cladding surface. A content of hydrogen of about 120 ppm was estimated. After the preliminary tests, ring samples were cut from the fuel rod, and were subject of tensile test on an INSTRON 5569 model machine in order to evaluate the changes of their mechanical properties as consequence of irradiation. Scanning electron microscopy was performed on a microscope model TESCAN MIRA II LMU CS with Schottky FE emitter and variable pressure. The analysis shows that the central zone has deeper dimples, whereas on the outer zone, the dimples are tilted and smaller. (authors)

Improvement of test methodology for evaluating diesel fuel stability

Energy Technology Data Exchange (ETDEWEB)

Gutman, M.; Tartakovsky, L.; Kirzhner, Y.; Zvirin, Y. [Internal Combustion Engines Lab., Haifa (Israel); Luria, D. [Fuel Authority, Tel Aviv (Israel); Weiss, A.; Shuftan, M. [Israel Defence Forces, Tel Aviv (Israel)

1995-05-01

The storage stability of diesel fuel has been extensively investigated for many years under laboratory conditions. Although continuous efforts have been made to improve testing techniques, there does not yet exist a generally accepted correlation between laboratory methods (such as chemical analysis of the fuel) and actual diesel engine tests. A testing method was developed by the Technion Internal Combustion Engines Laboratory (TICEL), in order to address this problem. The test procedure was designed to simulate diesel engine operation under field conditions. It is based on running a laboratory-modified single cylinder diesel engine for 50 h under cycling operating conditions. The overall rating of each test is based on individual evaluation of the deposits and residue formation in the fuel filter, nozzle body and needle, piston head, piston rings, exhaust valve, and combustion chamber (six parameters). Two methods for analyzing the test results were used: objective, based on measured data, and subjective, based on visual evaluation results of these deposits by a group of experts. Only the residual level in the fuel filter was evaluated quantitatively by measured results. In order to achieve higher accuracy of the method, the test procedure was improved by introducing the measured results of nozzle fouling as an additional objective evaluating (seventh) parameter. This factor is evaluated on the basis of the change in the air flow rate through the nozzle before and after the complete engine test. Other improvements in the method include the use of the nozzle assembly photograph in the test evaluation, and representation of all seven parameters on a continuous scale instead of the discrete scale used anteriorly, in order to achieve higher accuracy. This paper also contains the results obtained by application of this improved fuel stability test for a diesel fuel stored for a five-year period.

The 3rd irradiation test plan of DUPIC fuel

Energy Technology Data Exchange (ETDEWEB)

Yang, Myung Seung; Song, K. C.; Park, J. H. and others

2001-05-01

The objective of the 3rd irradiation test of DUPIC fuel at the HANARO is to estimate the in-core behaviour of a DUPIC pellet that is irradiated up to more than average burnup of CANDU fuel. The irradiation of DUPIC fuel is planned to start at May 21, 2001, and will be continued at least for 8 months. The burnup of DUPIC fuel through this irradiation test is thought to be more than 7,000 MWd/tHE. The DUPIC irradiation rig instrumented with three SPN detectors will be used to accumulate the experience for the instrumented irradiation and to estimate the burnup of irradiated DUPIC fuel more accurately. Under normal operating condition, the maximum linear power of DUPIC fuel was estimated as 55.06 kW/m, and the centerline temperature of a pellet was calculated as 2510 deg C. In order to assess the integrity of DUPIC fuel under the accident condition postulated at the HANARO, safety analyses on the locked rotor and reactivity insertion accidents were carried out. The maximum centerline temperature of DUPIC fuel was estimated 2590 deg C and 2094 deg C for each accident, respectively. From the results of the safety analysis, the integrity of DUPIC fuel during the HANARO irradiation test will be secured. The irradiated DUPIC fuel will be transported to the IMEF. The post-irradiation examinations are planned to be performed at the PIEF and IMEF.

Loss-of-flow test L5 on FFTF-type irradiated fuel

International Nuclear Information System (INIS)

Simms, R.; Gehl, S.M.; Lo, R.K.; Rothman, A.B.

1978-03-01

Test L5 simulated a hypothetical loss-of-flow accident in an LMFBR using three (Pu, U)O 2 fuel elements of the FTR type. The test elements were irradiated before TREAT Test L5 in the General Electric Test Reactor to 8 at. % burnup at about 40 kW/m. The preirradiation in GETR caused a fuel-restructuring range characteristic of moderate-power structure relative to the FTR. The test transient was devised so that a power burst would be initiated at incipient cladding melting after the loss of flow. The test simulation corresponds to a scenario for FTR in which fuel in high-power-structure subassemblies slump, resulting in a power excursion. The remaining subassemblies are subjected to this power burst. Test L5 addressed the fuel-motion behavior of the subassemblies in this latter category. Data from test-vehicle sensors, hodoscope, and post-mortem examinations were used to construct the sequence of events within the test zone. From these observations, the fuel underwent a predominantly dispersive event just after reaching a peak power six times nominal at, or after, scram. The fuel motion was apparently driven by the release of entrained fission-product gases, since fuel vapor pressure was deliberately kept below significant levels for the transient. The test remains show a wide range of microstructural evolution, depending on the extent of heat deposition along the active fuel column. Extensive fuel swelling was also observed as a result of the lack of the cladding restraint. The results of the thermal-hydraulic calculations with the SAS3A code agreed qualitatively with the postmortem results with respect to the extent of the melting and the dispersal of cladding and fuel. However, the calculated times of certain events did not agree with the observed times

Investigation of thermodynamic cycle for generic 1200 MW{sub el} pressure channel reactor with nuclear steam superheat

Energy Technology Data Exchange (ETDEWEB)

Vincze, A.; Sidawi, K.; Abdullah, R.; Baldock, M.; Saltanov, E.; Pioro, I., E-mail: andrei.vincze@uoit.net, E-mail: khalil.sidawi@uoit.net, E-mail: rand.abdullah@uoit.net, E-mail: matthew.baldock@uoit.net, E-mail: eugene.saltanov@uoit.ca, E-mail: igor.pioro@uoit.ca [Univ. of Ontario Inst. of Tech., Oshawa, ON (Canada)

2014-07-01

Current Nuclear Power Plants (NPPs) play a significant role in energy production around the world. All NPPs operating today employ a Rankine steam cycle for the conversion of thermal power to electricity. This paper will examine the steam cycle arrangement an experimental pressure channel reactor using Nuclear Steam Superheat (NSS) and compare it to two advanced reactor designs, the Advanced CANDU Reactor 1000 (ACR-1000) and the Advanced Boiling Water Reactor (ABWR) designs. The thermodynamic cycle layout and thermal efficiencies of the three reactor types will be discussed. (author)

The development of flow test technology for PWR fuel assembly

International Nuclear Information System (INIS)

Chung, Moon Ki; Cha, Chong Hee; Chung, Chang Hwan; Chun, Se Young; Song, Chul Hwa; Chung, Heung Joon; Won, Soon Yeun; Cho, Yeong Rho; Kim, Bok Deuk

1988-05-01

KAERI has an extensive program to develope PWR fuel assembly. In relation to the program, development of flow test technology is needed to evaluate the thermal hydraulic compactibility and mechanical integrity of domestically fabricated nuclear fuels. A high-pressure and high-temperature flow test facility was designed to test domestically fabricated fuel assembly. The test section of the facility has capacity of a 6x6 full length PWR fuel assembly. A flow test rig was designed and installed at Cold Test Loop to carry out model experiments with 5x5 rod assembly under atmosphere pressure to get information about the characteristics of pressure loss of spacer grids and velocity distribution in the subchannels. LDV measuring technology was established using TSI's Laser Dopper Velocimeter 9100-3 System

Irradiation testing of miniature fuel plates for the RERTR program

Energy Technology Data Exchange (ETDEWEB)

Senn, R L; Martin, M M [Oak Ridge National Laboratory, Oak Ridge, TN 37830 (United States)

1983-08-01

An irradiation test facility, which provides a test bed for irradiating a variety of miniature fuel plates miniplates) for the Reduced Enrichment Research and Test Reactors (RERTR) program, has been placed into operation. The objective of these tests is to screen various candidate fuel materials as to their suitability for replacing the highly enriched uranium fuel materials currently used by the world's test and research reactors with a lower enrichment fuel material, without significantly degrading reactor operating characteristics and power levels. The use of low uranium enrichment of about 20% {sup 235}U in place of highly enriched fuel for these reactors would reduce the potential for {sup 235}U diversion. Fuel materials currently being evaluated in this first phase of these screening tests include aluminum-base dispersion-type fuel plates with fuel cores of 1) high uranium content U{sup 3}){sup 8}-Al being developed by ORNL, 2) high uranium content UAI{sub x}-Al being developed by EG and G Idaho, Inc., and 3) very high uranium content U{sub 3}Si-Al- being developed by ANL. The miniplates are 115-mm long by 50-mm wide with overall plate thicknesses of 1.27 or 1.52 mm. The fuel core dimensions vary according to overall plate thicknesses with a minimal clad thickness requirement of 0.20 mm. Sixty such miniplates (thirty of each thickness) can be irradiated in one test facility. The irradiation test facility, designated as HFED-1 is operating in core position E-7 in the Oak Ridge Research Reactor (ORR), a 30-MW water-moderated reactor. The peak neutron flux measured for this experiment is 1.96 x 10{sup 18} neutrons m{sub -2} s{sub -1}. The various types of miniplates will achieve burnups of up to approximately 2.2x10{sup 27} fissions/m{sup 3} of fuel, which will require approximately eight full power months of irradiation. During reactor shutdown periods, the experiment is removed from the reactor, moved to a special poolside station, disassembled, and inspected

High Temperature Corrosion Problem of Boiler Components in presence of Sulfur and Alkali based Fuels

Science.gov (United States)

Ghosh, Debashis; Mitra, Swapan Kumar

2011-04-01

Material degradation and ageing is of particular concern for fossil fuel fired power plant components. New techniques/approaches have been explored in recent years for Residual Life assessment of aged components and material degradation due to different damage mechanism like creep, fatigue, corrosion and erosion etc. Apart from the creep, the high temperature corrosion problem in a fossil fuel fired boiler is a matter of great concern if the fuel contains sulfur, chlorine sodium, potassium and vanadium etc. This paper discusses the material degradation due to high temperature corrosion in different critical components of boiler like water wall, superheater and reheater tubes and also remedial measures to avoid the premature failure. This paper also high lights the Residual Life Assessment (RLA) methodology of the components based on high temperature fireside corrosion. of different critical components of boiler.

FAST FLUX TEST FACILITY DRIVER FUEL MEETING

Energy Technology Data Exchange (ETDEWEB)

None,

1966-06-01

The Pacific Northwest Laboratory has convened this meeting to enlist the best talents of our laboratories and industry in soliciting factual, technical information pertinent to the Pacific Northwest's Laboratory's evaluation of the potential fuel systems for the Fast Flux Test Facility. The particular factors emphasized for these fuel systems are those associated with safety, ability to meet testing objectives, and economics. The proceedings includes twenty-three presentations, along with a transcript of the discussion following each, as well as a summary discussion.

Characterization and testing of monolithic RERTR fuel plates

Energy Technology Data Exchange (ETDEWEB)

Keiser, D.D.; Jue, J.F.; Burkes, D.E. [Idaho National Lab., Idaho Falls, ID (United States)

2007-07-01

Monolithic fuel plates are being developed as a LEU (low enrichment uranium) fuel for application in research reactors throughout the world. These fuel plates are comprised of a U-Mo alloy foil encased in aluminum alloy cladding. Three different fabrication techniques have been looked at for producing monolithic fuel plates: hot isostatic pressing (HIP), transient liquid phase bonding (TLPB), and friction stir welding (FSW). Of these three techniques, HIP and FSW are currently being emphasized. As part of the development of these fabrication techniques, fuel plates are characterized and tested to determine properties like hardness and the bond strength at the interface between the fuel and cladding. Testing of HIP-made samples indicates that the foil/cladding interaction behavior depends on the Mo content in the UMo foil, the measured hardness values are quite different for the fuel, cladding, and interaction zone phase and Ti, Zr and Nb are the most effective diffusion barriers. For FSW samples, there is a dependence of the bond strength at the foil/cladding interface on the type of tool that is employed for performing the actual FSW process. (authors)

Metallographic analysis of irradiated RERTR-3 fuel test specimens

International Nuclear Information System (INIS)

Meyer, M. K.; Hofman, G. L.; Strain, R. V.; Clark, C. R.; Stuart, J. R.

2000-01-01

The RERTR-3 irradiation test was designed to investigate the irradiation behavior of aluminum matrix U-MO alloy dispersion fuels under high-temperature, high-fission-rate conditions. Initial postirradiation examination of RERTR-3 fuel specimens has concentrated on binary U-MO atomized fuels. The rate of matrix aluminum depletion was found to be higher than predictions based on low temperature irradiation data. Wavelength Dispersive X-ray Spectroscopy (WDS) indicates that aluminum is present in the interior of the fuel particles. WDS data is supported by a mass and volume balance calculation performed on the basis of image analysis results. The depletion of matrix aluminum seems to have no detrimental effects on fuel performance under the conditions tested to date

Drop testing of the Westinghouse fresh nuclear fuel package

International Nuclear Information System (INIS)

Shappert, L.B.; Sanders, C.F.

1992-01-01

In recent years, the Westinghouse Columbia Fuel Fabrication Facility has been faced with increasing pressure from utilities that wished to take the fuel in their nuclear power plants to higher burnups. To help accommodate this trend, Westinghouse has determined that it needs the ability to increase the enrichment of the fresh fuel it delivers to its customers. One critical step in this process is to certify a new (Type A, fissile) fresh fuel package design that has the capability to transport fuel with a higher enrichment than was previously available. A prototype package was tested in support of the Safety Analysis Report of the Packaging. This paper provides detailed information on those tests and their results

Initiation of depleted uranium oxide and spent fuel testing for the spent fuel sabotage aerosol ratio programme

International Nuclear Information System (INIS)

Molecke, M.A.; Gregson, M.W.; Sorenson, K.B.

2004-01-01

We provide a detailed overview of an on-going, multinational test programme that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolised materials plus volatilised fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high-energy/density device. The programme participants in the United States plus Germany, France and the United Kingdom, part of the international Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC) have strongly supported and coordinated this research programme. Sandia National Laboratories has the lead role for conducting this research programme; test programme support is provided by both the US Department of Energy and the US Nuclear Regulatory Commission. We provide a summary of the overall, multiphase test design and a description of all explosive containment and aerosol collection test components used. We focus on the recently initiated tests on 'surrogate' spent fuel, unirradiated depleted uranium oxide and forthcoming actual spent fuel tests. We briefly summarise similar results from completed surrogate tests that used non-radioactive, sintered cerium oxide ceramic pellets in test rods. (author)

Hydrogen Fuel Cell Vehicle Fuel Economy Testing at the U.S. EPA National Vehicle and Fuel Emissions Laboratory (SAE Paper 2004-01-2900)

Science.gov (United States)

The introduction of hydrogen fuel cell vehicles and their new technology has created the need for development of new fuel economy test procedures and safety procedures during testing. The United States Environmental Protection Agency-National Vehicle Fuels and Emissions Laborato...

Imminent: Irradiation Testing of (Th,Pu)O{sub 2} Fuel - 13560

Energy Technology Data Exchange (ETDEWEB)

Kelly, Julian F. [Thor Energy AS, Sommerrogaten 13-15, Oslo 0255 (Norway); Franceschini, Fausto [Westinghouse Electric Company LLC, 1000 Cranberry Woods Drive, Cranberry Township, PA 16066 (United States)

2013-07-01

Commercial-prototype thorium-plutonium oxide (Th-MOX) fuel pellets have been loaded into the material test reactor in Halden, Norway. The fuel is being operated at full power - with instrumentation - in simulated LWR / PHWR conditions and its behaviour is measured 'on-line' as it operates to high burn-up. This is a vital test on the commercialization pathway for this robust new thoria-based fuel. The performance data that is collected will support a fuel modeling effort to support its safety qualification. Several different samples of Th-MOX fuel will be tested, thereby collecting information on ceramic behaviours and their microstructure dependency. The fuel-cycle reasoning underpinning the test campaign is that commercial Th- MOX fuels are an achievable intermediate / near-term SNF management strategy that integrates well with a fast reactor future. (authors)

Updated FY12 Ceramic Fuels Irradiation Test Plan

International Nuclear Information System (INIS)

Nelson, Andrew T.

2012-01-01

The Fuel Cycle Research and Development program is currently devoting resources to study of numerous fuel types with the aim of furthering understanding applicable to a range of reactors and fuel cycles. In FY11, effort within the ceramic fuels campaign focused on planning and preparation for a series of rabbit irradiations to be conducted at the High Flux Isotope Reactor located at Oak Ridge National Laboratory. The emphasis of these planned tests was to study the evolution of thermal conductivity in uranium dioxide and derivative compositions as a function of damage induced by neutron damage. Current fiscal realities have resulted in a scenario where completion of the planned rabbit irradiations is unlikely. Possibilities for execution of irradiation testing within the ceramic fuels campaign in the next several years will thus likely be restricted to avenues where strong synergies exist both within and outside the Fuel Cycle Research and Development program. Opportunities to augment the interests and needs of modeling, advanced characterization, and other campaigns present the most likely avenues for further work. These possibilities will be pursued with the hope of securing future funding. Utilization of synthetic microstructures prepared to better understand the most relevant actors encountered during irradiation of ceramic fuels thus represents the ceramic fuel campaign's most efficient means to enhance understanding of fuel response to burnup. This approach offers many of the favorable attributes embraced by the Separate Effects Testing paradigm, namely production of samples suitable to study specific, isolated phenomena. The recent success of xenon-imbedded thick films is representative of this approach. In the coming years, this strategy will be expanded to address a wider range of problems in conjunction with use of national user facilities novel characterization techniques to best utilize programmatic resources to support a science-based research program.

Material test reactor fuel research at the BR2 reactor

Energy Technology Data Exchange (ETDEWEB)

Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

2012-03-15

The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

Blunt impact tests of retired passenger locomotive fuel tanks

Science.gov (United States)

2017-08-01

The Transportation Technology Center, Inc. conducted impact tests on three locomotive fuel tanks as part of the Federal Railroad Administrations locomotive fuel tank crashworthiness improvement program. Three fuel tanks, two from EMD F40PH locomot...

Municipal solid waste combustion: Fuel testing and characterization

Energy Technology Data Exchange (ETDEWEB)

Bushnell, D.J.; Canova, J.H.; Dadkhah-Nikoo, A.

1990-10-01

The objective of this study is to screen and characterize potential biomass fuels from waste streams. This will be accomplished by determining the types of pollutants produced while burning selected municipal waste, i.e., commercial mixed waste paper residential (curbside) mixed waste paper, and refuse derived fuel. These materials will be fired alone and in combination with wood, equal parts by weight. The data from these experiments could be utilized to size pollution control equipment required to meet emission standards. This document provides detailed descriptions of the testing methods and evaluation procedures used in the combustion testing and characterization project. The fuel samples will be examined thoroughly from the raw form to the exhaust emissions produced during the combustion test of a densified sample.

40 CFR 86.340-79 - Gasoline-fueled engine dynamometer test run.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 18 2010-07-01 2010-07-01 false Gasoline-fueled engine dynamometer... Emission Regulations for New Gasoline-Fueled and Diesel-Fueled Heavy-Duty Engines; Gaseous Exhaust Test Procedures § 86.340-79 Gasoline-fueled engine dynamometer test run. (a) This section applies to gasoline...

Operation of the nuclear fuel cycle test facilities -Operation of the hot test loop facilities

International Nuclear Information System (INIS)

Chun, S. Y.; Jeong, M. K.; Park, C. K.; Yang, S. K.; Won, S. Y.; Song, C. H.; Jeon, H. K.; Jeong, H. J.; Cho, S.; Min, K. H.; Jeong, J. H.

1997-01-01

A performance and reliability of a advanced nuclear fuel and reactor newly designed should be verified by performing the thermal hydraulics tests. In thermal hydraulics research team, the thermal hydraulics tests associated with the development of an advanced nuclear fuel and reactor haven been carried out with the test facilities, such as the Hot Test Loop operated under high temperature and pressure conditions, Cold Test Loop, RCS Loop and B and C Loop. The objective of this project is to obtain the available experimental data and to develop the advanced measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics research team have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for the double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of CANFLEX fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within HANARO fuel bundle and to study a thermal mixing characteristic of PWR fuel bundle. RCS thermal hydraulic loop was constructed and the experiments have been carried out to measure the critical heat flux. In B and C Loop, the performance tests for each component were carried out. (author). 19 tabs., 78 figs., 19 refs

Operation of the nuclear fuel cycle test facilities -Operation of the hot test loop facilities

Energy Technology Data Exchange (ETDEWEB)

Chun, S. Y.; Jeong, M. K.; Park, C. K.; Yang, S. K.; Won, S. Y.; Song, C. H.; Jeon, H. K.; Jeong, H. J.; Cho, S.; Min, K. H.; Jeong, J. H.

1997-01-01

A performance and reliability of a advanced nuclear fuel and reactor newly designed should be verified by performing the thermal hydraulics tests. In thermal hydraulics research team, the thermal hydraulics tests associated with the development of an advanced nuclear fuel and reactor haven been carried out with the test facilities, such as the Hot Test Loop operated under high temperature and pressure conditions, Cold Test Loop, RCS Loop and B and C Loop. The objective of this project is to obtain the available experimental data and to develop the advanced measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics research team have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for the double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of CANFLEX fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within HANARO fuel bundle and to study a thermal mixing characteristic of PWR fuel bundle. RCS thermal hydraulic loop was constructed and the experiments have been carried out to measure the critical heat flux. In B and C Loop, the performance tests for each component were carried out. (author). 19 tabs., 78 figs., 19 refs.

Full-length high-temperature severe fuel damage test No. 5

International Nuclear Information System (INIS)

Lanning, D.D.; Lombardo, N.J.; Hensley, W.K.; Fitzsimmons, D.E.; Panisko, F.E.; Hartwell, J.K.

1993-09-01

This report describes and presents data from a severe fuel damage test that was conducted in the National Research Universal (NRU) reactor at Chalk River Nuclear Laboratories (CRNL), Ontario, Canada. The test, designated FLHT-5, was the fourth in a series of full-length high-temperature (FLHT) tests on light-water reactor fuel. The tests were designed and performed by staff from the US Department of Energy's Pacific Northwest Laboratory (PNL), operated by Battelle Memorial Institute. The test operation and test results are described in this report. The fuel bundle in the FLHT-5 experiment included 10 unirradiated full-length pressurized-water reactor (PWR) rods, 1 irradiated PWR rod and 1 dummy gamma thermometer. The fuel rods were subjected to a very low coolant flow while operating at low fission power. This caused coolant boilaway, rod dryout and overheating to temperatures above 2600 K, severe fuel rod damage, hydrogen generation, and fission product release. The test assembly and its effluent path were extensively instrumented to record temperatures, pressures, flow rates, hydrogen evolution, and fission product release during the boilaway/heatup transient. Post-test gamma scanning of the upper plenum indicated significant iodine and cesium release and deposition. Both stack gas activity and on-line gamma spectrometer data indicated significant (∼50%) release of noble fission gases. Post-test visual examination of one side of the fuel bundle revealed no massive relocation and flow blockage; however, rundown of molten cladding was evident

Analisa Efisiensi Water Tube Boiler Berbahan Bakar Fiber, Cangkang Sawit dan Kulit Kayu Menggunakan Metode Langsung

OpenAIRE

Gaol, Dosma Putra Lumban

2016-01-01

Some of the factors that affect the efficiency of the boiler is a superheater pressure, water feed temperature, steam temperature, the amount of steam produced, the amount of fuel consumption and calorific value fuel combustion. Steamtab chemicallogic use companion software to calculate the value of enthalpy. The aim of this study is to get relations variations in pressure superheater with boiler efficiency, the relationship of variation of temperature feed water to the boiler efficiency, the...

Testing Systems and Results for Advanced Nuclear Fuel Materials

International Nuclear Information System (INIS)

Rooyen, I.J. van; Griffith, G.W.; Garnier, J.E.

2012-01-01

Light Water Reactor Sustainability (LWRS) Program Advanced LWR Nuclear Fuel Development (ALFD) Pathway. Development and testing of high performance fuel cladding identified as high priority to support: enhancement of fuel performance, reliability, and reactor safety. One of the technologies being examined is an advanced fuel cladding made from ceramic matrix composites (CMC) utilizing silicon carbide (SiC) as a structural material supplementing a commercial Zircaloy-4 (Zr-4) tube. A series of out-of-pile tests to fully characterize the SiC CMC hybrid design to produce baseline data. The planned tests are intended to either produce quantitative data or to demonstrate the properties required to achieve two initial performance conditions relative to standard zircaloybased cladding: decreased hydrogen uptake (corrosion) and decreased fretting of the cladding tube under normal operating and postulated accident conditions. These two failure mechanisms account for approximately 70% of all in-pile failures of LWR commercial fuel assemblies

Experimental data report for Test TS-2 reactivity initiated accident test in NSRR with pre-irradiated BWR fuel rod

International Nuclear Information System (INIS)

Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Kobayashi, Shinsho; Yamahara, Takeshi; Sukegawa, Tomohide; Kikuchi, Teruo

1993-02-01

This report presents experimental data for Test TS-2 which was the second test in a series of Reactivity Initiated Accident (RIA) condition test using pre-irradiated BWR fuel rods, performed at the Nuclear Safety Research Reactor (NSRR) in February, 1990. Test fuel rod used in the Test TS-2 was a short sized BWR (7x7) type rod which was fabricated from a commercial rod irradiated at Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79% and a burnup of 21.3Gwd/tU (bundle average). A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The energy deposition of the fuel rod in this test was evaluated to be 72±5cal/g·fuel (66±5cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and, results of pre and post pulse irradiation examinations are described in this report. (author)

Fuel performance analysis for the HAMP-1 mini plate test

Energy Technology Data Exchange (ETDEWEB)

Cho, Byoung Jin; Tahka, Y. W.; Yim, J. S.; Lee, B. H. [KAERI, Daejeon (Korea, Republic of)

2016-05-15

U-7wt%Mo/Al- 5wt%Si dispersion fuel with 8gU/cm{sup 3} is chosen to achieve more efficiency and higher performance than the conventional U{sub 3}Si{sub 2} fuel. As part of the fuel qualification program for the KiJang research reactor (KJRR), three irradiation tests with mini-plates are on the way at the High-flux Advanced Neutron Application Reactor (HANARO). The first test among three HANARO Mini-Plate Irradiation tests (HAMP-1, 2, 3) has completed. PLATE code has been initially developed to analyze the thermal performance of high density U-Mo/Al dispersion fuel plates during irradiation [1]. We upgraded the PLATE code with the latest irradiation results which were implemented by corrosion, thermal conductivity and swelling model. Fuel performance analysis for HAMP-1 was conducted with updated PLATE. This paper presents results of performance evaluation of the HAMP-1. Maximum fuel temperature was obtained 136 .deg., which is far below the preset limit of 200 .deg. for the irradiation test. The meat swelling and corrosion thickness was also confirmed that the developed fuel would behave as anticipated.

Characterization of spent fuel approved testing material: ATM-106

International Nuclear Information System (INIS)

Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thornhill, C.K.

1988-10-01

The characterization data obtained to date are described for Approved Testing Material (ATM)-106 spent fuel from Assembly BT03 of pressurized-water reactor Calvert Cliffs No. 1. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well- characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCWRM) program. ATM-106 consists of 20 full-length irradiated fuel rods with rod-average burnups of about 3700 GJ/kgM (43 MWd/kgM) and expected fission gas release of /approximately/10%. Characterization data include (1) as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) calculated nuclide inventories and radioactivities in the fuel and cladding; and (6) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel rod are being conducted and will be included in planned revisions of this report. 12 refs., 110 figs., 81 tabs

The Fuel Accident Condition Simulator (FACS) furnace system for high temperature performance testing of VHTR fuel

Energy Technology Data Exchange (ETDEWEB)

Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Laug, David V.; Scates, Dawn M.; Reber, Edward L.; Roybal, Lyle G.; Walter, John B.; Harp, Jason M. [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Morris, Robert N. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831 (United States)

2012-10-15

Highlights: Black-Right-Pointing-Pointer A system has been developed for safety testing of irradiated coated particle fuel. Black-Right-Pointing-Pointer FACS system is designed to facilitate remote operation in a shielded hot cell. Black-Right-Pointing-Pointer System will measure release of fission gases and condensable fission products. Black-Right-Pointing-Pointer Fuel performance can be evaluated at temperatures as high as 2000 Degree-Sign C in flowing helium. - Abstract: The AGR-1 irradiation of TRISO-coated particle fuel specimens was recently completed and represents the most successful such irradiation in US history, reaching peak burnups of greater than 19% FIMA with zero failures out of 300,000 particles. An extensive post-irradiation examination (PIE) campaign will be conducted on the AGR-1 fuel in order to characterize the irradiated fuel properties, assess the in-pile fuel performance in terms of coating integrity and fission metals release, and determine the fission product retention behavior during high temperature safety testing. A new furnace system has been designed, built, and tested to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000 Degree-Sign C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, and Eu), iodine, and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator furnace system and the associated

Spent fuel test. Climax data acquisition system integration report

International Nuclear Information System (INIS)

Nyholm, R.A.; Brough, W.G.; Rector, N.L.

1982-06-01

The Spent Fuel Test - Climax (SFT-C) is a test of the retrievable, deep geologic storage of commercially generated, spent nuclear reactor fuel in granitic rock. Eleven spent fuel assemblies, together with 6 electrical simulators and 20 guard heaters, are emplaced 420 m below the surface in the Climax granite at the Nevada Test Site. On June 2, 1978, Lawrence Livermore National Laboratory (LLNL) secured funding for the SFT-C, and completed spent fuel emplacement May 28, 1980. This multi-year duration test is located in a remote area and is unattended much of the time. An extensive array of radiological safety and geotechnical instrumentation is deployed to monitor the test performance. A dual minicomputer-based data acquisition system collects and processes data from more than 900 analog instruments. This report documents the design and functions of the hardware and software elements of the Data Acquisition System and describes the supporting facilities which include environmental enclosures, heating/air-conditioning/humidity systems, power distribution systems, fire suppression systems, remote terminal stations, telephone/modem communications, and workshop areas. 9 figures

Spent Fuel Test - Climax data acquisition system operations manual

International Nuclear Information System (INIS)

Nyholm, R.A.

1983-01-01

The Spent Fuel Test-Climax (SFT-C) is a test of the retrievable, deep geologic storage of commercially generated, spent nuclear reactor fuel in granite rock. Eleven spent fuel assemblies, together with 6 electrical simulators and 20 guard heaters, are emplaced 420 m below the surface in the Climax granite at the US Department of Energy Nevada Test Site. On June 2, 1978, Lawrence Livermore National Laboratory (LLNL) secured funding for the SFT-C, and completed spent fuel emplacement May 28, 1980. The multi-year duration test is located in a remote area and is unattended much of the time. An extensive array of radiological safety and geotechnical instrumentation is deployed to monitor the test performance. A dual minicomputer-based data acquisition system (DAS) collects and processes data from more than 900 analog instruments. This report documents the software element of the LLNL developed SFT-C Data Acquisition System. It defines the operating system and hardware interface configurations, the special applications software and data structures, and support software

RECH-1 test fuel irradiation status report

International Nuclear Information System (INIS)

Marin, J.; Lisboa, J.; Olivares, L.; Chavez, J.

2005-01-01

Since May 2003, one RECH-1 fuel element has been submitted to irradiation at the HFR-Petten, Holland. By November 2004 the irradiation has achieved its pursued goal of 55% burn up. This irradiation qualification service will finish in the year 2005 with PIE tests, as established in a contractual agreement between the IAEA, NRG, and CCHEN. This report presents the objectives and the current results of this fuel qualification under irradiation. Besides, a brief description of CHI/4/021, IAEA's Technical Cooperation Project that has supported this irradiation test, is also presented here. (author)

United States Advanced Ultra-Supercritical Component Test Facility for 760°C Steam Power Plants ComTest Project

Energy Technology Data Exchange (ETDEWEB)

Hack, Horst [Electric Power Research Institute (EPRI); Purgert, Robert Michael [Energy Industries of Ohio

2017-12-13

Following the successful completion of a 15-year effort to develop and test materials that would allow coal-fired power plants to be operated at advanced ultra-supercritical (A-USC) steam conditions, a United States-based consortium is presently engaged in a project to build an A-USC component test facility (ComTest). A-USC steam cycles have the potential to improve cycle efficiency, reduce fuel costs, and reduce greenhouse gas emissions. Current development and demonstration efforts are focused on enabling the construction of A-USC plants, operating with steam temperatures as high as 1400°F (760°C) and steam pressures up to 5000 psi (35 MPa), which can potentially increase cycle efficiencies to 47% HHV (higher heating value), or approximately 50% LHV (lower heating value), and reduce CO₂ emissions by roughly 25%, compared to today’s U.S. fleet. A-USC technology provides a lower-cost method to reduce CO₂ emissions, compared to CO₂ capture technologies, while retaining a viable coal option for owners of coal generation assets. Among the goals of the ComTest facility are to validate that components made from advanced nickel-based alloys can operate and perform under A-USC conditions, to accelerate the development of a U.S.-based supply chain for the full complement of A-USC components, and to decrease the uncertainty of cost estimates for future A-USC power plants. The configuration of the ComTest facility would include the key A-USC technology components that were identified for expanded operational testing, including a gas-fired superheater, high-temperature steam piping, steam turbine valve, and cycling header component. Membrane walls in the superheater have been designed to operate at the full temperatures expected in a commercial A-USC boiler, but at a lower (intermediate) operating pressure. This superheater has been designed to increase the temperature of the steam supplied by the host utility boiler up to 1400°F (760�

Development of a non-engine fuel injector deposit test for alternative fuels (ENIAK-project)

Energy Technology Data Exchange (ETDEWEB)

Hoffmann, Hajo; Pohland vom Schloss, Heide [OWI - Oel Waerme Institut GmbH, Herzogenrath (Germany)

2013-06-01

Deposit formation in and on the injectors of diesel engines may lead to injector malfunction, resulting in a loss in power, rough engine operation and poor emission levels. Poor Biodiesel quality, contamination with copper and zinc as well as undesired reactions between (several) additives and biodiesel components are known causes for nozzle fouling. Therefore, good housekeeping when using biodiesel is required, and all additives have to pass a no-harm test concerning injector fouling. The standard fouling tests are two engine tests: The XUD9-test (CEC F-23-01) and the DW-10-test (CEC DF 98-08). The XUD9 is a cost efficient, fast and proven testing method. It uses, however, an obsolete indirect injection diesel engine and cannot reproduce internal diesel injector deposits (IDID). The newer DW10 test is complex, costly and designed for high stress. This reduces the engine life and leads to a fuel consumption of approximately 1,000 1 per test, both contributing to the high costs of the test. The ENIAK-Project is funded by the FNR (''Fachagentur Nachwachsende Rohstoffe'', Agency for Renewable Resources) and conducted in cooperation with AGQM, ASG and ERC. Its main goal is the development, assembly, commissioning, and evaluation of a non-engine fuel injector test. It uses a complete common rail system. The injection takes place in a self-designed reactor instead of an engine, and the fuel is not combusted, but re-condensed and pumped in a circle, leading to a low amount of fuel required. If the test method proves to be as reliable as expected, it can be used as an alternative test method for injector fouling with low requirements regarding infrastructure on the testing site and sample volume. (orig.)

Autonomy-Enabled Fuel Savings for Military Vehicles: Report on 2016 Aberdeen Test Center Testing

Energy Technology Data Exchange (ETDEWEB)

Ragatz, Adam [National Renewable Energy Lab. (NREL), Golden, CO (United States); Prohaska, Robert [National Renewable Energy Lab. (NREL), Golden, CO (United States); Gonder, Jeff [National Renewable Energy Lab. (NREL), Golden, CO (United States)

2017-05-26

Fuel savings have never been the primary focus for autonomy-enabled military vehicles. However, studies have estimated that autonomy in passenger and commercial vehicles could improve fuel economy by as much as 22%-33% over various drive cycles. If even a fraction of this saving could be realized in military vehicles, significant cost savings could be realized each year through reduced fuel transport missions, reduced fuel purchases, less maintenance, fewer required personnel, and increased vehicle range. Researchers from the National Renewable Energy Laboratory installed advanced data logging equipment and instrumentation on two autonomy-enabled convoy vehicles configured with Lockheed Martin's Autonomous Mobility Applique System to determine system performance and improve on the overall vehicle control strategies of the vehicles. Initial test results from testing conducted at the U.S. Army Aberdeen Test Center at the Aberdeen Proving Grounds are included in this report. Lessons learned from in-use testing and performance results have been provided to the project partners for continued system refinement.

Design criteria of out-pile system of HANARO fuel test loop

Energy Technology Data Exchange (ETDEWEB)

Kim, J. Y.

1997-07-01

The objective of HANARO aims at the development and localization of nuclear technologies through the engineering tests. Thus it is very important the design and installation of the irradiation test facilities to be installed at the irradiation hole for verification test of the fuel performance are in connection with maximization of the utilization of HANARO. The principle subjects of this study are to presend and informed the detail design criteria and technical specification of out-pile system of HANARO fuel test loop for the developing of the fuel and reactor material. This results will become guidance for the planning of the irradiation testing using the HANARO fuel test loop. (author). 16 refs., 31 tabs., 9 figs.

Characterization of spent fuel approved testing material---ATM-105

Energy Technology Data Exchange (ETDEWEB)

Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

1991-12-01

The characterization data obtained to data are described for Approved Testing Material 105 (ATM-105), which is spent fuel from Bundles CZ346 and CZ348 of the Cooper Nuclear Power Plant, a boiling-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-105 consists of 88 full-length irradiated fuel rods with rod-average burnups of about 2400 GJ/kgM (28 MWd/kgM) and expected fission gas release of about 1%. Characterization data include (1) descriptions of as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel are being conducted and will be included in planned revisions of this report.

Fabrication, inspection, and test plan for the Advanced Test Reactor (ATR) Mixed-Oxide (MOX) fuel irradiation project

International Nuclear Information System (INIS)

Wachs, G.W.

1997-11-01

The Department of Energy (DOE) Fissile Materials Disposition Materials Disposition Program (FMDP) has announced that reactor irradiation of MOX fuel is one of the preferred alternatives for disposal of surplus weapons-usable plutonium (Pu). MOX fuel has been utilized domestically in test reactors and on an experimental basis in a number of Commercial Light Water Reactors (CLWRs). Most of this experience has been with Pu derived from spent low enriched uranium (LEU) fuel, known as reactor grade (RG) Pu. The MOX fuel test will be irradiated in the ATR to provide preliminary data to demonstrate that the unique properties of surplus weapons-derived or weapons-grade (WG) plutonium (Pu) do not compromise the applicability of this MOX experience base. In addition, the test will contribute experience with irradiation of gallium-containing fuel to the data base required for resolution of generic CLWR fuel design issues (ORNL/MD/LTR-76). This Fabrication, Inspection, and Test Plan (FITP) is a level 2 document as defined in the FMDP LWR MOX Fuel Irradiation Test Project Plan (ORNL/MD/LTR-78)

Advanced Fuel/Cladding Testing Capabilities in the ORNL High Flux Isotope Reactor

International Nuclear Information System (INIS)

Ott, Larry J.; Ellis, Ronald James; McDuffee, Joel Lee; Spellman, Donald J.; Bevard, Bruce Balkcom

2009-01-01

The ability to test advanced fuels and cladding materials under reactor operating conditions in the United States is limited. The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) and the newly expanded post-irradiation examination (PIE) capability at the ORNL Irradiated Fuels Examination Laboratory provide unique support for this type of advanced fuel/cladding development effort. The wide breadth of ORNL's fuels and materials research divisions provides all the necessary fuel development capabilities in one location. At ORNL, facilities are available from test fuel fabrication, to irradiation in HFIR under either thermal or fast reactor conditions, to a complete suite of PIEs, and to final product disposal. There are very few locations in the world where this full range of capabilities exists. New testing capabilities at HFIR have been developed that allow testing of advanced nuclear fuels and cladding materials under prototypic operating conditions (i.e., for both fast-spectrum conditions and light-water-reactor conditions). This paper will describe the HFIR testing capabilities, the new advanced fuel/cladding testing facilities, and the initial cooperative irradiation experiment that begins this year.

Test of Flow Characteristics in Tubular Fuel Assembly I - Establishment of test loop and measurement validation test

International Nuclear Information System (INIS)

Park, Jong Hark; Chae, H. T.; Park, C.; Kim, H.

2005-12-01

Tubular type fuel has been developed as one of candidates for Advanced HANARO Reactor(AHR). It is necessary to test the flow characteristics such as velocity in each flow channels and pressure drop of tubular type fuel. A hydraulic test-loop to examine the hydraulic characteristics for a tubular type fuel has been designed and constructed. It consists of three parts; a) piping-loop including pump and motor, magnetic flow meter and valves etc, b) test-section part where a simulated tubular type fuel is located, and 3) data acquisition system to get reading signals from sensors or instruments. In this report, considerations during the design and installation of the facility and the selection of data acquisition sensors and instruments are described in detail. Before doing the experiment to measure the flow velocities in flow channels, a preliminary tests have been done for measuring the coolant velocities using pitot-tube and for validating the measurement accuracy as well. Local velocities of the radial direction in circular tubes are measured at regular intervals of 60 degrees by three pitot-tubes. Flow rate inside the circular flow channel can be obtained by integrating the velocity distribution in radial direction. The measured flow rate was compared to that of magnetic flow meter. According to the results, two values had a good agreement, which means that the measurement of coolant velocity by using pitot-tube and the flow rate measured by the magnetic flow meter are reliable. Uncertainty analysis showed that the error of velocity measurement by pitot-tube is less than ±2.21%. The hydraulic test-loop also can be adapted to others such as HANARO 18 and 36 fuel, in-pile system of FTL(Fuel Test Loop), etc

Development of TIG Welding System for a Nuclear Fuel Test Rig

Energy Technology Data Exchange (ETDEWEB)

Joung, Changyoung; Ahn, Sungho; Hong, Jintae; Kim, Kahye [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rig and rods. To manufacture the nuclear fuel test rig, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rig and rods jointing the various sensors and end caps on a fuel cladding tube, which is charged with fuel pellets and component parts. Thus, we designed and fabricated the precision welding system consisting of an orbital TIG welder, a low-pressure chamber, and a high-pressure chamber. Using this system, the performance tests were performed with the round and seal spot welds for each welding condition. This paper describes not only the contents for the fabrication of precision TIG welding system but also some results from weld tests using the low-pressure and high-pressure chambers to verify the performance of this system. The TIG welding system was developed to manufacture the nuclear fuel test rig and rods. It has been configured to be able to weld the nuclear fuel test rigs and rods by applying the TIG welder using a low-pressure chamber and a high-pressure chamber. The performance tests using this system were performed with the round and seal spot welds for the welding conditions. The soundness of the orbital TIG welding system was confirmed through performance tests in the low-pressure and high-pressure chambers.

Development of TIG Welding System for a Nuclear Fuel Test Rig

International Nuclear Information System (INIS)

Joung, Changyoung; Ahn, Sungho; Hong, Jintae; Kim, Kahye

2013-01-01

The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rig and rods. To manufacture the nuclear fuel test rig, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rig and rods jointing the various sensors and end caps on a fuel cladding tube, which is charged with fuel pellets and component parts. Thus, we designed and fabricated the precision welding system consisting of an orbital TIG welder, a low-pressure chamber, and a high-pressure chamber. Using this system, the performance tests were performed with the round and seal spot welds for each welding condition. This paper describes not only the contents for the fabrication of precision TIG welding system but also some results from weld tests using the low-pressure and high-pressure chambers to verify the performance of this system. The TIG welding system was developed to manufacture the nuclear fuel test rig and rods. It has been configured to be able to weld the nuclear fuel test rigs and rods by applying the TIG welder using a low-pressure chamber and a high-pressure chamber. The performance tests using this system were performed with the round and seal spot welds for the welding conditions. The soundness of the orbital TIG welding system was confirmed through performance tests in the low-pressure and high-pressure chambers

Experimental data report for Test TS-1 Reactivity Initiated Accident Test in NSRR with pre-irradiated BWR fuel rod

International Nuclear Information System (INIS)

Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Horiki, Ohichiro; Yamahara, Takeshi; Ichihashi, Yoshinori; Kikuchi, Teruo

1992-01-01

This report presents experimental data for Test TS-1 which was the first in a series of tests, simulating Reactivity Initiated Accident (RIA) conditions using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in October, 1989. Test fuel rod used in the Test TS-1 was a short-sized BWR (7 x 7) type rod which was fabricated from a commercial rod provided from Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79 % and burnup of 21.3 GWd/t (bundle average). Pulse irradiation was performed at a condition of stagnant water cooling, atmospheric pressure and ambient temperature using a newly developed double container-type capsule. Energy deposition of the rod in this test was evaluated to be about 61 cal/g·fuel (55 cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, fuel burnup measurements, transient behavior of the test rod during pulse irradiation and results of post pulse irradiation examinations are contained in this report. (author)

Pie technique of LWR fuel cladding fracture toughness test

International Nuclear Information System (INIS)

Endo, Shinya; Usami, Koji; Nakata, Masahito; Fukuda, Takuji; Numata, Masami; Kizaki, Minoru; Nishino, Yasuharu

2006-01-01

Remote-handling techniques were developed by cooperative research between the Department of Hot Laboratories in the Japan Atomic Energy Research Institute (JAERI) and the Nuclear Fuel Industries Ltd. (NFI) for evaluating the fracture toughness on irradiated LWR fuel cladding. The developed techniques, sample machining by using the electrical discharge machine (EDM), pre-cracking by fatigue tester, sample assembling to the compact tension (CT) shaped test fixture gave a satisfied result for a fracture toughness test developed by NFL. And post-irradiation examination (PIE) using the remote-handling techniques were carried out to evaluate the fracture toughness on BWR spent fuel cladding in the Waste Safety Testing Facility (WASTEF). (author)

Demonstration tests for HTGR fuel elements and core components with test sections in HENDEL

Energy Technology Data Exchange (ETDEWEB)

Miyamoto, Yoshiaki; Hino, Ryutaro; Inagaki, Yoshiyuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

1995-03-01

In the fuel stack test section (T{sub 1}) of the Helium Engineering Demonstration Loop (HENDEL), thermal and hydraulic performances of helium gas flows through a fuel rod channel and a fuel stack have been investigated for the High-Temperature Engineering Test Reactor (HTTR) core thermal design. The test data showed that the turbulent characteristics appearing in the Reynolds number above 2000: no typical behavior in the transition zone, and friction factors and heat transfer coefficients in the fuel channel were found to be higher than those in a smooth annular channel. Heat transfer behavior of gas flow in a fuel element channel with blockage and cross-flow through a gap between upper and lower fuel elements stacked was revealed using the mock-up models. On the other hand, demonstration tests have been performed to verify thermal and hydraulic characteristics and structural integrity related to the core bottom structure using a full-scale test facility named as the in-core structure test section (T{sub 2}). The sealing performance test revealed that the leakage of low-temperature helium gas through gaps between the permanent reflector blocks to the core was very low level compared with the HTTR design value and no change of the leakage flow rate were observed after a long term operation. The heat transfer tests including thermal transient at shutdown of gas circulators verified good insulating performance of core insulation structures in the core bottom structure and the hot gas duct; the temperature of the metal portion of these structure was below the design value. Examination of the thermal mixing characteristics indicated that the mixing of the hot helium gas started at a hot plenum and finished completely at downstream of the outlet hot gas duct. The present results obtained from these demonstration tests have been practically applied to the detailed design works and licensing procedures of the HTTR. (J.P.N.) 92 refs.

Corrosion of spent Advanced Test Reactor fuel

International Nuclear Information System (INIS)

Lundberg, L.B.; Croson, M.L.

1994-01-01

The results of a study of the condition of spent nuclear fuel elements from the Advanced Test Reactor (ATR) currently being stored underwater at the Idaho National Engineering Laboratory (INEL) are presented. This study was motivated by a need to estimate the corrosion behavior of dried, spent ATR fuel elements during dry storage for periods up to 50 years. The study indicated that the condition of spent ATR fuel elements currently stored underwater at the INEL is not very well known. Based on the limited data and observed corrosion behavior in the reactor and in underwater storage, it was concluded that many of the fuel elements currently stored under water in the facility called ICPP-603 FSF are in a degraded condition, and it is probable that many have breached cladding. The anticipated dehydration behavior of corroded spent ATR fuel elements was also studied, and a list of issues to be addressed by fuel element characterization before and after forced drying of the fuel elements and during dry storage is presented

Final report of fuel dynamics Test E7

International Nuclear Information System (INIS)

Doerner, R.C.; Murphy, W.F.; Stanford, G.S.; Froehle, P.H.

1977-04-01

Test data from an in-pile failure experiment of high-power LMFBR-type fuel pins in a simulated $3/s transient-overpower (TOP) accident are reported and analyzed. Major conclusions are that (1) a series of cladding ruptures during the 100-ms period preceding fuel release injected small bursts of fission gas into the flow stream; (2) gas release influenced subsequent cladding melting and fuel release [there were no measurable FCI's (fuel-coolant interactions), and all fuel motion observed by the hodoscope was very slow]; (3) the predominant postfailure fuel motion appears to be radial swelling that left a spongy fuel crust on the holder wall; (4) less than 4 to 6 percent of the fuel moved axially out of the original fuel zone, and most of this froze within a 10-cm region above the original top of the fuel zone to form the outlet blockage. An inlet blockage approximately 1 cm long was formed and consisted of large interconnected void regions. Both blockages began just beyond the ends of the fuel pellets

New JMTR irradiation test plan on fuels and materials

International Nuclear Information System (INIS)

Nakamura, Takehiko; Nishiyama, Yutaka; Chimi, Yasuhiro; Sasajima, Hideo; Ogiyanagi, Jin; Nakamura, Jinichi; Suzuki, Masahide; Kawamura, Hiroshi

2009-01-01

In order to maintain and enhance safety of light water reactors (LWRs) in long-term and up-graded operations, proper understanding of irradiation behavior of fuels and materials is essentially important. Japanese government and the Japan Atomic Energy Agency (JAEA) have decided to refurbish the Japan Materials Testing Reactor (JMTR) and to install new tests rigs, in order to play an active role for solving irradiation related issues on plant aging and high-duty uses of the current LWRs and on development of next-generation reactors. New tests on fuel integrity under simulated abnormal transients and high-duty irradiation conditions are planned in the JMTR. Power ramp tests of newdesign fuel rods will also be performed in the first stage of the program, which is expected to start in year 2011 after refurbishment of the JMTR. Combination of the JMTR tests with simulated reactivity initiated accident tests in the Nuclear Safety Research Reactor (NSRR) and loss of coolant accident tests in hot laboratories would serve as the integrated fuel safety research on the high performance fuels at extended burnups, covering from the normal to the accident conditions, including abnormal transients. For the materials irradiation, fracture toughness of reactor vessel steels and stress corrosion cracking behavior of stainless steels are being studied in addition to basic irradiation behavior of nuclear materials such as hafnium. The irradiation studies would contribute not only to solve the current problems but also to identify possible seeds of troubles and to make proactive responses. (author)

Spent fuel test-climax: a test of geologic storage of high-level waste in granite

International Nuclear Information System (INIS)

Ramspott, L.D.; Ballou, L.B.; Patrick, W.C.

1981-01-01

A test of retrievable geologic storage of spent fuel assemblies from an operating commercial nuclear reactor is underway at the Nevada Test Site (NTS) of the US Department of Energy. This generic test is located 420 m below the surface in the Climax granitic stock. Eleven canisters of spent fuel approximately 2.5 years out of reactor core (about 1.6 kW/canister thermal output) were emplaced in a storage drift along with 6 electrical simulator canisters. Two adjacent drifts contain electrical heaters, which are operated to simulate within the test array the thermal field of a large repository. Fuel was loaded during April to May 1980 and initial results of the test will be presented

Development of Out-pile Test Technology for Fuel Assembly Performance Verification

Energy Technology Data Exchange (ETDEWEB)

Chun, Tae Hyun; In, W. K.; Oh, D. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)] (and others)

2007-03-15

Out-pile tests with full scale fuel assembly are to verify the design and to evaluate the performance of the final products. HTL for the hydraulic tests and FAMeCT for mechanical/structural tests were constructed in this project. The maximum operating conditions of HTL are 30 bar, 320 .deg. C, and 500 m3/hr. This facility can perform the pressure drop test, fuel assembly uplift test, and flow induced vibration test. FAMeCT can perform the bending and vibration tests. The verification of the developed facilities were carried out by comparing the reference data of the fuel assembly which was obtained at the Westinghouse Co. The compared data showed a good coincidence within uncertainties. FRETONUS was developed for high temperature and high pressure fretting wear simulator and performance test. A performance test was conducted for 500 hours to check the integrity, endurance, data acquisition capability of the simulator. The technology of turbulent flow analysis and finite element analysis by computation was developed. From the establishments of out-pile test facilities for full scale fuel assembly, the domestic infrastructure for PWR fuel development has been greatly upgraded.

Performance evaluation of UO2-Zr fuel in power ramp tests

International Nuclear Information System (INIS)

Knudsen, P.; Bagger, C.

1977-01-01

In power reactors using UO 2 -Zr fuel, rapid power increases may lead to failures in fuel pins that have been irradiated at steady or decreasing heat loads. This paper presents results which extend the experience with power ramp performance of high burn-up fuel pins. A test fuel element containing both pellet and vipac UO 2 -Zr fuel pins was irradiated in the HBWR at Halden for effectively 2 1/2 years to an average burn-up of 21,000 MWD/te UO 2 at gradually decreasing power levels. The subsequent non-destructive characterization revealed formation of transverse cracks in the vipac fuel columns. After the HBWR irradiation, five of the fuel pins were power ramp tested individually in the DR 3 Reactor at Riso. The ramp rates in this test series were in the range 3-60 W/cm min. The maximum local heat loads seen in the ramp tests were 20-120% above the highest levels experienced at the same axial positions during the HBWR irradiation. Three pellets and one vipac fuel pin failed, whereas another vipac pin gave no indication of clad penetration. Profilometry after the ramp testing indicated the formation of small ridges for both types of fuel pins. For vipac fuel, the ridges were less regularly distributed along the pin length than for pellet fuel. Neutron radiography revealed the formation of additional transverse and longitudinal fuel cracks during the power ramps for both types of fuel pins. The observed failures seemed to be marginal since little or no indication as to the locations of the clad penetrations could be derived from the non-destructive post-irradiation examinations. The cases have been analyzed by means of the Danish fuel performance codes. The calculations, which are in general agreement with the observations, are discussed. The results of the investigations indicate qualitative similarities in over power performance of the two fuel types

Masonry fireplace emissions test method: Repeatability and sensitivity to fueling protocol.

Science.gov (United States)

Stern, C H; Jaasma, D R; Champion, M R

1993-03-01

A test method for masonry fireplaces has been evaluated during testing on six masonry fireplace configurations. The method determines carbon monoxide and particulate matter emission rates (g/h) and factors (g/kg) and does not require weighing of the appliance to determine the timing of fuel loading.The intralaboratory repeatability of the test method has been determined from multiple tests on the six fireplaces. For the tested fireplaces, the ratio of the highest to lowest measured PM rate averaged 1.17 and in no case was greater than 1.32. The data suggest that some of the variation is due to differences in fuel properties.The influence of fueling protocol on emissions has also been studied. A modified fueling protocol, tested in large and small fireplaces, reduced CO and PM emission factors by roughly 40% and reduced CO and PM rates from 0 to 30%. For both of these fireplaces, emission rates were less sensitive to fueling protocol than emission factors.

76 FR 5319 - Regulation of Fuel and Fuel Additives: Alternative Test Method for Olefins in Gasoline

Science.gov (United States)

2011-01-31

... Regulation of Fuel and Fuel Additives: Alternative Test Method for Olefins in Gasoline AGENCY: Environmental... gasoline. This proposed rule will provide flexibility to the regulated community by allowing an additional... A. Alternative Test Method for Olefins in Gasoline III. Statutory and Executive Order Reviews A...

Mechanical characterization tests of the X2-Gen fuel assembly and skeleton

International Nuclear Information System (INIS)

Kim, Hyung Kyu; Yoon, Kyung Ho; Lee, Kang Hee; Kim, Jae Yong; Lee, Young Ho; Kang, Heung Seok

2011-01-01

The KNF (KEPCO Nuclear Fuel) requested mechanical characterization tests of a fuel assembly and a skeleton of the X2-Gen fuel. The tests consisted of the lateral vibration and lateral/axial stiffness, lateral/axial impact and combined deflection tests carried out by using the FAMeCT (Fuel Assembly Mechanical Characterization Tester) in KAERI. The upper and lower core plate simulators were newly designed and manufactured because the fuel geometry of the X2-Gen was different from the KSNP type fuel assembly. In addition to this, the upper carriage was also revised with the LM guide system from the previous two guide rods system. Therefore, the axial and combined deflection tests were soundly executed. Each test was repeated twice to confirm the repeatability. The discrepancy from the repetition was small enough to be neglected. The mechanical characterization tests were accredited with the KOLAS (Korea Laboratory Accreditation Scheme) standard, and the certified test reports (lateral vibration, lateral/axial bending and lateral/axial impact) and the uncertified test report (combined deflection) were issued together with the current test result report

Transient fuel and target performance testing for the HWR-NPR

International Nuclear Information System (INIS)

Jicha, J.J. Jr.

1990-01-01

This paper describes a five year program of fuel target transient performance testing and model development required for the safety assessment of the HWR new production reactor. Technical issues are described, focusing on fuel and target behavior during extremely low probability transients which can lead to fuel melting. Early work on these issues is reviewed. The program to meet remaining needs is described. Three major transient-testing activities are included: in-cell experiments on small samples of irradiated fuel and target, small-scale phenomenological experiments in the ACRR reactor, and limited-integral experiments in the TREAT reactor. A coordinated development of detailed fuel and target behavior models is also described

Combustion tests in a solid fuel boiler to clarify the emissions when co-firing refuse; Proveldning i fastbraenslepanna foer att kartlaegga emissioner vid inblandning av olika avfallsfraktioner

Energy Technology Data Exchange (ETDEWEB)

Blom, Elisabet; Lundborg, Rickard; Wrangensten, Lars

2002-04-01

}); The highest HCI emission, before the flue gas condensing plant, was noted for the reference fuel (approx. 245 Mg/nm{sup 3}); The highest SO{sub 2} emission, in the upper furnace, was also noted for the reference fuel (over 50 mg/nm{sup 3}). Levels after the condensing plant could not be measured. 25 % meat powder injection gave the highest NO{sub x}-emission (approx. 50 mg/MJ). The high heavy metal content in ash could not bee correlated to the content in the fuel fractions. The unburnt carbon content in bottom ash is remarkably lower for 25 % meat powder injection into reference fuel. The conclusion is that the operation parameters for the oiler are optimized for this specific fuel fraction and the amount injected. Tests with a fouling probe near the first superheater showed that at a temperature of 400 deg C 10% napkin waste gave most fouling. This is interesting as the lowest HCl-emission level was noted for napkins, maybe depending on that the chlorine is present in the depositions. The project results point in the direction that co-combustion of different refuse fractions will be possible within the existing environmental legislation and maybe also within the new EU directive 2000/76/EG. Although flue gas condensation is necessary to get acceptable emissions of HCl and SO{sub 2}. Further general research is thus needed and several R and D project areas are listed in the report. Deeper studies of the combustion process in furnace are important in future project.

Development of the advanced PHWR technology -Verification tests for CANDU advanced fuel-

Energy Technology Data Exchange (ETDEWEB)

Jung, Jang Hwan; Suk, Hoh Chun; Jung, Moon Kee; Oh, Duk Joo; Park, Joo Hwan; Shim, Kee Sub; Jang, Suk Kyoo; Jung, Heung Joon; Park, Jin Suk; Jung, Seung Hoh; Jun, Ji Soo; Lee, Yung Wook; Jung, Chang Joon; Byun, Taek Sang; Park, Kwang Suk; Kim, Bok Deuk; Min, Kyung Hoh [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-07-01

This is the `94 annual report of the CANDU advanced fuel verification test project. This report describes the out-of pile hydraulic tests at CANDU-hot test loop for verification of CANFLEX fuel bundle. It is also describes the reactor thermal-hydraulic analysis for thermal margin and flow stability. The contents in this report are as follows; (1) Out-of pile hydraulic tests for verification of CANFLEX fuel bundle. (a) Pressure drop tests at reactor operation condition (b) Strength test during reload at static condition (c) Impact test during reload at impact load condition (d) Endurance test for verification of fuel integrity during life time (2) Reactor thermal-hydraulic analysis with CANFLEX fuel bundle. (a) Critical channel power sensitivity analysis (b) CANDU-6 channel flow analysis (c) Flow instability analysis. 61 figs, 29 tabs, 21 refs. (Author).

Development of the advanced PHWR technology -Verification tests for CANDU advanced fuel-

International Nuclear Information System (INIS)

Jung, Jang Hwan; Suk, Hoh Chun; Jung, Moon Kee; Oh, Duk Joo; Park, Joo Hwan; Shim, Kee Sub; Jang, Suk Kyoo; Jung, Heung Joon; Park, Jin Suk; Jung, Seung Hoh; Jun, Ji Soo; Lee, Yung Wook; Jung, Chang Joon; Byun, Taek Sang; Park, Kwang Suk; Kim, Bok Deuk; Min, Kyung Hoh

1995-07-01

This is the '94 annual report of the CANDU advanced fuel verification test project. This report describes the out-of pile hydraulic tests at CANDU-hot test loop for verification of CANFLEX fuel bundle. It is also describes the reactor thermal-hydraulic analysis for thermal margin and flow stability. The contents in this report are as follows; (1) Out-of pile hydraulic tests for verification of CANFLEX fuel bundle. (a) Pressure drop tests at reactor operation condition (b) Strength test during reload at static condition (c) Impact test during reload at impact load condition (d) Endurance test for verification of fuel integrity during life time (2) Reactor thermal-hydraulic analysis with CANFLEX fuel bundle. (a) Critical channel power sensitivity analysis (b) CANDU-6 channel flow analysis (c) Flow instability analysis. 61 figs, 29 tabs, 21 refs. (Author)

Alpha Fuels Environmental Test Facility impact gun

International Nuclear Information System (INIS)

Anderson, C.G.

1978-01-01

The Alpha Fuels Environmental Test Facility (AFETF) impact gun is a unique tool for impact testing 238 PuO 2 -fueled heat sources of up to 178-mm dia at velocities to 300 m/s. An environmentally-sealed vacuum chamber at the muzzle of the gun allows preheating of the projectile to 1,000 0 C. Immediately prior to impact, the heat source projectile is completely sealed in a vacuum-tight catching container to prevent escape of its radioactive contents should rupture occur. The impact velocity delivered by this gas-powered gun can be regulated to within +-2%

Fuels for research and test reactors, status review: July 1982

International Nuclear Information System (INIS)

Stahl, D.

1982-12-01

A thorough review is provided on nuclear fuels for steady-state thermal research and test reactors. The review was conducted to provide a documented data base in support of recent advances in research and test reactor fuel development, manufacture, and demonstration in response to current US policy on availability of enriched uranium. The review covers current fabrication practice, fabrication development efforts, irradiation performance, and properties affecting fuel utilization, including thermal conductivity, specific heat, density, thermal expansion, corrosion, phase stability, mechanical properties, and fission-product release. The emphasis is on US activities, but major work in Europe and elsewhere is included. The standard fuel types discussed are the U-Al alloy, UZrH/sub x/, and UO 2 rod fuels. Among new fuels, those given major emphasis include H 3 Si-Al dispersion and UO 2 caramel plate fuels

Spent nuclear fuel integrity during dry storage - performance tests and demonstrations

International Nuclear Information System (INIS)

McKinnon, M.A.; Doherty, A.L.

1997-06-01

This report summarizes the results of fuel integrity surveillance determined from gas sampling during and after performance tests and demonstrations conducted from 1983 through 1996 by or in cooperation with the US DOE Office of Commercial Radioactive Waste Management (OCRWM). The cask performance tests were conducted at Idaho National Engineering Laboratory (INEL) between 1984 and 1991 and included visual observation and ultrasonic examination of the condition of the cladding, fuel rods, and fuel assembly hardware before dry storage and consolidation of fuel, and a qualitative determination of the effects of dry storage and fuel consolidation on fission gas release from the spent fuel rods. The performance tests consisted of 6 to 14 runs involving one or two loading, usually three backfill environments (helium, nitrogen, and vacuum backfills), and one or two storage system orientations. The nitrogen and helium backfills were sampled and analyzed to detect leaking spent fuel rods. At the end of each performance test, periodic gas sampling was conducted on each cask. A spent fuel behavior project (i.e., enhanced surveillance, monitoring, and gas sampling activities) was initiated by DOE in 1994 for intact fuel in a CASTOR V/21 cask and for consolidated fuel in a VSC-17 cask. The results of the gas sampling activities are included in this report. Information on spent fuel integrity is of interest in evaluating the impact of long-term dry storage on the behavior of spent fuel rods. Spent fuel used during cask performance tests at INEL offers significant opportunities for confirmation of the benign nature of long-term dry storage. Supporting cask demonstration included licensing and operation of an independent spent fuel storage installation (ISFSI) at the Virginia Power (VP) Surry reactor site. A CASTOR V/21, an MC-10, and a Nuclear Assurance NAC-I28 have been loaded and placed at the VP ISFSI as part of the demonstration program. 13 refs., 14 figs., 9 tabs

40 CFR 86.335-79 - Gasoline-fueled engine test cycle.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 18 2010-07-01 2010-07-01 false Gasoline-fueled engine test cycle. 86....335-79 Gasoline-fueled engine test cycle. (a) The following test sequence shall be followed in... operating the engine at the higher approved load setting during cycle 1 and at the lower approved load...

Performance test of the I and C system (GSF - 2002) for the irradiation tests using a fuel capsule

Energy Technology Data Exchange (ETDEWEB)

Kang, Young Hwan; Park, S. J.; Kim, B. G.; Ahn, D. H

2004-12-01

HANARO is a very important facility in Korea. It offers various types of irradiation tests of nuclear fuels and materials. With the various applications of the HANARO capsule for the academic and industrial applications, new technologies and relevant facilities will become more important especially for the advanced nuclear fuels and materials development. A new I and C system for an irradiation test using an instrumented fuel capsule have been designed and manufactured to provide more qualified data to fuel developer. The performance test which started in 2004, was done to investigate the thermal response of the capsule connected to the gas mixing system of the new I and C system(GSF-2002) in the cold test loop under the HANARO hydraulic operational condition. Main test parameters are mass flow rate of 25, 50 and 100 cc/min of He/Ne gas, gas pressure of 1 to 3 kg/cm{sup 2}, heater power of 1 to 3.4kW and different gas mixing ratios of He to Ne. From the out-pile tests, it was confirmed that the I and C system(GSF-2002) would be feasible for the fuel irradiation tests. Both analytical and test data prepared by this study are directly used for the fuel experiments related to advanced fuel development program.

Simulation of integral local tests with high-burnup fuel

International Nuclear Information System (INIS)

Gyori, G.

2011-01-01

The behaviour of nuclear fuel under LOCA conditions may strongly depend on the burnup-dependent fuel characteristics, as it has been indicated by recent integral experiments. Fuel fragmentation and the associated fission gas release can influence the integral fuel behaviour, the rod rupture and the radiological release. The TRANSURANUS fuel performance code is a proper tool for the consistent simulation of burnup-dependent phenomena during normal operation and the thermo-mechanical behaviour of the fuel rod in a subsequent accident. The code has been extended with an empirical model for micro-cracking induced FGR and fuel fragmentation and verified against integral LOCA tests of international projects. (author)

Corrosion Tests of LWR Fuels - Nuclide Release

International Nuclear Information System (INIS)

P.A. Finn; Y. Tsai; J.C. Cunnane

2001-01-01

Two BWR fuels [64 and 71 (MWd)/kgU], one of which contained 2% Gd, and two PWR fuels [30 and 45 (MWd)/kgU], are tested by dripping groundwater on the fuels under oxidizing and hydrologically unsaturated conditions for times ranging from 2.4 to 8.2 yr at 90 C. The 99 Tc, 129 I, 137 Cs, 97 Mo, and 90 Sr releases are presented to show the effects of long reaction times and of gadolinium on nuclide release. This investigation showed that the five nuclides at long reaction times have similar fractional release rates and that the presence of 2% Gd reduced the 99 Tc cumulative release fraction by about an order of magnitude over that of a fuel with a similar burnup

Methodology of thermalhydraulic tests of fuel assemblies for WWER-1000

International Nuclear Information System (INIS)

Archipov, A.; Kolochko, V.N.

2001-01-01

At present 11 units with WWER-1000 are in operation in Ukraine. The NPPs are provided with nuclear fuel from Russia. The fuel assemblies are fabricated and delivered to Ukrainian NPPs from Russia. However the contemporary tendencies of nuclear energy development in the world assume a diversification of nuclear fuel vendors. Therefore the creation of the own nuclear fuel cycle of Ukraine is in mind in the strategy of nuclear energy development of Ukraine. As a part of the fuel assemblies fabrication process complex of the thermalhydraulic tests should be carried out to confirm design characteristics of the fuel assemblies before they are loaded in the reactor facility. The experimental basis and scientific infrastructure for the thermalhydraulic tests arrangement and realization of the programs and procedures for the core equipment examination are under consideration. (author)

Power Burst Facility severe-fuel-damage test program

International Nuclear Information System (INIS)

McCardell, R.K.; MacDonald, P.E.

1982-01-01

As a result of the Three Mile Island Unit 2 (TMI-2) accident, the United States Nuclear Regulatory Commission (USNRC) has initiated a severe fuel damage research program to investigate fuel rod and core response, and fission product and hydrogen release and transport during degraded core cooling accidents. This paper presents a discussion of the expected benefits of the PBF severe fuel damage tests to the nuclear industry, a description of the first five planned experiments, the results of pretest analysis performed to predict the fuel bundle heatup for the first two experiments, and a discussion of Phase II severe fuel damage experiments. Modifications to the fission product detection system envisioned for the later experiments are also described

Experimental data report for test TS-3 Reactivity Initiated Accident test in the NSRR with pre-irradiated BWR fuel rod

International Nuclear Information System (INIS)

Nakamura, Takehiko; Yoshinaga, Makio; Fujishiro, Toshio; Kobayashi, Shinsho; Yamahara, Takeshi; Sukegawa, Tomohide; Kikuchi, Teruo; Sobajima, Makoto.

1993-09-01

This report presents experimental data for Test TS-3 which was the third test in a series of Reactivity Initiated Accident (RIA) tests using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in September, 1990. Test fuel rod used in the Test TS-3 was a short-sized BWR (7 x 7) type rod which was re-fabricated from a commercial rod irradiated in the Tsuruga Unit 1 power reactor of Japan Atomic Power Co. The fuel had an initial enrichment of 2.79 % and a burnup of 26 Gwd/tU. A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The energy deposition of the fuel rod in this test was evaluated to be 94 ± 4 cal/g · fuel (88 ± 4 cal/g · fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and results of pre-pulse and post-pulse irradiation examinations are described in this report. (author)

The US Army Foreign Comparative Test fuel cell program

Science.gov (United States)

Bostic, Elizabeth; Sifer, Nicholas; Bolton, Christopher; Ritter, Uli; Dubois, Terry

The US Army RDECOM initiated a Foreign Comparative Test (FCT) Program to acquire lightweight, high-energy dense fuel cell systems from across the globe for evaluation as portable power sources in military applications. Five foreign companies, including NovArs, Smart Fuel Cell, Intelligent Energy, Ballard Power Systems, and Hydrogenics, Inc., were awarded competitive contracts under the RDECOM effort. This paper will report on the status of the program as well as the experimental results obtained from one of the units. The US Army has interests in evaluating and deploying a variety of fuel cell systems, where these systems show added value when compared to current power sources in use. For low-power applications, fuel cells utilizing high-energy dense fuels offer significant weight savings over current battery technologies. This helps reduce the load a solider must carry for longer missions. For high-power applications, the low operating signatures (acoustic and thermal) of fuel cell systems make them ideal power generators in stealth operations. Recent testing has been completed on the Smart Fuel Cell A25 system that was procured through the FCT program. The "A-25" is a direct methanol fuel cell hybrid and was evaluated as a potential candidate for soldier and sensor power applications.

Conceptual development of a test facility for spent fuel management

Energy Technology Data Exchange (ETDEWEB)

Park, S.W.; Lee, H.H.; Lee, J.Y.; Lee, J.S.; Ro, S.G. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

Spent fuel management is an important issue for nuclear power program, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts are directed at KAERI to develop advanced technologies for safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility of pilot scale is being developed with provisions for integral demonstration of a multitude of technical functions required for spent fuel management. The facility, baptized SMART (Spent fuel MAnagement technology Research and Test facility), is to be capable of handling full size assembly of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 10 MTU/y (about 24 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half-volume package and optionally transformation of the fuel rod into a fuel of CANDU type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer-term dry storage or direct disposal ) in the former case and direct refabrication of the spent PWR fuel into CANDU-type DUPIC fuel for reuse in CANDU reactors in the latter case, respectively. In addition to these major functions, there are other associated technologies to be demonstrated : such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive materials is configured in such way to maximize costs within the given functional and operational requirements. (author). 3 tabs., 4 figs.

Conceptual development of a test facility for spent fuel management

International Nuclear Information System (INIS)

Park, S.W.; Lee, H.H.; Lee, J.Y.; Lee, J.S.; Ro, S.G.

1997-01-01

Spent fuel management is an important issue for nuclear power program, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts are directed at KAERI to develop advanced technologies for safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility of pilot scale is being developed with provisions for integral demonstration of a multitude of technical functions required for spent fuel management. The facility, baptized SMART (Spent fuel MAnagement technology Research and Test facility), is to be capable of handling full size assembly of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 10 MTU/y (about 24 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half-volume package and optionally transformation of the fuel rod into a fuel of CANDU type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer-term dry storage or direct disposal ) in the former case and direct refabrication of the spent PWR fuel into CANDU-type DUPIC fuel for reuse in CANDU reactors in the latter case, respectively. In addition to these major functions, there are other associated technologies to be demonstrated : such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive materials is configured in such way to maximize costs within the given functional and operational requirements. (author). 3 tabs., 4 figs

Irradiation testing of high-density uranium alloy dispersion fuels

International Nuclear Information System (INIS)

Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

1997-01-01

Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 'microplates'. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U10Mo-0.05Sn, U2Mo, or U 3 Si 2 . These experiments will be discharged at peak fuel burnups of approximately 40 and 80 at.% U 235 . Of particular interest are the extent of reaction of the fuel and matrix phases and the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions. (author)

Prototype spent-fuel canister design, analysis, and test

International Nuclear Information System (INIS)

Leisher, W.B.; Eakes, R.G.; Duffey, T.A.

1982-03-01

Sandia National Laboratories was asked by the US Energy Research and Development Administration (now US Department of Energy) to design the spent fuel shipping cask system for the Clinch River Breeder Reactor Plant (CRBRP). As a part of this task, a canister which holds liquid sodium and the spent fuel assembly was designed, analyzed, and tested. The canister body survived the regulatory Type-B 9.1-m (30-ft) drop test with no apparent leakage. However, the commercially available metal seal used in this design leaked after the tests. This report describes the design approach, analysis, and prototype canister testing. Recommended work for completing the design, when funding is available, is included

Fabrication and testing of uranium nitride fuel for space power reactors

Science.gov (United States)

Matthews, R. B.; Chidester, K. M.; Hoth, C. W.; Mason, R. E.; Petty, R. L.

1988-02-01

Uranium nitride fuel was selected for previous space power reactors because of its attractive thermal and physical properties; however, all UN fabrication and testing activities were terminated over ten years ago. An accelerated irradiation test, SP-1, was designed to demonstrate the irradiation performance of Nb-1 Zr clad UN fuel pins for the SP-100 program. A carbothermic-reduction/nitriding process was developed to synthesize UN powders. These powders were fabricated into fuel pellets by conventional cold-pressing and sintering. The pellets were loaded into Nb-1 Zr cladding tubes, irradiated in a fast-test reactor, and destructively examined after 0.8 at% burnup. Preliminary postirradiation examination (PIE) results show that the fuel pins behaved as designed. Fuel swelling, fission-gas release, and microstructural data are presented, and suggestions to enhance the reliability of UN fuel pins are discussed.

Comprehensive exergetic and economic comparison of PWR and hybrid fossil fuel-PWR power plants

International Nuclear Information System (INIS)

Sayyaadi, Hoseyn; Sabzaligol, Tooraj

2010-01-01

A typical 1000 MW Pressurized Water Reactor (PWR) nuclear power plant and two similar hybrid 1000 MW PWR plants operate with natural gas and coal fired fossil fuel superheater-economizers (Hybrid PWR-Fossil fuel plants) are compared exergetically and economically. Comparison is performed based on energetic and economic features of three systems. In order to compare system at their optimum operating point, three workable base case systems including the conventional PWR, and gas and coal fired hybrid PWR-Fossil fuel power plants considered and optimized in exergetic and exergoeconomic optimization scenarios, separately. The thermodynamic modeling of three systems is performed based on energy and exergy analyses, while an economic model is developed according to the exergoeconomic analysis and Total Revenue Requirement (TRR) method. The objective functions based on exergetic and exergoeconomic analyses are developed. The exergetic and exergoeconomic optimizations are performed using the Genetic Algorithm (GA). Energetic and economic features of exergetic and exergoeconomic optimized conventional PWR and gas and coal fired Hybrid PWR-Fossil fuel power plants are compared and discussed comprehensively.

Fuel Subsystems Flight Test Handbook

Science.gov (United States)

1981-12-01

detailed, accessible .-ltand complete test records for his own protection and for the benefit of his successor in case of promotion, transfer or...and pilot display of fuel quantity, low level warning and a " Bingo " fudl warning. 3.0 TEST OBJECT1VES: ACTION Orriclc. On POUTION•’PHONWE [ DATE PCR...TIS No. 46, paragraph 3.11 3.6 To demonstrate that the low level and bingo warning system are consistent L •and meet the requirements of paragraph

Reactor transients tests for SNR fuel elements in HFR reactor

International Nuclear Information System (INIS)

Plitz, H.

1989-01-01

In HFR reactor, fuel pins of LMFBR reactors are putted in irradiation specimen capsules cooled with sodium for reactor transients tests. These irradiation capsules are instrumented and the experiences realized until this day give results on: - Fuel pins subjected at a continual variation of power - melting fuel - axial differential elongation of fuel pins

Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

Energy Technology Data Exchange (ETDEWEB)

Purcell, P.C. [BNFL International Transport, Spent Fuel Services (United Kingdom); Dallongeville, M. [COGEMA Logistics (AREVA Group) (France)

2004-07-01

For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme.

Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

International Nuclear Information System (INIS)

Purcell, P.C.; Dallongeville, M.

2004-01-01

For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme

Power technology complex for production of motor fuel from brown coals with power supply from NPPs

International Nuclear Information System (INIS)

Troyanov, M.F.; Poplavskij, V.M.; Sidorov, G.I.; Bondarenko, A.V.; Chebeskov, A.N.; Chushkin, V.N.; Karabash, A.A.; Krichko, A.A.; Maloletnev, A.S.

1998-01-01

With the present-day challenge of efficient use of low-grade coals and current restructuring of coal industry in the Russian Federation, it is urgent to organise the motor fuel production by the synthesis from low grade coals and heavy petroleum residues. With this objective in view, the Institute of Physics and Power Engineering of RF Minatom and Combustible Resources Institute of RF Mintopenergo proposed a project of a standard nuclear power technology complex for synthetic liquid fuel (SLF) production using fast neutron reactors for power supply. The proposed project has two main objectives: (1) Engineering and economical optimization of the nuclear power supply for SLF production; and (2) Engineering and economical optimization of the SLF production by hydrogenisation of brown coals and heavy petroleum residues with a complex development of advanced coal chemistry. As a first approach, a scheme is proposed with the use of existing reactor cooling equipment, in particular, steam generators of BN-600, limiting the effect on safety of reactor facility operation at minimum in case of deviations and abnormalities in the operation of technological complex. The possibility to exclude additional requirements to the equipment for nuclear facility cooling was also taken into account. It was proposed to use an intermediate steam-water circuit between the secondary circuit sodium and the coolant to heat the technological equipment. The only change required for the BN-600 equipment will be the replacement of sections of intermediate steam superheaters at the section of main steam superheaters. The economic aspects of synthetic motor fuel production proposed by the joint project depend on the evaluation of integral balances: thermal power engineering, chemical technology, the development of advanced large scale coal chemistry of high profitability; utilisation of ash and precious microelements in waste-free technology; production of valuable isotopes; radical solution of

LOCA scenario tests of irradiated fuel rod specimens

International Nuclear Information System (INIS)

Scott, Harold

2004-01-01

Full text: The NRC's cladding performance program at Argonne National Laboratory (ANL) is testing fueled high-burnup segments subjected to LOCA integral phenomena. The data are provided to NRC and the nuclear industry for their independent assessment of the adequacy of licensing criteria for LOCA events. The tests are being conducted with high-burnup 30 cm segments from Limerick (9x9 Zry-2) and H.B. Robinson (15x15 Zry-4) reactors. Prior to testing, sibling samples are characterized with respect to fuel morphology, fuel-cladding bond, cladding oxide layer thickness, hydrogen content and high-temperature steam oxidation kinetics. Specimens that survive quench are subjected to four-point bend tests, followed by local diametral compression tests. The retention of post-quench ductility is a more limiting requirement than surviving thermal stresses during quench. Companion tests are conducted with unirradiated cladding to generate baseline data for comparison with the high-burnup fuel results. LOCA integral tests have the following sequential steps: stabilization of temperature, internal pressure and steam flow at 300 C, ramping of temperature (∼5C/s) through ballooning and burst to 1204 C, hold at 1204 C for 1-5 minutes, slow-cooling (∼3C/s) to 800 C, and water quenching at ∼800C. Two high-burnup tests were completed in 2002 with Limerick BWR rod segments: ramp to burst in argon followed by slow cooling; and the LOCA test with 5-minute hold time at 1204 C, followed by slow cooling. With the exception of burst-opening shape, results for burst temperature, burst pressure, burst length, and ballooning strain profile are more similar to, than different from, results for unirradiated Zry-2 cladding exposed to the same time-temperature history. The 3rd Limerick test with quench was performed in December 2003, and a 4th Limerick test was performed in March 2004. Tests on high-burnup Robinson PWR fuel segments are scheduled to begin in June 2004. The presentation points

Tomographic imaging of severely disrupted fuel assemblies tested in TREAT

International Nuclear Information System (INIS)

Morman, J.A.; Froehle, P.H.; Holland, J.W.; Bennett, J.D.

1990-01-01

A series of CT codes is under development in the Reactor Analysis and Safety Division of Argonne National Laboratory for use as a post-test examination tool to analyze segments of the final fuel-bundle configuration of TREAT tests. This paper presents the results of CT analysis for fuel assemblies using neutron radiography. Fuel relocation following overpower transients in the TREAT reactor is examined for sections of the assemblies, and results are compared to metallographic sections. Further improvements are expected to increase the use and reliability of CT analysis as a standard post-test examination tool

Spent fuel test-climax: technical measurements interim report, FY 1980

International Nuclear Information System (INIS)

Carlson, R.C.; Patrick, W.C.; Wilder, D.G.; Brough, W.G.; Montan, D.N.; Harben, P.E.; Ballou, L.B.; Heard, H.C.

1980-01-01

The Spent Fuel Test--Climax (SFT-C), a test of the retrievable geologic storage of spent fuel assemblies from an operating commercial power reactor, is under way at the Nevada Test Site of the US Department of Energy. Although the main thrust of the project is a demonstration of the feasibility of packaging, handling, storing, and retrieving the highly radioactive fuel assemblies, over 800 data channels have been installed to monitor the response of the rock to the heat and radiation produced by the fuel assemblies and to distinguish in that response the effect due to heat alone. Temperatures in the test array are tracking well with thermal modeling calculations performed before the test was started. The fuel assemblies have been in place since May 1980. The canisters have passed through skin temperature maxima of about 145 0 C and are currently declining in temperature. Evidence is emerging that the thermomechanical response of the rock surrounding the SFT-C is strongly affected by fractures and other discontinuities inthe rock. Most of the effort to date has been in project construction, design, and installation of the instrumentation. Although the data are available in raw form for verification purposes, the data are not as yet in a suitable form for detailed analyses. Work continues on the data management aspects of the project and in continued monitoring of the test

Evaluation of the linear power of HANARO test fuel bundles

Energy Technology Data Exchange (ETDEWEB)

Lee, Choong Sung; Seo, C. G.; Lee, B. C.; Kim, H. R

2001-02-01

The HANARO fuel was developed by AECL and it is configured in a bundle of rods containing uranium silicide. AECL has conducted a variety of tests using specimen in order to achieve its qualification and licensing and the highest linear power was evaluated to be 112.8kW/m. In design stage of HANARO, the best estimated maximum linear power at hot spot was found to occur in the transition core from the initial to the equilibrium and its value was 108kW/m, which exceeds 112.8kW/m if the physics uncertainty of the HANARO nuclear design model is taken into account. Consequently, the licensing body issued the conditional permit to operate HANARO and the fuel integrity at the linear power higher than 112.8kW/m was requested to be confirmed through irradiation tests by realizing its repeatability. Hereby, KAERI designed uninstrumented and instrumented test fuel bundles and conducted their burnup tests. In parallel with the tests, the nuclear design model has been revised and updated to enable us to pursue the pin-by-pin power history. This report describes the best estimated power history of the test fuel bundles using the revised model. In conclusion, HANARO fuel keeps its integrity at power condition greater than 120kW/m.

16 x 16 Vantage+ Fuel Assembly Flow Vibrational Testing

International Nuclear Information System (INIS)

Chambers, Martin; Kurincic, Bojan

2014-01-01

Nuklearna Elektrarna Krsko (NEK) has experienced leaking fuel after increasing the cycle duration to 18 months. The leaking fuel mechanism has predominantly been consistent over multiple cycles and is typically observed in highly irradiated Fuel Assemblies (FA) after around 4 years of continuous operation that were located at the core periphery (baffle). The cause of the leaking fuel is due to Grid-To-Rod-Fretting (GRTF) and occasional debris fretting. NEK utilises a 16x16 Vantage+ FA design with all Inconel structural mixing vane grids (8 in total), Zirlo thimbles, Integral Fuel Burnable Absorber (IFBA) rods with enriched ZrB2, enriched Annular Blanket, Debris Filter Bottom Nozzle (DFBN), Removable Top Nozzle (RTN) and Zirlo fuel cladding material with a high burnup capability of 60 GWD/MTU. Numerous design and operational changes are thought to have reduced the original 16x16 FA design margin to fretting resistance of either vibration or its wear work rate, such as significant power uprate (spring force loss, rod creep down...), operational cycle duration increase from 12 to 18 months (increasing residence time as well as lead FA and fuel rod burnup values), Reactor Coolant System flow increase (increased vibration), removal of Thimble Plugs (increased bypass flow, increased vibration) and Zirc-4 to Zirlo cladding change (decreasing wear work rate). The fuel rod to grid spring as well as dimple contact areas are relatively smaller than other FA designs that exhibit good in-reactor fretting performance. A FA design change project to address the small rod to dimple / spring contact area and utilise fuel cladding oxide coating is currently being pursued with the fuel supplier. The FA vibrational properties are very important to the in-reactor FA performance and reliability. The 16x16 Vantage+ vibrational testing was performed with a full size FA in the Fuel Assembly Compatibility Testing (FACTS) loop that is able to provide full flow rates at elevated temperature

Corrosion Investigations in Straw-Fired Power Plants in Denmark

DEFF Research Database (Denmark)

Montgomery, Melanie; Frandsen, Flemming; Karlsson, A

2001-01-01

of accelerated corrosion. The type of corrosion attack can be directly ascribed to the composition of the deposit and the metal surface temperature. A series of field tests have been undertaken in the various straw-fired power plants in Denmark, namely the Masnedø, Rudkøbing and Ensted CHP plants. Three types......In Denmark, straw and other types of biomass are used for generating energy in power plants. Straw has the advantage that it is a "carbon dioxide neutral fuel" and therefore environmentally acceptable. Straw combustion is associated with corrosion problems which are not encountered in coal-fired...... of exposure were undertaken to investigate corrosion: a) the exposure of metal rings on water/air cooled probes, b) the exposure of test tubes in a test superheater, and c) the exposure of test tubes in existing superheaters. Thus both austenitic steels and ferritic steels were exposed in the steam...

Results of a conventional fuel tank blunt impact test

Science.gov (United States)

2015-03-23

The Federal Railroad Administrations Office of Research : and Development is conducting research into passenger : locomotive fuel tank crashworthiness. A series of impact tests is : being conducted to measure fuel tank deformation under two : type...

Testing of FFTF fuel handling equipment

International Nuclear Information System (INIS)

Coleman, D.W.; Grazzini, E.D.; Hill, L.F.

1977-07-01

The Fast Flux Test Facility has several manual/computer controlled fuel handling machines which are exposed to severe environments during plant operation but still must operate reliably when called upon for reactor refueling. The test programs for two such machines--the Closed Loop Ex-Vessel Machine and the In-Vessel Handling Machine--are described. The discussion centers on those areas where design corrections or equipment repairs substantiated the benefits of a test program prior to plant operation

Back pressure helium leak testing of fuel elements for Dhruva research reactor

Energy Technology Data Exchange (ETDEWEB)

Dutta, N G; Ahmad, Anis; Kulkarni, P G; Purushotham, D S.C. [Bhabha Atomic Research Centre, Bombay (India). Atomic Fuels Div.

1994-12-31

Leak tightness specification on fuel elements for reactor use is always very stringent. The fuel element fabricated for Dhruva reactor is specified to be leak-tight up to 1 x 10{sup -8} std. cc/sec. The fuel element consists of natural metallic uranium rod around 12.5 mm diameter and 3 meter long in encased in aluminium tube and seal welded at both ends. Since helium gas is not filled inside the fuel element while doing seal welding, the only way to do helium leak testing of such fuel rods is by back-pressure technique. This paper describes the development of test facility for carrying out such test and discusses the experiences of carrying out helium leak testing by back-pressure technique on more than 700 numbers of fuel rods for Dhruva reactor. (author). 4 refs., 3 figs., 1 tab.

Development of Start-up and Shutdown Procedure for the HANARO Fuel Test Loop

International Nuclear Information System (INIS)

Park, S. K.; Sim, B. S.; Chi, D. Y.; Lee, J. M.; Lee, C. Y.; Ahn, S. H.

2009-06-01

A start-up and shutdown procedure for the HANARO fuel test loop has been developed. This is a facility for fuel and material irradiation tests. The facility provides experimental conditions similar to the normal operational pressures and temperatures of commercial PWR and CANDU plants. The normal operation modes of the HANARO fuel test loop are classified into loop shutdown, cold stand-by 1, cold stand-by 2, hot stand-by, and hot operation. The operation modes depend on the fission power of test fuels and the coolant temperature at the inlet of the in-pile test section. The HANARO must maintain a shutdown mode if the HANARO fuel test loop is loop shutdown, cold stand-by 1, cold stand-by 2, or hot stand-by. As the HANARO becomes power operation mode, the operation mode of the HANARO fuel test loop comes to hot operation from hot stand-by. The procedure for the HANARO fuel test loop consists of four main parts such as check of initial conditions, stat-up operation procedure, shutdown operation procedure, and check lists for operations. Several hot test operations ensure that the procedure is appropriate

40 CFR 1060.521 - How do I test fuel caps for permeation emissions?

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 32 2010-07-01 2010-07-01 false How do I test fuel caps for permeation... EQUIPMENT Test Procedures § 1060.521 How do I test fuel caps for permeation emissions? If you measure a fuel tank's permeation emissions with a nonpermeable covering in place of the fuel cap under § 1060.520(b)(5...

CFD analysis of temperature imbalance in superheater/reheater region of tangentially coal-fired boiler

Science.gov (United States)

Zainudin, A. F.; Hasini, H.; Fadhil, S. S. A.

2017-10-01

This paper presents a CFD analysis of the flow, velocity and temperature distribution in a 700 MW tangentially coal-fired boiler operating in Malaysia. The main objective of the analysis is to gain insights on the occurrences in the boiler so as to understand the inherent steam temperature imbalance problem. The results show that the root cause of the problem comes from the residual swirl in the horizontal pass. The deflection of the residual swirl due to the sudden reduction and expansion of the flow cross-sectional area causes velocity deviation between the left and right side of the boiler. This consequently results in flue gas temperature imbalance which has often caused tube leaks in the superheater/reheater region. Therefore, eliminating the residual swirl or restraining it from being diverted might help to alleviate the problem.

Testing of HTR UO{sub 2} TRISO fuels in AVR and in material test reactors

Energy Technology Data Exchange (ETDEWEB)

Kania, Michael J., E-mail: MichaelJKania@googlemail.com [Retired from Lockheed Martin Corp, 20 Beach Road, Averill Park, NY 12018 (United States); Nabielek, Heinz, E-mail: heinznabielek@me.com [Retired from Research Center Jülich, Monschauerstrasse 61, 52355 Düren (Germany); Verfondern, Karl [Research Center Juelich,Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Allelein, Hans-Josef [Research Center Juelich,Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); RWTH Aachen, 52072 Aachen (Germany)

2013-10-15

The German High Temperature Reactor Fuel Development Program successfully developed, licensed and manufactured many thousands of spherical fuel elements that were used to power the experimental AVR reactor and the commercial THTR reactor. In the 1970s, this program extended the performance envelope of HTR fuels by developing and qualifying the TRISO-coated particle system. Irradiation testing in real-time AVR tests and accelerated MTR tests demonstrated the superior manufacturing process of this fuel and its irradiation performance. In the 1980s, another program direction change was made to a low enriched UO{sub 2} TRISO-coated particle system coupled with high-quality manufacturing specifications designed to meet new HTR plant design needs. These needs included requirements for inherent safety under normal operation and accident conditions. Again, the German fuel development program met and exceeded these challenges by manufacturing and qualifying the low-enriched UO{sub 2} TRISO-fuel system for HTR systems with steam generation, gas-turbine systems and very high temperature process heat applications. Fuel elements were manufactured in production scale facilities that contained near defect free UO{sub 2} TRISO coated particles, homogeneously distributed within a graphite matrix with very low levels of uranium contamination. Good irradiation performance for these elements was demonstrated under normal operating conditions to 12% FIMA and under accident conditions not exceeding 1600 °C.

Design and fuel fabrication processes for the AC-3 mixed-carbide irradiation test

International Nuclear Information System (INIS)

Latimer, T.W.; Chidester, K.M.; Stratton, R.W.; Ledergerber, G.; Ingold, F.

1992-01-01

The AC-3 test was a cooperative U.S./Swiss irradiation test of 91 wire-wrapped helium-bonded U-20% Pu carbide fuel pins irradiated to 8.3 at % peak burnup in the Fast Flux Test Facility. The test consisted of 25 pins that contained spherepac fuel fabricated by the Paul Scherrer Institute (PSI) and 66 pins that contained pelletized fuel fabricated by the Los Alamos National Laboratory. Design of AC-3 by LANL and PSI was begun in 1981, the fuel pins were fabricated from 1983 to 1985, and the test was irradiated from 1986 to 1988. The principal objective of the AC-3 test was to compare the irradiation performance of mixed-carbide fuel pins that contained either pelletized or sphere-pac fuel at prototypic fluence and burnup levels for a fast breeder reactor

Radiochemical analyses of several spent fuel Approved Testing Materials

International Nuclear Information System (INIS)

Guenther, R.J.; Blahnik, D.E.; Wildung, N.J.

1994-09-01

Radiochemical characterization data are described for UO 2 and UO 2 plus 3 wt% Gd 2 O 3 commercial spent nuclear fuel taken from a series of Approved Testing Materials (ATMs). These full-length nuclear fuel rods include MLA091 of ATM-103, MKP070 of ATM-104, NBD095 and NBD131 of ATM-106, and ADN0206 of ATM-108. ATMs 103, 104, and 106 were all irradiated in the Calvert Cliffs Nuclear Power Plant (Reactor No.1), a pressurized-water reactor that used fuel fabricated by Combustion Engineering. ATM-108 was part of the same fuel bundle designed as ATM-105 and came from boiling-water reactor fuel fabricated by General Electric and irradiated in the Cooper Nuclear Power Plant. Rod average burnups and expected fission gas releases ranged from 2,400 to 3,700 GJ/kgM. (25 to 40 Mwd/kgM) and from less than 1% to greater than 10%, respectively, depending on the specific ATM. The radiochemical analyses included uranium and plutonium isotopes in the fuel, selected fission products in the fuel, fuel burnup, cesium and iodine on the inner surfaces of the cladding, 14 C in the fuel and cladding, and analyses of the gases released to the rod plenum. Supporting examinations such as fuel rod design and material descriptions, power histories, and gamma scans used for sectioning diagrams are also included. These ATMs were examined as part of the Materials Characterization Center Program conducted at Pacific Northwest Laboratory provide a source of well-characterized spent fuel for testing in support of the US Department of Energy Office of Civilian Radioactive Waste Management Program

Storage of LWR spent fuel in air: Volume 1: Design and operation of a spent fuel oxidation test facility

International Nuclear Information System (INIS)

Thornhill, C.K.; Campbell, T.K.; Thornhill, R.E.

1988-12-01

This report describes the design and operation and technical accomplishments of a spent-fuel oxidation test facility at the Pacific Northwest Laboratory. The objective of the experiments conducted in this facility was to develop a data base for determining spent-fuel dry storage temperature limits by characterizing the oxidation behavior of light-water reactor (LWR) spent fuels in air. These data are needed to support licensing of dry storage in air as an alternative to spent-fuel storage in water pools. They are to be used to develop and validate predictive models of spent-fuel behavior during dry air storage in an Independent Spent Fuel Storage Installation (ISFSI). The present licensed alternative to pool storage of spent fuel is dry storage in an inert gas environment, which is called inerted dry storage (IDS). Licensed air storage, however, would not require monitoring for maintenance of an inert-gas environment (which IDS requires) but does require the development of allowable temperature limits below which UO 2 oxidation in breached fuel rods would not become a problem. Scoping tests at PNL with nonirradiated UO 2 pellets and spent-fuel fragment specimens identified the need for a statistically designed test matrix with test temperatures bounding anticipated maximum acceptable air-storage temperatures. This facility was designed and operated to satisfy that need. 7 refs

TS-1 and TS-2 transient overpower tests on FFTF fuel

International Nuclear Information System (INIS)

Pitner, A.L.; Ferrell, P.C.; Culley, G.E.; Weber, E.T.

1985-01-01

The TS-1 and TS-2 TREAT transient experiments subjected a low burnup (2 MWd/kg) and a medium burnup (58 MWd/kg), respectively, FFTF irradiated fuel pin to unprotected 5 cents/s overpower transient conditions. The fuel pin failure response was similar in the two tests, which demonstrated a large margin to failure (P/P 0 > 3) and a favorable upper level failure location. Thus, for these transient conditions, burnup effects on transient performance appeared to be minimal in the range tested. Pin disruption in the medium burnup TS-2 test was more severe due to the higher fission gas pressurization, but failure occurred at only a 5% lower power level than for the low burnup TS-1 fuel pin. Both tests exhibited axial extrusion of molten fuel to the region above the fuel column several seconds before pin failure, demonstrating a potentially beneficial inherent safety mechanism to delay failure and mitigate accident consequences

First TREAT [Transient Reactor Test Facility] transient overpower tests on U-Pu-Zr fuel: M5 and M6

International Nuclear Information System (INIS)

Robinson, W.R.; Bauer, T.H.; Wright, A.E.; Rhodes, E.A.; Stanford, G.S.; Klickman, A.E.

1987-01-01

Transient Reactor Test Facility (TREAT) tests M5 and M6 were the first transient overpower (TOP) tests of the margin to cladding breach and prefailure elongation of metallic U-Pu-Zr ternary fuel, the reference fuel of the Integral Fast Reactor concept. Similar tests on U-Fs fueled EBR-II driver pins were previously performed and reported [1,2]. Results from these earlier tests indicated a margin to failure of about 4 times nominal power and significant axial elongation prior to failure, a feature that was very pronounced at low burnups. While these two fuel types are similar in many respects, the ternary alloy exhibits a much more complex physical structure and is typically irradiated at much higher temperatures. Thus, a prime motivation for performing M5 and M6 was to compare the safety related fuel performance characteristics of U-Fs and U-Pu-Zr. This report described conditions, results, and conclusions of testing of these fuel types

LOFT fuel rod surface temperature measurement testing

International Nuclear Information System (INIS)

Eaton, A.M.; Tolman, E.L.; Solbrig, C.W.

1978-01-01

Testing of the LOFT fuel rod cladding surface thermocouples has been performed to evaluate how accurately the LOFT thermocouples measure the cladding surface temperature during a loss-of-coolant accident (LOCA) sequence and what effect, if any, the thermocouple would have on core performance. Extensive testing has been done to characterize the thermocouple design. Thermal cycling and corrosion testing of the thermocouple weld design have provided an expected lifetime of 6000 hours when exposed to reactor coolant conditions of 620 K and 15.9 MPa and to sixteen thermal cycles with an initial temperature of 480 K and peak temperatures ranging from 870 to 1200K. Departure from nucleate boiling (DNB) tests have indicated a DNB penalty (5 to 28% lower) during steady state operation and negligible effects during LOCA blowdown caused by the LOFT fuel rod surface thermocouple arrangement. Experience with the thermocouple design in Power Burst Facility (PBF) and LOFT nonnuclear blowdown testing has been quite satisfactory. Tests discussed here were conducted using both stainless steel and zircaloy-clad electrically heated rod in the LOFT Test Support Facility (LTSF) blowdown simulation loop

PIE and separate effect test of high burnup UO2 fuel

International Nuclear Information System (INIS)

Yang, Yong Sik; Kim, S.K.; Kim, D.H.

2005-01-01

To investigate the performance of a high burnup UO 2 fuel, the highest burnup fuel assembly in KOREA was transported to the PIE facility in KAERI. It was a 17·17 fuel assembly irradiated at the Ulchin Unit 2 PWR. The peak fuel rod average burnup was about 57MWd/kgU and locally 65MWd/kgU. The general PIE was performed to investigate the fuel rod irradiation performance. Fission gas release, burnup, oxide thickness, hydrogen pickup, CRUD, and density change were measured by destructive of non-destructive test. Microstructure change, bubble and pore size distributions were observed by optical microscopy, SEM and EPMA. All generated and available PIE results were used to verify high burnup fuel performance code INFRA. Several rods were cut for additional separate effect test. For the high burnup fission gas release behaviour analysis, annealing apparatus were developed and installed in hot cell and preliminary test was performed. In addition to current apparatus new induction furnace will be installed in hot cell to investigate the high temperature and transient fission gas release behaviour. Ring tensile test was performed to analyze the material property degradation which caused by the oxidation and hydride, and additional mechanical tests will be performed. (Author)

Irradiation testing of high density uranium alloy dispersion fuels

International Nuclear Information System (INIS)

Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

1997-10-01

Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 microplates. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U-10Mo-0.05Sn, U 2 Mo, or U 3 Si 2 . These experiments will be discharged at peak fuel burnups of 40% and 80%. Of particular interest is the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions

Corrosion testing of uranium silicide fuel specimens

International Nuclear Information System (INIS)

Bourns, W.T.

1968-09-01

U 3 Si is the most promising high density natural uranium fuel for water-cooled power reactors. Power reactors fuelled with this material are expected to produce cheaper electricity than those fuelled with uranium dioxide. Corrosion tests in 300 o C water preceded extensive in-reactor performance tests of fuel elements and bundles. Proper heat-treatment of U-3.9 wt% Si gives a U 3 5i specimen which corrodes at less than 2 mg/cm 2 h in 300 o C water. This is an order of magnitude lower than the maximum corrosion rate tolerable in a water-cooled reactor. U 3 Si in a defected unbonded Zircaloy-2 sheath showed only a slow uniform sheath expansion in 300 o C water. All tests were done under isothermal conditions in an out-reactor loop. (author)

Testing system for a fuel cells stack

International Nuclear Information System (INIS)

Culcer, Mihai; Iliescu, Mariana; Stefanescu, Ioan; Raceanu, Mircea; Enache, Adrian; Lazar, Roxana Elena

2006-01-01

Hydrogen and electricity together represent one of the most promising ways to realize sustainable energy, whilst fuel cells provide the most efficient conversion devices for converting hydrogen and possibly other fuels into electricity. Thus, the development of fuel cell technology is currently being actively pursued worldwide. Due to its simple operation and other fair characteristics, the Proton Exchange Membrane Fuel Cell (PEMFC) is especially suitable as a replacement for the internal combustion engine. The PEMFC is also being developed for decentralized electricity and heat generation in buildings and mobile applications. Starting with 2001 the Institute of Research - Development for Cryogenics and Isotopic Technologies - ICIT - Rm. Valcea developed research activities supported by the Romanian Ministry of Education and Research within the National Research Program in order to bridge the gap to European competencies in the area of hydrogen and fuel cells. The paper deals with the testing system designed and developed in ICIT Rm. Valcea as a flexible and versatile tool allowing a large scale of parameter settings and measurements on a single cell or on a fuel cells stack onto a wind range of output power values. (authors)

On-Line Fuel Failure Monitor for Fuel Testing and Monitoring of Gas Cooled Very High Temperature Reactors

International Nuclear Information System (INIS)

Hawari, Ayman I.; Bourham, Mohamed A.

2010-01-01

Very High Temperature Reactors (VHTR) utilize the TRISO microsphere as the fundamental fuel unit in the core. The TRISO microsphere (∼ 1-mm diameter) is composed of a UO2 kernel surrounded by a porous pyrolytic graphite buffer, an inner pyrolytic graphite layer, a silicon carbide (SiC) coating, and an outer pyrolytic graphite layer. The U-235 enrichment of the fuel is expected to range from 4%-10% (higher enrichments are also being considered). The layer/coating system that surrounds the UO2 kernel acts as the containment and main barrier against the environmental release of radioactivity. To understand better the behavior of this fuel under in-core conditions (e.g., high temperature, intense fast neutron flux, etc.), the US Department of Energy (DOE) is launching a fuel testing program that will take place at the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL). During this project North Carolina State University (NCSU) researchers will collaborate with INL staff for establishing an optimized system for fuel monitoring for the ATR tests. In addition, it is expected that the developed system and methods will be of general use for fuel failure monitoring in gas cooled VHTRs.

PETER loop. Multifunctional test facility for thermal hydraulic investigations of PWR fuel elements

International Nuclear Information System (INIS)

Ganzmann, I.; Hille, D.; Staude, U.

2009-01-01

The reliable fuel element behavior during the complete fuel cycle is one of the fundamental prerequisites of a safe and efficient nuclear power plant operation. The fuel element behavior with respect to pressure drop and vibration impact cannot be simulated by means of fluid-structure interaction codes. Therefore it is necessary to perform tests using fuel element mock-ups (1:1). AREVA NP has constructed the test facility PETER (PWR fuel element tests in Erlangen) loop. The modular construction allows maximum flexibility for any type of fuel elements. Modern measuring instrumentation for flow, pressure and vibration characterization allows the analysis of cause and consequences of thermal hydraulic phenomena. PETER loop is the standard test facility for the qualification of dynamic fuel element behavior in flowing fluid and is used for failure mode analysis.

Fission product behavior during the first two PBF severe fuel damage tests

International Nuclear Information System (INIS)

Osetek, D.J.; Cronenberg, A.W.; Hobbins, R.R.; Vinjamuri, K.

1984-01-01

The results of the first two severe fuel damage tests performed in the Power Burst Facility are assessed in terms of fission product release and chemical behavior. On-line gamma spectroscopy and grab sample data indicate limited release during solid-phase fuel heatup. Analysis indicates that the fuel morphology conditions for the trace-irradiated fuel employed in these two tests limit initial release. Only upon high temperature fuel restructuring and liquefaction is significant release indicated. Chemical equilibrium predictions, based on steam oxidation or reduction conditions, indicate I to be the primary iodine species during trnsport in the steam environment of the first test and CsI to be the primary species during transport in the hydrogen environment of the second test. However, the higher steam flow rate conditions of the first test transported the released iodine through the sample system; whereas, low-hydrogen flow rate of the second test apparently allowed the vast majority of iodine-bearing compounds to plateout during transport

Power ramping test in the JMTR for PCI study of water reactor fuel

International Nuclear Information System (INIS)

Nakata, H.; Kanbara, M.; Ichikawa, M.

1984-01-01

Power ramping test is essential for PCI study of water reactor fuel. Boiling water capsules have been used for the tests in the JMTR. Heat generation of fuel rod in the capsule can be changed by the He-3 power control facility during reactor operation. Four specially designed fuel rods have been ramped to about 41-43 kW/m; two of them have small gaps filled with iodine, the other two are equipped with centerline temperature thermocouple. Fuel rod elongation detector is equipped to each capsule. For the fuel rods with small gap, unique contraction followed by ordinary fuel relaxation behaviour was observed right after the fast ramping. None of them failed. Future programme includes a series of tests of fuel rods irradiated in the high-pressure water loop at the JMTR and a verification test of remedy fuel which allows daily-load-following operation of BWRs. (author)

Potential high temperature corrosion problems due to co-firing of biomass and fossil fuels

DEFF Research Database (Denmark)

Montgomery, Melanie; Vilhelmsen, T.; Jensen, S.A.

2007-01-01

Over the past years, considerable high temperature corrosion problems have been encountered when firing biomass in power plants due to the high content of potassium chloride in the deposits. Therefore to combat chloride corrosion problems co-firing of biomass with a fossil fuel has been undertaken....... This results in potassium chloride being converted to potassium sulphate in the combustion chamber and it is sulphate rich deposits that are deposited on the vulnerable metallic surfaces such as high temperature superheaters. Although this removes the problem of chloride corrosion, other corrosion mechanisms...... appear such as sulphidation and hot corrosion due to sulphate deposits. At Studstrup power plant Unit 4, based on trials with exposure times of 3000 hours using 0-20% straw co-firing with coal, the plant now runs with a fuel of 10% straw + coal. After three years exposure in this environment...

Fuel staging tests at the Kymijaervi power plant

International Nuclear Information System (INIS)

Kivelae, M.; Rotter, H.; Virkki, J.

1990-01-01

The aim of this study was to measure nitrogen oxide (NO x ) emissions and find the methods to reduce them in plants using coal and natural gas as fuel. The tests involved were made at the Kymijaervi Power Plant, Lahti, Finland. Coal and natural gas was used alone or mixed. With natural gas when using flue gas recirculation, the NO x emission level dropped from 330 mg/m 3 down to 60 mg/m 3 . A negative side effect was that the flue gas temperature increased. At coal combustion and staged combustion, the flue gas recirculation had no significant effect on the NO x emission level. At coal combustion, the staging of combustion air halved the NO x emission but the combustibles increased strongly. With fuel staging, using coal as main fuel and gas as staging fuel, the NO x emission level was decreased from 340 mg/m 3 to 170 mg/m 3 . At the same time the combustibles increased 2 %- units. Also the flue gas temperature increased a little. At the tests, the proportion of natural gas was rather high, one third of the fuel energy input, but it could not be decreased, because the gas flow ratio was already too low to ensure good mixing

Development and testing of the EDF-2 reactor fuel element

International Nuclear Information System (INIS)

Delpeyroux, P.

1964-01-01

This technical report reviews the work which has been necessary for defining the EDF-2 fuel element. After giving briefly the EDF-2 reactor characteristics and the preliminary choice of parameters which made it possible to draw up a draft plan for the fuel element, the authors consider the research proper: - Uranium studies: tests on the passage into the β phase of an internal crown of a tube, bending of the tube under the effect of a localized force, welding of the end-pellets and testing for leaks. The resistance of the tube to crushing and of the pellets to yielding under the external pressure have been studied in detail in another CEA report. - Can studies: conditions of production and leak proof testing of the can, resistance of the fins to creep due to the effect of the gas flow. - Studies of the extremities of the element: creep under compression and welding of the plugs to the can. - Cartridge studies: determination of the characteristics of the can fuel fixing grooves and of the canning conditions, verification of the resistance of the fuel element to thermal cycling, determination of the temperature drop at the can-fuel interface dealt with in more detail in another CEA report. - Studies of the whole assembly: this work which concerns the graphite jacket, the support and the cartridge vibrations has been carried out by the Mechanical and Thermal Study Service (Mechanics Section). In this field the Fuel Element Study Section has investigated the behaviour of the centering devices in a gas current. The outcome of this research is the defining of the plan of the element the production process and the production specifications. The validity of ail these out-of-pile tests will be confirmed by the in-pile tests already under way and by irradiation of the elements in the EDF-2 reactor itself. In conclusion the programme is given for improving the fuel element and for defining the fuel element for the second charge. (authors) [fr

Post irradiation test report of irradiated DUPIC simulated fuel

International Nuclear Information System (INIS)

Yang, Myung Seung; Jung, I. H.; Moon, J. S. and others

2001-12-01

The post-irradiation examination of irradiated DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) simulated fuel in HANARO was performed at IMEF (Irradiated Material Examination Facility) in KAERI during 6 months from October 1999 to March 2000. The objectives of this post-irradiation test are i) the integrity of the capsule to be used for DUPIC fuel, ii) ensuring the irradiation requirements of DUPIC fuel at HANARO, iii) performance verification in-core behavior at HANARO of DUPIC simulated fuel, iv) establishing and improvement the data base for DUPIC fuel performance verification codes, and v) establishing the irradiation procedure in HANARO for DUPIC fuel. The post-irradiation examination performed are γ-scanning, profilometry, density, hardness, observation the microstructure and fission product distribution by optical microscope and electron probe microanalyser (EPMA)

Technology Implementation Plan: Irradiation Testing and Qualification for Nuclear Thermal Propulsion Fuel

Energy Technology Data Exchange (ETDEWEB)

Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Rader, Jordan D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-01

This document is a notional technology implementation plan (TIP) for the development, testing, and qualification of a prototypic fuel element to support design and construction of a nuclear thermal propulsion (NTP) engine, specifically its pre-flight ground test. This TIP outlines a generic methodology for the progression from non-nuclear out-of-pile (OOP) testing through nuclear in-pile (IP) testing, at operational temperatures, flows, and specific powers, of an NTP fuel element in an existing test reactor. Subsequent post-irradiation examination (PIE) will occur in existing radiological facilities. Further, the methodology is intended to be nonspecific with respect to fuel types and irradiation or examination facilities. The goals of OOP and IP testing are to provide confidence in the operational performance of fuel system concepts and provide data to program leadership for system optimization and fuel down-selection. The test methodology, parameters, collected data, and analytical results from OOP, IP, and PIE will be documented for reference by the NTP operator and the appropriate regulatory and oversight authorities. Final full-scale integrated testing would be performed separately by the reactor operator as part of the preflight ground test.

Improving the AGR fuel testing power density profile versus irradiation-time in the advanced test reactor

International Nuclear Information System (INIS)

Chang, Gray S.; Lillo, Misti A.; Maki, John T.; Petti, David A.

2009-01-01

The Very High Temperature gas-cooled Reactor (VHTR), which is currently being developed, achieves simplification of safety through reliance on ceramic-coated fuel particles. Each TRISO-coated fuel particle has its own containment which serves as the principal barrier against radionuclide release under normal operating and accident conditions. These fuel particles, in the form of graphite fuel compacts, are currently undergoing a series of irradiation tests in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) to support the Advanced Gas-Cooled Reactor (AGR) fuel qualification program. A representive coated fuel particle with an 235 U enrichment of 19.8 wt% was used in this analysis. The fuel burnup analysis tool used to perform the neutronics study reported herein, couples the Monte Carlo transport code MCNP, with the radioactive decay and burnup code ORIGEN2. The fuel burnup methodology known as Monte-Carlo with ORIGEN2 (MCWO) was used to evaluate the AGR experiment assembly and demonstrate compliance with ATR safety requirements. For the AGR graphite fuel compacts, the MCWO-calculated fission power density (FPD) due to neutron fission in 235 U is an important design parameter. One of the more important AGR fuel testing requirements is to maintain the peak fuel compact temperature close to 1250degC throughout the proposed irradiation campaign of 550 effective full power days (EFPDs). Based on the MCWO-calculated FPD, a fixed gas gap size was designed to allow regulation of the fuel compact temperatures throughout the entire fuel irradiation campaign by filling the gap with a mixture of helium and neon gases. The chosen fixed gas gap can only regulate the peak fuel compact temperature in the desired range during the irradiation test if the ratio of the peak power density to the time-dependent low power density (P/T) at 550 EFPDs is less than 2.5. However, given the near constant neutron flux within the ATR driver core and the depletion of 235 U

Basic research on high-uranium density fuels for research and test reactors

International Nuclear Information System (INIS)

Ugajin, M.; Itoh, A.; Akabori, M.

1992-01-01

High-uranium density fuels, uranium silicides (U 3 Si 2 , U 3 Si) and U 6 Me-type uranium alloys (Me = Fe, Mn, Ni), were prepared and examined metallurgically as low-enriched uranium (LEU) fuels for research and test reactors. Miniature aluminum-dispersion plate-type fuel (miniplate) and aluminum-clad disk-type fuel specimens were fabricated and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Fuel-aluminum compatibility tests were conducted to elucidate the extent of reaction and to identify reaction products. The relative stability of the fuels in an aluminum matrix was established at 350degC or above. Experiments were also performed to predict the chemical form of the solid fission-products in the uranium silicide (U 3 Si 2 ) simulating a high burnup anticipated for reactor service. (author)

Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

Science.gov (United States)

Bradley, David E.; Mireles, Omar R.; Hickman, Robert R.

2011-01-01

Deep space missions with large payloads require high specific impulse (Isp) and relatively high thrust in order to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average Isp. Nuclear thermal rockets (NTR) capable of high Isp thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high temperature hydrogen exposure on fuel elements is limited. The primary concern is the mechanical failure of fuel elements which employ high-melting-point metals, ceramics or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via non-contact RF heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

Irradiation of MEU and LEU test fuel elements in DR 3

International Nuclear Information System (INIS)

Haack, K.

1984-01-01

Irradiation of three MEU and three LEU fuel elements in the Danish reactor DR 3. Thermal and fast neutron flux density scans of the core have been made and the results, related to the U235-content of each fuel element, are compared with the values from HEU fuel elements. The test elements were taken to burn-up percentages of 50-60%. Reactivity values of the test elements at charge and at discharge have been measured and the values are compared with those of HEU fuel elements. (author)

Degradation mechanisms and accelerated testing in PEM fuel cells

Energy Technology Data Exchange (ETDEWEB)

Borup, Rodney L [Los Alamos National Laboratory; Mukundan, Rangachary [Los Alamos National Laboratory

2010-01-01

The durability of PEM fuel cells is a major barrier to the commercialization of these systems for stationary and transportation power applications. Although there has been recent progress in improving durability, further improvements are needed to meet the commercialization targets. Past improvements have largely been made possible because of the fundamental understanding of the underlying degradation mechanisms. By investigating component and cell degradation modes; defining the fundamental degradation mechanisms of components and component interactions new materials can be designed to improve durability. Various factors have been shown to affect the useful life of PEM fuel cells. Other issues arise from component optimization. Operational conditions (such as impurities in either the fuel and oxidant stream), cell environment, temperature (including subfreezing exposure), pressure, current, voltage, etc.; or transient versus continuous operation, including start-up and shutdown procedures, represent other factors that can affect cell performance and durability. The need for Accelerated Stress Tests (ASTs) can be quickly understood given the target lives for fuel cell systems: 5000 hours ({approx} 7 months) for automotive, and 40,000 hrs ({approx} 4.6 years) for stationary systems. Thus testing methods that enable more rapid screening of individual components to determine their durability characteristics, such as off-line environmental testing, are needed for evaluating new component durability in a reasonable turn-around time. This allows proposed improvements in a component to be evaluated rapidly and independently, subsequently allowing rapid advancement in PEM fuel cell durability. These tests are also crucial to developers in order to make sure that they do not sacrifice durability while making improvements in costs (e.g. lower platinum group metal [PGM] loading) and performance (e.g. thinner membrane or a GDL with better water management properties). To

An inspection standard of fuel for the high temperature engineering test reactor

International Nuclear Information System (INIS)

Kobayashi, Fumiaki; Shiozawa, Shusaku; Sawa, Kazuhiro; Sato, Sadao; Hayashi, Kimio; Fukuda, Kosaku; Kaneko, Mitsunobu; Sato, Tsutomu.

1992-06-01

The High Temperature Engineering Test Reactor (HTTR) uses the fuel comprising coated fuel particles. A general inspection standard for the coated particle fuel, however, has not been established in Japan. Therefore, it has been necessary to prescribe the inspection standard of the fuel for HTTR. Under these circumstances, a fuel inspection standard of HTTR has been established under cooperation of fuel specialists both inside and outside of JAERI on referring to the inspection methods adopted in USA, Germany and Japan for HTGR fuels. Since a large number of coated fuel particle samples is needed to inspect the HTTR fuel, the sampling inspection standard has also been established considering the inspection efficiency. This report presents the inspection and the sampling standards together with an explanation of these standards. These standards will be applied to the HTTR fuel acceptance tests. (author)

New phenomena observed during fuel assemblies testing

International Nuclear Information System (INIS)

Tzotcheva, V.

2001-01-01

The paper presents a new attempt to explain inexplicable increase of specific activity for some of the fuel assemblies during the fuel tightness testing procedures on Kozloduy NPP. A brief description of established procedure for fuel tightness control is presented in the paper. Special emphasis is given on a hypothesis that explains the fact of existence of deviation in Iodine activity more than usual, which have no reasonable interpretation. The reasons for uniform high Iodine activity for reloaded assemblies, that have kept in the open measuring can for a long time (1-3 hours), is found to be the process of Iodine dissolving in the water and the accelerated process of natural degassing. A proposal to use the 134 Cs and 137 Cs as stand-alone criteria for more precise results is made in respect to increase the reliability of fuel reloading and storage procedures

Research and Test Reactor Fuel Elements (RTRFE)

International Nuclear Information System (INIS)

Pace, Brett W.; Marinak, Edward A.

1999-01-01

BWX Technologies Inc. (BWXT) has experienced several production improvements over the past year. The homogeneity yields in 4.8 gU/cc U 3 Si 2 plates have increased over last year's already high yields. Through teamwork and innovative manufacturing techniques, maintaining high quality surface finishes on plates and elements is becoming easier and less expensive. Currently, BWXT is designing a fabrication development plan to reach a fuel loading of 9 gU/cc within 2 - 4 years. This development will involve a step approach requested by ANL to produce plates using U-8Mo at a loading of 6 gU/cc first and qualify the fuel at those levels. In achieving the goal of a very high-density fuel loading of 9 gU/cc, BWXT is considering employing several new, state of the art, ultrasonic testing techniques for fuel core evaluation. (author)

Modelling of LOCA Tests with the BISON Fuel Performance Code