WorldWideScience

Sample records for superheater test fuel

  1. Superheater hydraulic model test plan

    Energy Technology Data Exchange (ETDEWEB)

    Gabler, M.; Oliva, R.M.

    1973-10-01

    The plan for conducting a hydraulic test on a full scale model of the AI Steam Generator Module design is presented. The model will incorporate all items necessary to simulate the hydraulic performance characteristics of the superheater but will utilize materials other than the 2-1/4 Cr - 1 Mo in its construction in order to minimize costs and expedite schedule. Testing will be performed in the Rockwell International Rocketdyne High Flow Test Facility which is capable of flowing up to 32,00 gpm of water at ambient temperatures. All necessary support instrumentation is also available at this facility.

  2. Superheater fouling in a BFB boiler firing wood-based fuel blends

    NARCIS (Netherlands)

    Stam, A.F.; Haasnoot, K.; Brem, Gerrit

    2014-01-01

    Four different fuel blends have been fired in a 28 MWel BFB. Wood pellets (test 0) were not problematic for about ten years, contrary to a mixture of demolition wood, wood cuttings, compost overflow, paper sludge and roadside grass (test 1) which caused excessive fouling at a superheater bundle

  3. Superheater corrosion in a boiler fired with refuse-derived fuel

    Energy Technology Data Exchange (ETDEWEB)

    Blough, J.L.; Stanko, G.J. [Foster Wheeler Development Corp., Livingston, NJ (United States); Bakker, W.T. [Electric Power Research Inst., Palo Alto, CA (United States); Steinbeck, T. [United Power Association, Elk River, MN (United States)

    1995-12-31

    The environment in the superheater of a boiler firing refuse-derived fuel (RDF) is very aggressive. The high wastage rate for the standard T-22 material necessitated a materials testing program. Simples of Types 304H, HR3C, T-22 chromized, 825 and 625 were assembled into tubular test sections and welded into the superheater tubing. After 1,180 hours the test sections were evaluated and the wastage rates calculated for each material. The chlorides contained in the RDF are believed to be the primary corrodent. The chlorine may be interacting with the metal samples as HCl, a low-melting-point eutectic or a combination of these. Of the six materials tested, Alloy 625 exhibited the best resistance--substantially better than the next-best Type 304. Alloy 825 and HR3C corroded approximately 1.5 times the rate of Type 304. The chromized layer on T-22 showed no resistance to the environment and was consumed in large areas.

  4. Superheater corrosion in biomass boiler - theories and tests in Vaestermalmsverket, Falun; Oeverhettarkorrosion i bioeldad panna - teorier och prov i Vaestermalmsverket, Falun

    Energy Technology Data Exchange (ETDEWEB)

    Roennquist, Eva-Marie

    2000-10-01

    It has lately been evident that a number of biomass-fired plants are experiencing major problems with corrosion of their superheaters. The major aim with this project is to contribute with increased knowledge in this area. The efforts to build up experience around different materials applicable for superheaters with high steam data is of great importance for future plants in Sweden. The main objective for 'Vaermeforsk' has been to transfer the experiences from this investigation to other boilers or plants with different types of fuel. This investigation has therefore been focused on the verification of SYCON's assumptions regarding the roles of chloride and alkali and the possibility of influencing/minimising superheater corrosion by optimisation of the fuel mix. Another important part was to verify that the boiler design, as such, does not create an unfavourable environment for the superheaters by producing reducing zones due to plume formation. Based on the above, the investigation has been divided into three loosely connected parts. (1) The role and reaction by chlorides in the deposits on a superheater, (2) Reducing environment - plume formation of non-combusted fuel, and (3) Choice of materials in the superheater. Serious corrosion has been detected in the superheater tubes of 'Vaestermalmsverket' in Falun. The material temperature was below 530 deg C. No serious inhomogeneous combustion problems or areas with reducing environments have been detected. The corrosion was therefore judged to be caused by alkali chlorides which condense on the superheater tubes. Tests with minor amounts of sulphur added to the biomass fuel have been shown to suppress the generation of alkali chlorides and their condensation on the superheater surfaces. A good correlation between calculated and measured values have been achieved. Very low corrosion rates have been measured on the test probes, constructed with different superheater material and placed in the

  5. Can the lifetime of the superheater tubes be predicted according to the fuel analyses? Assessment from field and laboratory data

    Energy Technology Data Exchange (ETDEWEB)

    Salmenoja, K. [Kvaerner Pulping Oy, Tampere (Finland)

    1998-12-31

    Lifetime of the superheaters in different power boilers is more or less still a mystery. This is especially true in firing biomass based fuels (biofuels), such as bark, forest residues, and straw. Due to the unhomogeneous nature of the biofuels, the lifetime of the superheaters may vary from case to case. Sometimes the lifetime is significantly shorter than originally expected, sometimes no corrosion even in the hottest tubes is observed. This is one of the main reasons why the boiler operators often demand for a better predictability on the corrosion resistance of the materials to avoid unscheduled shutdowns. (orig.) 9 refs.

  6. 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.

  7. Microwave superheaters for fusion

    Energy Technology Data Exchange (ETDEWEB)

    Campbell, R.B.; Hoffman, M.A.; Logan, B.G.

    1987-10-16

    The microwave superheater uses the synchrotron radiation from a thermonuclear plasma to heat gas seeded with an alkali metal to temperatures far above the temperature of material walls. It can improve the efficiency of the Compact Fusion Advanced Rankine (CFAR) cycle described elsewhere in these proceedings. For a proof-of-principle experiment using helium, calculations show that a gas superheat ..delta..T of 2000/sup 0/K is possible when the wall temperature is maintained at 1000/sup 0/K. The concept can be scaled to reactor grade systems. Because of the need for synchrotron radiation, the microwave superheater is best suited for use with plasmas burning an advanced fuel such as D-/sup 3/He. 5 refs.

  8. 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...... thermal cycling took place, both the Ni and Ni2Al3 coatings failed. This highlights the differences between the two biomass plants and demonstrates thata coating solution has to be tailored to the operation conditions of a specific boiler....

  9. SEM Investigation of Superheater Deposits from Biomass-Fired Boilers

    DEFF Research Database (Denmark)

    Jensen, Peter Arendt; Frandsen, Flemming; Hansen, Jørn

    2004-01-01

    Straw is used as fuel in relatively small-scale combined heat and power producing (CHP) grate boilers in Denmark. The large content of potassium and chlorine in straw greatly increases the deposit formation and corrosion of the superheater coils, compared to boilers firing coal. In this study......, mature superheater deposit samples were extracted from two straw-fired boilers, Masnedø and Ensted, with fuel inputs of 33 MWth and 100 MWth, respectively. SEM (scanning electron microscopy) images and EDX (energy dispersive X-ray) analyses were performed on the deposit samples. Different strategies...... are adopted to minimize deposit problems at the two boilers. At Masnedø the final superheater steam temperature is 520 °C, no soot blowing of the superheaters is applied and a relatively large superheater area is used. At Ensted, an external wood-fired superheater is used in order to obtain a final steam...

  10. Superheater Corrosion In Biomass Boilers: Today's Science and Technology

    Energy Technology Data Exchange (ETDEWEB)

    Sharp, William (Sandy) [SharpConsultant

    2011-12-01

    the measured first melting point of fly ash deposits does not necessarily produce a step increase in corrosion rate. Corrosion rate typically accelerates at temperatures below the first melting temperature and mixed deposits may have a broad melting temperature range. Although the environment at a superheater tube surface is initially that of the ash deposits, this chemistry typically changes as the deposits mature. The corrosion rate is controlled by the environment and temperature at the tube surface, which can only be measured indirectly. Some results are counter-intuitive. Two boiler manufacturers and a consortium have developed models to predict fouling and corrosion in biomass boilers in order to specify tube materials for particular operating conditions. It would be very useful to compare the predictions of these models regarding corrosion rates and recommended alloys in the boiler environments where field tests will be performed in the current program. Manufacturers of biomass boilers have concluded that it is more cost-effective to restrict steam temperatures, to co-fire biofuels with high sulfur fuels and/or to use fuel additives rather than try to increase fuel efficiency by operating with superheater tube temperatures above melting temperature of fly ash deposits. Similar strategies have been developed for coal fired and waste-fired boilers. Additives are primarily used to replace alkali metal chloride deposits with higher melting temperature and less corrosive alkali metal sulfate or alkali aluminum silicate deposits. Design modifications that have been shown to control superheater corrosion include adding a radiant pass (empty chamber) between the furnace and the superheater, installing cool tubes immediately upstream of the superheater to trap high chloride deposits, designing superheater banks for quick replacement, using an external superheater that burns a less corrosive biomass fuel, moving circulating fluidized bed (CFB) superheaters from the

  11. NUCLEAR SUPERHEATER FOR BOILING WATER REACTOR

    Science.gov (United States)

    Holl, R.J.; Klecker, R.W.; Graham, C.B.

    1962-05-15

    A description is given of a boiling water reactor having a superheating region integral with the core. The core consists essentially of an annular boiling region surrounding an inner superheating region. Both regions contain fuel elements and are separated by a cylindrical wall, perforations being provided in the lower portion of the cylindrical wall to permit circulation of a common water moderator between the two regions. The superheater region comprises a plurality of tubular fuel assemblies through which the steam emanating from the boiling region passes to the steam outlet. Each superheater fuel assembly has an outer double-walled cylinder, the double walls being concentrically spaced and connected together at their upper ends but open at the bottom to provide for differential thermal expansion of the inner and outer walls. Gas is entrapped in the annulus between the walls which acts as an insulating space between the fissionable material inside and the moderator outside. (AEC)

  12. Neural network for prediction of superheater fireside corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Makkonen, P. [Foster Wheeler Energia Oy, Karhula R and D Center, Karhula (Finland)

    1998-12-31

    Superheater corrosion causes vast annual losses to the power companies. If the corrosion could be reliably predicted, new power plants could be designed accordingly, and knowledge of fuel selection and determination of process conditions could be utilized to minimize superheater corrosion. If relations between inputs and the output are poorly known, conventional models depending on corrosion theories will fail. A prediction model based on a neural network is capable of learning from errors and improving its performance as the amount of data increases. The neural network developed during this study predicts superheater corrosion with 80 % accuracy at early stage of the project. (orig.) 10 refs.

  13. Analysis of Superheater Work Under Creep Conditions

    Directory of Open Access Journals (Sweden)

    Piotr Duda

    2015-03-01

    Full Text Available The aim of this article is work modelling of superheater SH3. It is made of the austenitic stainless steel Super 304H. Its design temperature T is 604 C, and the design pressure P acting on the inner surface of the pipes is 284 bar. The high temperature is the reason of the superheater work under creep conditions. In this article calculations of the optimally mounted coil superheater SH3 are presented. The calculations are carried out first on the basis of the applicable European standards and with the help of the Auto Pipe program. Then, calculations are performed using the ANSYS program based on conducted creep tests and proposed creep equation. The coefficients in creep equation are determined based on the research conducted at the Instytut Metalurgii Żelaza in Gliwice. The model approximates the creep strain as the function of time and stress and this function is presented in the form of a three-dimensional surface . The results of calculations by both methods will be compared and conclusions will be presented. The performed analyzes can estimate the superheater coil remnant life and the usage after the selected time of its operation.

  14. Current Status of Superheat Spray Modeling With NCC

    Science.gov (United States)

    Raju, M. S.; Bulzan, Dan L.

    2012-01-01

    An understanding of liquid fuel behavior at superheat conditions is identified to be a topic of importance in the design of modern supersonic engines. As a part of the NASA's supersonics project office initiative on high altitude emissions, we have undertaken an effort to assess the accuracy of various existing CFD models used in the modeling of superheated sprays. As a part of this investigation, we have completed the implementation of a modeling approach into the national combustion code (NCC), and then applied it to investigate the following three cases: (1) the validation of a flashing jet generated by the sudden release of pressurized R134A from a cylindrical nozzle, (2) the differences between two superheat vaporization models were studied based on both hot and cold flow calculations of a Parker-Hannifin pressure swirl atomizer, (3) the spray characteristics generated by a single-element LDI (Lean Direct Injector) experiment were studied to investigate the differences between superheat and non-superheat conditions. Further details can be found in the paper.

  15. 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

  16. Field test corrosion experiments in Denmark with biomass fuels Part I Straw firing

    DEFF Research Database (Denmark)

    Montgomery, Melanie; Karlsson, A; Larsen, OH

    2002-01-01

    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 temperature range of 450-600°C...

  17. Computer modeling of a convective steam superheater

    Science.gov (United States)

    Trojan, Marcin

    2015-03-01

    Superheater is for generating superheated steam from the saturated steam from the evaporator outlet. In the case of pulverized coal fired boiler, a relatively small amount of ash causes problems with ash fouling on the heating surfaces, including the superheaters. In the convection pass of the boiler, the flue gas temperature is lower and ash deposits can be loose or sintered. Ash fouling not only reduces heat transfer from the flue gas to the steam, but also is the cause of a higher pressure drop on the flue gas flow path. In the case the pressure drop is greater than the power consumed by the fan increases. If the superheater surfaces are covered with ash than the steam temperature at the outlet of the superheater stages falls, and the flow rates of the water injected into attemperator should be reduced. There is also an increase in flue gas temperature after the different stages of the superheater. Consequently, this leads to a reduction in boiler efficiency. The paper presents the results of computational fluid dynamics simulations of the first stage superheater of both the boiler OP-210M using the commercial software. The temperature distributions of the steam and flue gas along the way they flow together with temperature of the tube walls and temperature of the ash deposits will be determined. The calculated steam temperature is compared with measurement results. Knowledge of these temperatures is of great practical importance because it allows to choose the grade of steel for a given superheater stage. Using the developed model of the superheater to determine its degree of ash fouling in the on-line mode one can control the activation frequency of steam sootblowers.

  18. Irradiation test of Dupic fuel

    International Nuclear Information System (INIS)

    Kee, Chan Song; Myung, Seung Yang; Hyun, Soo Park

    2002-01-01

    Simulated DUPIC fuel that had been fabricated from natural uranium oxide with simulated fission products was irradiated at the HANARO research reactor at KAERI in 1999. The objectives of this irradiation test were to estimate the in-core behaviour of DUPIC fuel, to verify the design of the non-instrumented irradiation rig developed for the irradiation test of DUPIC fuel and to ensure the irradiation requirements of DUPIC fuel at HANARO. The post-irradiation examinations, such as dimensional measurement, γ-scanning and EPMA, for irradiated simulated DUPIC fuel have been performed at the IMEF. The irradiation test of DUPIC fuel, fabricated with spent PWR fuel material, was performed at HANARO for two months as of May 2000. The resultant burn-up of irradiated DUPIC fuel was estimated to be 1 800 MWd/MTU. The irradiation behaviour of DUPIC fuel will be investigated based on the data from post-irradiation examinations. (authors)

  19. A fault tolerant superheat control strategy for supermarket refrigeration systems

    DEFF Research Database (Denmark)

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

    2013-01-01

    In this paper, a fault tolerant control (FTC) strategy is proposed for evaporator superheat control in supermarket refrigeration systems. Conventional control uses a pressure and temperature sensor for this purpose, however, the pressure sensor can fail to function. A contingency control strategy......, 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...... in a plug & play fashion. The strategy is outlined by means of procedural steps as well as a flow chart that also illustrates the process of automatic tuning of the maximum slope-seeking controller. Test results are furthermore presented for a display case in a full scale CO2 supermarket refrigeration...

  20. 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

  1. Lifetime evaluation of superheater tubes exposed to steam oxidation, high temperature corrosion and creep

    Energy Technology Data Exchange (ETDEWEB)

    Henriksen, N. [Elsamprojekt A/S, Faelleskemikerne, Fredericia (Denmark); Hede Larsen, O.; Blum, R. [I/S Fynsvaerket, Faelleskemikerne, Odense (Denmark)

    1996-12-01

    Advanced fossil fired plants operating at high steam temperatures require careful design of the superheaters. The German TRD design code normally used in Denmark is not precise enough for the design of superheaters with long lifetimes. The authors have developed a computer program to be used in the evaluation of superheater tube lifetime based on input related to tube dimensions, material, pressure, steam temperature, mass flux, heat flux and estimated corrosion rates. The program is described in the paper. As far as practically feasible, the model seems to give a true picture of the reality. For superheaters exposed to high heat fluxes or low internal heat transfer coefficients as is the case for superheaters located in fluidized bed environments or radiant environments, the program has been extremely useful for evaluation of surface temperature, oxide formation and lifetime. The total uncertainty of the method is mainly influenced by the uncertainty of the determination of the corrosion rate. More precise models describing the corrosion rate as a function of tube surface temperature, fuel parameters and boiler parameters need to be developed. (au) 21 refs.

  2. 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.

  3. 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.

  4. 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...... at low superheat. The parameters in the proposed controller structure can automatically be chosen based on two open loop tests. Results from tests on two different refrigeration systems indicate that the proposed controller can control the evaporator superheat to a low level giving close to optimal...... 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....

  5. FUEL ASSEMBLY SHAKER TEST SIMULATION

    Energy Technology Data Exchange (ETDEWEB)

    Klymyshyn, Nicholas A.; Sanborn, Scott E.; Adkins, Harold E.; Hanson, Brady D.

    2013-05-30

    This report describes the modeling of a PWR fuel assembly under dynamic shock loading in support of the Sandia National Laboratories (SNL) shaker test campaign. The focus of the test campaign is on evaluating the response of used fuel to shock and vibration loads that a can occur during highway transport. Modeling began in 2012 using an LS-DYNA fuel assembly model that was first created for modeling impact scenarios. SNL’s proposed test scenario was simulated through analysis and the calculated results helped guide the instrumentation and other aspects of the testing. During FY 2013, the fuel assembly model was refined to better represent the test surrogate. Analysis of the proposed loads suggested the frequency band needed to be lowered to attempt to excite the lower natural frequencies of the fuel assembly. Despite SNL’s expansion of lower frequency components in their five shock realizations, pretest predictions suggested a very mild dynamic response to the test loading. After testing was completed, one specific shock case was modeled, using recorded accelerometer data to excite the model. Direct comparison of predicted strain in the cladding was made to the recorded strain gauge data. The magnitude of both sets of strain (calculated and recorded) are very low, compared to the expected yield strength of the Zircaloy-4 material. The model was accurate enough to predict that no yielding of the cladding was expected, but its precision at predicting micro strains is questionable. The SNL test data offers some opportunity for validation of the finite element model, but the specific loading conditions of the testing only excite the fuel assembly to respond in a limited manner. For example, the test accelerations were not strong enough to substantially drive the fuel assembly out of contact with the basket. Under this test scenario, the fuel assembly model does a reasonable job of approximating actual fuel assembly response, a claim that can be verified through

  6. 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...... a pressure and a temperature sensor. In this paper we show, through extensive testing, that the superheat or filling of the evaporator can actually be controlled using only a single temperature sensor. This can either reduce commissioning costs by lowering the necessary amount of sensors or add fault...... tolerance in existing systems if a sensor fails (e.g. pressure sensor). The solution is based on a novel maximum slope-seeking control method, where a perturbation signal is added to the valve opening degree, which gives additional information about the system for control purposes. Furthermore, the method...

  7. 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...

  8. Compact Fuel Element Environment Test

    Science.gov (United States)

    Bradley, D. E.; Mireles, O. R.; Hickman, R. R.; Broadway, J. W.

    2012-01-01

    Deep space missions with large payloads require high specific impulse (I(sub sp)) and relatively high thrust to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average I(sub sp). Nuclear thermal rockets (NTRs) capable of high I(sub sp) 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 3,000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements that 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 noncontact radio frequency heating and expose samples to hydrogen for typical mission durations has been developed to assist in optimal material and manufacturing process selection without employing fissile material. This Technical Memorandum details the test bed design and results of testing conducted to date.

  9. 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)

  10. CANFLEX fuel bundle impact test

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. Study made of corrosion resistance of stainless steel and nickel alloys in nuclear reactor superheaters

    Science.gov (United States)

    Greenberg, S.; Hart, R. K.; Lee, R. H.; Ruther, W. E.; Schlueter, R. R.

    1967-01-01

    Experiments performed under conditions found in nuclear reactor superheaters determine the corrosion rate of stainless steel and nickel alloys used in them. Electropolishing was the primary surface treatment before the corrosion test. Corrosion is determined by weight loss of specimens after defilming.

  12. 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....

  13. 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...

  14. 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...

  15. 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

  16. Biomass fired superheater for more efficient electricity generation from waste incineration plants. Final technical report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    A full scale Demonstration Plant has been constructed and tested at Haslev Combined Heat and Power Plant, which produces about 5 MW electric power and 13 MJ/s heat, enough to supply approximately 2500 households. The demonstration plant was commissioned in the spring of 1996. The results of testing are successful and show a capacity well above the design value. The concept is now developed to a commercial stage. The main component of the demonstration plant is a full scale carbonising plant of 3 MW. The plant is fired with straw, which is carbonised to 1 MJ/s product gas and 2 MJ/s char. Due to a low process temperature and absence of air, the corrosive matters remain in the char fraction, which is used as topping fuel in the steam boiler. The produced gas is utilised in a separate combustion chamber producing a flue gas with very low content of alkali metals, chlorides and particles, making the flue gas much less corrosive at high steam temperatures than flue gasses from direct combustion of municipal solid waste or straw. Therefore, the flue gas is used to boost steam data in an external superheater without increasing the corrosion speed. If the concept is used, when building new built plants, the electricity gain can be increased by typically 10-15% on a given heat demand, this is an improvement of approximately 3% points in electricity efficiency. (au)

  17. 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...

  18. 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

  19. 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

  20. 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....

  1. 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

  2. Tier 3 Certification Fuel Impacts Test Program

    Science.gov (United States)

    The recent Tier 3 regulations for light duty vehicles introduced a new certification fuel designed to be more characteristic of current market fuels. A laboratory test program was conducted to measure differences in CO2 and fuel economy between the current and future certificatio...

  3. High-Speed Imaging of Explosive Droplet Boiling at the Superheat Limit

    Science.gov (United States)

    Ferris, F. Robert; Hermanson, Jim; Asadollahi, Arash; Esmaeeli, Asghar

    2017-11-01

    The explosive boiling processes of droplets of diethyl ether (1-2 mm in diameter) at the superheat limit were examined both experimentally and computationally. Experimentally, droplet explosion was studied using a heated bubble column to bring the test droplet to the superheat limit. The droplet fluid was diethyl ether (superheat limit 147 C at 1 bar) with immiscible glycerol employed as the heated host fluid. Tests were carried out at pressures between 0.5 and 4 bar absolute. The pressure rise associated with the explosive boiling event was captured using a piezoelectric quartz pressure transducer with a 1 MHz DAQ system. High-speed imaging of the interfacial behavior during explosive boiling was performed using a Phantom v12.1 camera at a frame rate of up to one million frames per second with the droplets illuminated by diffuse back-lighting. The imaging reveals features of the Rayleigh-Taylor instability at the vapor-liquid interface resulting from the unstable boiling process. Computationally, Direct Numerical Simulations are performed at Southern Illinois University Carbondale to compliment the experimental tests. NSF Award Number 1511152.

  4. 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.

  5. Nonlinear FOPDT Model Identification for the Superheat Dynamic in a Refrigeration System

    DEFF Research Database (Denmark)

    Yang, Zhenyu; Sun, Zhen; Andersen, Casper

    2011-01-01

    An on-line nonlinear FOPDT system identification method is proposed and applied to model the superheat dynamic in a supermarket refrigeration system. The considered nonlinear FOPDT model is an extension of the standard FOPDT model by means that its parameters are time dependent. After...... the considered system is discretized, the nonlinear FOPDT identification problem is formulated as a Mixed Integer Non-Linear Programming problem, and then an identification algorithm is proposed by combining the Branch-and-Bound method and Least Square technique, in order to on-line identify these time......-dependent parameters. The proposed method is firstly tested through a number of numerical examples, and then applied to model the superheat dynamic in a supermarket refrigeration system based on experimental data. As shown in these studies, the proposed method is quite promising in terms of reasonable accuracy, large...

  6. 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...

  7. REACTOR FUEL ELEMENTS TESTING CONTAINER

    Science.gov (United States)

    Whitham, G.K.; Smith, R.R.

    1963-01-15

    This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

  8. 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.

  9. Fuel cell hybrid drive train test facility

    NARCIS (Netherlands)

    J. Bruinsma; I. Zafina; H. Bosma; Edwin Tazelaar; Bram Veenhuizen

    2009-01-01

    Fuel cells are expected to play an important role in the near future as prime energy source on board of road-going vehicles. In order to be able to test all important functional aspects of a fuel cell hybrid drive train, the Automotive Institute of the HAN University has decided to realize a

  10. 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

  11. 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)

  12. 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

  13. 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

  14. 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...

  15. 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.

  16. 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)

  17. Test report : alternative fuels propulsion durability evaluation

    Science.gov (United States)

    2012-08-28

    This document, prepared by Honeywell Aerospace, Phoenix, AZ (Honeywell), contains the final : test report (public version) for the U.S. Department of Transportation/Federal Aviation : Administration (USDOT/FAA) Alternative Fuels Propulsion Engine Dur...

  18. 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.

  19. Inert tube coatings as a method to reduce deposit adhesion on superheaters; Inerta tubytbelaeggningar som metod att minska paaslagens vidhaeftning paa oeverhettare

    Energy Technology Data Exchange (ETDEWEB)

    Almark, Matts; Staalenheim, Annika; Henderson, Pamela

    2007-12-15

    In many biofuel and waste fired boilers there are significant deposit related problems. The fouling of the superheaters and other heat transfer surfaces reduces the heat transfer, which leads to increased flue gas temperatures, or might block the entire flue gas channel in extreme cases. An increasing share of waste and refuse derived fuels as well as the use of new types of biomass have led to increased problems with deposit formation. In order to minimize the formation of deposits and to maintain low flue gas temperatures the superheaters are subjected to frequent soot blowing, which in turn may lead to increased material loss of the tubes. The purpose of this project is to try to show that tube surfaces that may be regarded as inert have a 'non-stick' effect and that the fouling is reduced when the deposits do not react with the tube surface layer, and the adhesive force of the deposits are reduced. Nickel based alloy Sanicro 63, which forms a nickel oxide surface layer, and Kanthal APM, which forms an aluminum oxide surface layer, are compared with 15Mo3, a common superheater steels forming iron oxide and Sanicro 28, a stainless steel which forms iron and chromium oxides. Applied coatings are also tested, welded Alloy 625 and sprayed Kanthal APM, in order to investigate how ht application method interferes with the results from the pure material. A ceramic coating material that is claimed to give good results in waste and coal fired boilers is also tested. Tests with cooled probes, on which the tested materials are mounted, are performed in two different boilers, Haendeloe P14 representing waste fired boilers with fouling issues related to chlorine and heavy metals, and Myllykoski K7 representing a forest industry with non-corrosive sulfate containing deposits. The results show that a tube surface of a nickel-based alloy can reduce the formation of chlorine rich, partly melted deposits. No effect was shown on the dry, chlorine-free deposits. The

  20. 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

  1. Irradiated MTR fuel assemblies sipping test

    Energy Technology Data Exchange (ETDEWEB)

    Perrotta, J.A.; Terremoto, Luis A.A.; Zeituni, Carlos A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Div. de Engenharia do Nucleo

    1997-10-01

    This paper describes the procedure and methodology used to perform sipping test with the IEA-R1 fuel assemblies at the storage pool, and presents the results obtained for Cs-137 sipping water activity for each fuel assembly analyzed. Discussion is made correlating corrosion pits to the activity values measured. A Cs-137 leaking rate is determined which can be compared to the criteria established for canning spent fuel assemblies inside the pool of for shipment abroad. 3 refs., 13 figs., 1 tab.

  2. Irradiated MTR fuel assemblies sipping test

    International Nuclear Information System (INIS)

    Perrotta, J.A.; Terremoto, Luis A.A.; Zeituni, Carlos A.

    1997-01-01

    This paper describes the procedure and methodology used to perform sipping test with the IEA-R1 fuel assemblies at the storage pool, and presents the results obtained for Cs-137 sipping water activity for each fuel assembly analyzed. Discussion is made correlating corrosion pits to the activity values measured. A Cs-137 leaking rate is determined which can be compared to the criteria established for canning spent fuel assemblies inside the pool of for shipment abroad. 3 refs., 13 figs., 1 tab

  3. Fuel accident performance testing for small HTRs

    Science.gov (United States)

    Schenk, W.; Pott, G.; Nabielek, H.

    1990-04-01

    Irradiated spherical fuel elements containing 16400 coated UO 2 particles each were heated at temperatures between 1600 and 1800°C and the fission product release was measured. The demonstrated fission product retention at 1600°C establishes the basis for the design of small modular HTRs which inherently limit the temperature to 1600°C by passive means. In addition to this demonstration, the test data show that modern TRISO fuels provide an ample performance margin: release normally sets in at 1800°C; this occurs at 1600°C only with fuels irradiated under conditions which significantly exceed current reactor design requirements.

  4. 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

  5. CANDU fuel compression tests at elevated temperatures

    International Nuclear Information System (INIS)

    Koehn, E.; Chan, J.K.; Langman, V.J.; Hadaller, G.I.; Fortman, R.A.

    1995-01-01

    An inlet header large break loss of coolant accident (LOCA) in CANDU reactors with fuelling against flow can cause the fuel to shift in the channels with a consequent reactivity insertion. This results in an increased fuel power transient, and a potential increase in the mialyzed consequences for such events. As the reactor's age and the channel axial gaps increase, the magnitude of the predicted power u-dmient increases. A design solution to reduce the power transient is to limit the amount of fuel movement by reducing the channel axial gap. This solution was implemented into Ontario Hydro's Bruce B and Darlington reactors. A consequence of a reduced channel axial gap is the potential for the fuel column axial expansion to become constrained by the channel end components in large break LOCAs. This experimental program investigated the effects of pellet cracking and elevated sheath temperatures on the ability of the fuel elements, of the 37-element bundle design, to sustain axial loads. The unirradiated fuel elements tested were either in the as-received condition or with the U0 2 fuel pellets cracked in a mechanical process to simulate the effect of inufflation. The load deformation characteristics demonstrated that, for a given amount of axial compression. the loads sustainable by the elements at elevated sheath temperatures were low. As a result. excess axial expansion would be easily accommodated without further challenge to pressure tube integrity. (author)

  6. 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)

  7. 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

  8. The OSU Hydro-Mechanical Fuel Test Facility: Standard Fuel Element Testing

    Energy Technology Data Exchange (ETDEWEB)

    Wade R. Marcum; Brian G. Woods; Ann Marie Phillips; Richard G. Ambrosek; James D. Wiest; Daniel M. Wachs

    2001-10-01

    Oregon State University (OSU) and the Idaho National Laboratory (INL) are currently collaborating on a test program which entails hydro-mechanical testing of a generic plate type fuel element, or standard fuel element (SFE), for the purpose of qualitatively demonstrating mechanical integrity of uranium-molybdenum monolithic plates as compared to that of uranium aluminum dispersion, and aluminum fuel plates due to hydraulic forces. This test program supports ongoing work conducted for/by the fuel development program and will take place at OSU in the Hydro-Mechanical Fuel Test Facility (HMFTF). Discussion of a preliminary test matrix, SFE design, measurement and instrumentation techniques, and facility description are detailed in this paper.

  9. 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)

  10. 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%

  11. 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

  12. 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.)

  13. Fuel tank integrity research : fuel tank analyses and test plans

    Science.gov (United States)

    2013-04-15

    The Federal Railroad Administrations Office of Research : and Development is conducting research into fuel tank : crashworthiness. Fuel tank research is being performed to : determine strategies for increasing the fuel tank impact : resistance to ...

  14. Fully Fueled TACOM Vehicle Storage Test Program.

    Science.gov (United States)

    1981-12-01

    AFLRL with a water bottom were tested as control samples. This fuel sample had been previously innoculated with a culture of Cladosporium resinae and was...turbid, light pink color * Containing active growth of Cladosporium resinae ** Sample was shaken and allowed to stand for 24 hours prior to obtaining

  15. 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......, pressure, and stoichiometry at varying current density. Furthermore, procedures for polarization curve recording were also tested both in ascending and descending current directions....

  16. 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.

  17. 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)

  18. 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

  19. 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

  20. 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

  1. 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.

  2. 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

  3. 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

  4. 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.

  5. Out of pile testing of the PHWR fuel bundles

    International Nuclear Information System (INIS)

    Mahender Dev; Raghunathan, S.; Agarwal, G.K.; Patel, R.J.; Agarwal, R.G.

    2002-01-01

    In PHWRs fuel bundle resides in the form of a string in the coolant channels. These fuel bundles are required to be replaced periodically with the help of fuelling machine and spent fuel is discharged to the spent-fuel bay through fuel transfer system. During complete refuelling operation, and during residence in channel fuel bundle experiences various kinds of loads like drag force, impact force, force applied by Fuelling Machine ram and force applied by various actuators in fuel transfer system. These fuel bundles are manufactured indigenously and require out of pile testing for qualification of design as well as manufacturing process. In 220 MWe PHWRs, 19-element split spacer fuel bundle is used whereas in 500 MWe PHWRs 37-element fuel bundle will be used. A comprehensive programme was conducted to generate, basic data like estimation of loads coming on fuel bundles, experimental data generation about friction factor and pressure drop and carrying out of pile testing of 19-element fuel bundles in Integral Thermal Facility at Hall-7. The 37-element fuel bundles were tested in fuel locator test facility at simulated reactor conditions for pressure drop test, endurance test and cross flow test. The 37-element bundles have also been tested for flow-induced vibration during residence in the reactor. The paper describes the experimental techniques and setups, for simulating the reactor condition and determining the effect of those conditions on the fuel bundles. (author)

  6. Development of Failed Fuel Detection System for HANARO Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sung Ho; Sim, Bong Sik; Park, Su Ki; Chi, Dae Young; Park, Kook Nam; Lee, Chung Young; Lee, Jong Min; Kim, Hark Rho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-10-15

    FTL (Fuel Test Loop) is a test facility which could conduct fuel irradiation test at HANARO reactor. The FFDS (Failed Fuel Detection System) is installed in the FTL for detection of test fuel failure. The signals from the FFDS are interconnected with HANARO reactor protection system via FTL protection panels for reactor trip. In this paper, a design, manufacturing and functional test results of the FFDS are introduced.

  7. Test Results of PBMR Fuel Spheres

    International Nuclear Information System (INIS)

    Koshcheev, Konstantin; Diakov, Alexander; Beltyukov, Igor; Barybin, Andrey; Chernetsov, Mikhail

    2014-01-01

    Results of pre-irradiation testing of fuel spheres (FS) and coated particles (CP) manufactured by PBMR SOC (Republic of South Africa) are described. The stable high quality level of major characteristics (dimensions, CP coating structure, uranium-235 contamination of the FS matrix graphite and the outer PyC layer of the CP coating) are shown. Results of a methodical irradiation test of two FS in helium and neon medium at temperatures of 800 to 1300 °C with simultaneous determination of release-to-birth ratios for major gaseous fission products (GFP) are described. (author)

  8. 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.

  9. TESTING AND ACCEPTANCE OF FUEL PLATES FOR RERTR FUEL DEVELOPMENT EXPERIMENTS

    Energy Technology Data Exchange (ETDEWEB)

    J.M. Wight; G.A. Moore; S.C. Taylor

    2008-10-01

    This paper discusses how candidate fuel plates for RERTR Fuel Development experiments are examined and tested for acceptance prior to reactor insertion. These tests include destructive and nondestructive examinations (DE and NDE). The DE includes blister annealing for dispersion fuel plates, bend testing of adjacent cladding, and microscopic examination of archive fuel plates. The NDE includes Ultrasonic (UT) scanning and radiography. UT tests include an ultrasonic scan for areas of “debonds” and a high frequency ultrasonic scan to determine the "minimum cladding" over the fuel. Radiography inspections include identifying fuel outside of the maximum fuel zone and measurements and calculations for fuel density. Details of each test are provided and acceptance criteria are defined. These tests help to provide a high level of confidence the fuel plate will perform in the reactor without a breach in the cladding.

  10. Alkali chloride induced corrosion of superheaters under biomass firing conditions: Improved insights from laboratory scale studies

    DEFF Research Database (Denmark)

    Okoro, Sunday Chukwudi; Montgomery, Melanie; Jappe Frandsen, Flemming

    2015-01-01

    One of the major operational challenges experienced by power plants firing biomass is the high corrosion rate of superheaters. This limits the outlet steam temperature of the superheaters and consequently, the efficiency of the power plants. The high corrosion rates have been attributed to the fo...

  11. Liquefied Gaseous Fuels Spill Test Facility

    International Nuclear Information System (INIS)

    1993-02-01

    The US Department of Energy's liquefied Gaseous Fuels Spill Test Facility is a research and demonstration facility available on a user-fee basis to private and public sector test and training sponsors concerned with safety aspects of hazardous chemicals. Though initially designed to accommodate large liquefied natural gas releases, the Spill Test Facility (STF) can also accommodate hazardous materials training and safety-related testing of most chemicals in commercial use. The STF is located at DOE's Nevada Test Site near Mercury, Nevada, USA. Utilization of the Spill Test Facility provides a unique opportunity for industry and other users to conduct hazardous materials testing and training. The Spill Test Facility is the only facility of its kind for either large- or small-scale testing of hazardous and toxic fluids including wind tunnel testing under controlled conditions. It is ideally suited for test sponsors to develop verified data on prevention, mitigation, clean-up, and environmental effects of toxic and hazardous gaseous liquids. The facility site also supports structured training for hazardous spills, mitigation, and clean-up. Since 1986, the Spill Test Facility has been utilized for releases to evaluate the patterns of dispersion, mitigation techniques, and combustion characteristics of select materials. Use of the facility can also aid users in developing emergency planning under US P.L 99-499, the Superfund Amendments and Reauthorization Act of 1986 (SARA) and other regulations. The Spill Test Facility Program is managed by the US Department of Energy (DOE), Office of Fossil Energy (FE) with the support and assistance of other divisions of US DOE and the US Government. DOE/FE serves as facilitator and business manager for the Spill Test Facility and site. This brief document is designed to acquaint a potential user of the Spill Test Facility with an outline of the procedures and policies associated with the use of the facility

  12. Kerosene-Fuel Engine Testing Under Way

    Science.gov (United States)

    2003-01-01

    NASA Stennis Space Center engineers conducted a successful cold-flow test of an RS-84 engine component Sept. 24. The RS-84 is a reusable engine fueled by rocket propellant - a special blend of kerosene - designed to power future flight vehicles. Liquid oxygen was blown through the RS-84 subscale preburner to characterize the test facility's performance and the hardware's resistance. Engineers are now moving into the next phase, hot-fire testing, which is expected to continue into February 2004. The RS-84 engine prototype, developed by the Rocketdyne Propulsion and Power division of The Boeing Co. of Canoga Park, Calif., is one of two competing Rocket Engine Prototype technologies - a key element of NASA's Next Generation Launch Technology program.

  13. 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)

  14. TRESP II testing of AFS dual fuel system.

    Science.gov (United States)

    2013-11-01

    A dual fuel CNG and diesel system was retrofitted to a 13 L Volvo semi tractor for testing to verify the : fuel economy and CNG substitution rate. The semi tractor was tested on interstate and mountainous : highway routes with a loaded trailer. Fuel ...

  15. 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

  16. 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)

  17. MCO Engineering Test Report Fuel Basket Handling Grapple Acceptance Test

    International Nuclear Information System (INIS)

    CHENAULT, D.M.

    2000-01-01

    Acceptance testing of the production SNF Fuel Basket lift grapples to the required 150 percent maximum lift load is documented herein. The report shows the results affirming the proof test passage. The primary objective of this test was to confirm the load rating of the grapple per applicable requirements of ANSI 14 6 American National Standard For Radioactive Materials Special Lifting Devices for Shipping Containers Weighing 10,000 pounds (4500kg) or More. The above Standard requires a load test of 150% of the design load which must be held for a minimum of 10 minutes followed by a Liquid Penetrant or Magnetic Particle examination of critical areas and welds in accordance with the ANSI/ASME Boiler and Pressure Vessel Code 1989 Section 111 Division 1 section NF 5350

  18. 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.

  19. Canadian experience in irradiation and testing of MOX fuel

    Science.gov (United States)

    Yatabe, S.; Floyd, M.; Dimayuga, F.

    2018-04-01

    Experimental irradiation and performance testing of Mixed OXide (MOX) fuel at the Canadian Nuclear Laboratories (CNL) has taken place for more than 40 years. These experiments investigated MOX fuel behaviour and compared it with UO2 behaviour to develop and verify fuel performance models. This article compares the performance of MOX of various concentrations and homogeneities, under different irradiation conditions. These results can be applied to future fuel designs. MOX fuel irradiated by CNL was found to be comparable in performance to similarly designed and operated UO2 fuel. MOX differs in behaviour from UO2 fuel in several ways. Fission-gas release, grain growth and the thickness of zirconium oxide on the inner sheath appear to be related to MOX fuel homogeneity. Columnar grains formed at the pellet centre begin to develop at lower powers in MOX than in UO2 fuel.

  20. 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

  1. 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

  2. Sodium flow distribution in test fuel assembly P-23B

    International Nuclear Information System (INIS)

    Taylor, J.P.S.

    1978-08-01

    Relatively large cladding diametral increases in the exterior fuel pins of HEDL's test fuel subassembly P-23B were successfully explained by a thermal-hydraulic/solid mechanics analysis. This analysis indicates that while at power, the subassembly flow was less than planned and that the fuel pins were considerably displaced and bowed from their nominal position. In accomplishing this analysis, a method was developed to estimate the sodium flow distribution and pin distortions in a fuel subassembly at power

  3. Test plan for K-Basin fuel handling tools

    Energy Technology Data Exchange (ETDEWEB)

    Bridges, A.E.

    1995-02-08

    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.

  4. 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

  5. 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.

  6. 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

  7. Simultaneous control of drying temperature and superheat for a closed-loop heat pump dryer

    International Nuclear Information System (INIS)

    Yang, Zhao; Zhu, Zongsheng; Zhao, Feng

    2016-01-01

    Highlights: • In a closed-loop heat pump drying system, the drying process is divided into two processes: heating and cooling. • A simultaneous control strategy is proposed to improve the precision of superheat and drying temperature. • Two distinct fuzzy controllers are employed to achieve more stable drying temperature. • The superheat controller is mainly composed of a PID controller and a conversion emergency controller. • The experimental data show that the simultaneous controller decreases the nonlinearities of superheat and drying temperature. - Abstract: Closed-loop heat pump drying system has its particular characteristics, and in this paper a simultaneous control strategy is proposed to improve the precision of superheat and drying temperature. The stability of drying temperature guarantees the quality of drying material. And on the premise, superheat should be accurate and stable to improve energy usage. Two fuzzy controllers employing different membership functions and control rules are used to achieve more stable drying temperature. The superheat controller is mainly composed of a PID controller and a conversion emergency controller which accelerates the response of electronic expansion valve. The experimental results show that the temperatures at the outlet of indoor and outdoor condensers give smaller fluctuations. The variations of superheat are investigated when the temperature of drying chamber changes from ambient temperature to 25/30 °C, from 25 to 30 °C and from 30 to 25 °C, respectively. All the experimental data confirm that the new controller is applicable to decrease the nonlinearities of superheat and drying temperature.

  8. 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

  9. Structural geology report: Spent Fuel Test - Climax Nevada Test Site

    International Nuclear Information System (INIS)

    Wilder, D.G.; Yow, J.L. Jr.

    1984-10-01

    We performed underground mapping and core logging in the Climax Stock, a granitic intrusive at the Nevada Test Site, as part of a major field test to determine the feasibility of using granitic or crystalline rock for the underground storage of spent fuel from a nuclear reactor. This mapping and logging identified more than 2500 fractures, over 1500 of which were described in enough detail to allow statistical analyses and orientation studies to be performed. We identified eight joint sets, three major shear sets, and a fault zone within the Spent Fuel Test - Climax (SFT-C) portion of the Stock. Joint sets identified within the SFT-C and elsewhere in the Stock correlated well. The orientations of joint sets identified by other investigators were consistent with our findings, indicating that the joint sets are persistent and have a relatively uniform orientation throughout a major portion of the Stock. The one joint set not seen elsewhere in the Stock is healed and the wall rock is altered, implying that healed joints were not included in the mapping criteria used by other investigators. The shear sets were distinguished from the joint sets by virtue of crushed minerals, continuous clay infilling, and other evidences of shearing, and from faults by the lack of offsetting. Previous investigators working mainly in the Pile Driver Drifts identified two of the shear sets. The third set, being nearly parallel to these Drifts had not been identified previously. The fault zone identified at the far (Receiving Room) end of the project is oriented approximately N45 0 E-75 0 SE, similar to both the Boundary and Shaft Station Faults. We have, therefore, concluded that the Receiving Room Fault is one of a series of normal faults that occur within the Climax Stock and that are possibly related, in both age and genesis, to the Boundary Fault. 52 refs., 26 figs., 11 tabs

  10. Overview on MOX fuel for LWRs: Design, performance and testing

    International Nuclear Information System (INIS)

    Brown, C.; Callens, C.; Goll, W.; Lippens, M.

    2000-01-01

    This overview looks at the historical background to the design, performance and testing of LWR MOX fuel over the last 30 to 40 years. It briefly examines the scenarios which encouraged the development of MOX fuel for utilisation in LWRs and looks at the design changes required on moving from UO 2 to MOX fuel. The paper summarises the national irradiation testing programmes, the commercial developments and performance data obtained throughout this period, highlighting those aspects which have had an impact on manufacturing and design choices. The paper thus provides the historical background information for the contributed papers in Session 3 (Fuel Design, Performance and Testing) of the symposium. (author)

  11. 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

  12. Fuel Cell Testing - Degradation of Fuel Cells and its Impact on Fuel Cell Applications

    OpenAIRE

    PFRANG Andreas

    2008-01-01

    Fuel cells are expected to play a major role in the future energy supply, especially polymer electrolyte membrane fuel cells could become an integral part in future cars. Reduction of degradation of fuel cell performance while keeping fuel cell cost under control is the key for an introduction into mass markets.

  13. Fuel and Fuel Additive Registration Testing of Ethanol-Diesel Blend for O2Diesel, Inc.

    Energy Technology Data Exchange (ETDEWEB)

    Fanick, E. R.

    2004-02-01

    O2 Diesel Inc. (formerly AAE Technologies Inc.) tested a heavy duty engine with O2Diesel (diesel fuel with 7.7% ethanol and additives) for regulated emissions and speciation of vapor-phase and semi-volatile hydrocarbon compounds. This testing was performed in support of EPA requirements for registering designated fuels and fuel additives as stipulated by sections 211(b) and 211(e) of the Clean Air Act.

  14. Kinetic parameters of a material test research reactor fueled with various low enriched uranium dispersion fuels

    International Nuclear Information System (INIS)

    Muhammad, Farhan; Majid, Asad

    2009-01-01

    The effects of using different low enriched uranium fuels, having same uranium density, on the kinetic parameters of a material test research reactor were studied. For this purpose, the original aluminide fuel (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -Al) dispersion fuels having the same uranium density as of the original fuel. Simulations were carried out to calculate prompt neutron generation time, effective delayed-neutron fraction, core excess reactivity and neutron flux spectrum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were used to carry out these calculations. It was observed that both the silicide fuels had the same prompt neutron generation time 0.02% more than that of the original aluminide fuel, while the oxide fuel had a prompt neutron generation time 0.05% less than that of the original aluminide fuel. The effective delayed-neutron fraction decreased for all the fuels; the decrease was maximum at 0.06% for U 3 Si 2 -Al followed by 0.03% for U 3 Si-Al, and 0.01% for U 3 O 8 -Al fuel. The U 3 O 8 -Al fueled reactor gave the maximum ρ excess at BOL which was 21.67% more than the original fuel followed by U 3 Si-Al which was 2.55% more, while that of U 3 Si 2 -Al was 2.50% more than the original UAl x -Al fuel. The neutron flux of all the fuels was more thermalized, than in the original fuel, in the active fuel region of the core. The thermalization was maximum for U 3 O 8 -Al followed by U 3 Si-Al and then U 3 Si 2 -Al fuel.

  15. 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 Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR TESTING, EVALUATION, AND APPROVAL OF... closure restrict the outflow of liquid fuel. ...

  16. 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

  17. 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

  18. Fast Flux Test Facility metal fuel pin fabrication

    International Nuclear Information System (INIS)

    Benecke, M.W.; Dittmer, J.O.; Feigenbutz, L.V.

    1989-05-01

    A new initiative to develop, irradiate, and qualify a binary uranium/zirconium metal fuel system in the Fast Flux Test Facility (FFTF) has been implemented by the Westinghouse Hanford Company for the US Department of Energy. Metal fuel test assemblies have been designed and fabricated and are now being irradiated in FFTF to provide the data needed to support the potential use of binary fuels in FFTF and other liquid metal reactors. These development efforts support licensing activities for metal fuel use in near-term advanced liquid metal reactors. New metal fuel pin design features, fabrication development, and fabrication processes for three metal fuel tests will be described and their irradiation status reported in this paper. 11 figs

  19. 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

  20. 76 FR 5319 - Regulation of Fuel and Fuel Additives: Alternative Test Method for Olefins in Gasoline

    Science.gov (United States)

    2011-01-31

    ... producing gasoline are required to test Reformulated Gasoline (RFG), and conventional gasoline (CG) for... 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...

  1. 76 FR 65382 - Regulation of Fuel and Fuel Additives: Alternative Test Method for Olefins in Gasoline

    Science.gov (United States)

    2011-10-21

    ... blenders producing gasoline are required to test Reformulated Gasoline (RFG), and conventional gasoline (CG... Regulation of Fuel and Fuel Additives: Alternative Test Method for Olefins in Gasoline AGENCY: Environmental... gasoline. This final rule will provide flexibility to the regulated community by allowing an additional...

  2. 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)

  3. A higher chromium weld overlay alloy for waterwalls and superheaters

    Energy Technology Data Exchange (ETDEWEB)

    Paul, Larry; Clark, Gregg [ThyssenKrupp VDM USA, Inc., Tipton, IN (United States)

    2009-08-15

    The use of corrosion resistant weld overlay materials has proven to be a very effective method to extend the life of boiler tubes in coal-fired boilers. In order to properly select the best material as a weld overlay, the demands placed upon the material need to be understood. The required material properties for a weld overlay can change with boiler design (such as once-through versus drum boilers), tube function (evaporator versus superheater tubes), and for various regions within the boiler (such as elevation or proximity to over-fire air ports). Material properties that need to be considered include the physical, mechanical, and corrosion properties, as well as the cost and ease of fabrication. The score card in some of these areas has been less than desired for certain weld overlay materials. Nonetheless, there are usually multiple material choices that will work in most cases. When multiple choices are available, asset owners will generally select the material with the most experience. Since its introduction in 2003, Alloy 33 has continued to gain positive experience as a weld overlay in coal-fired boilers and is therefore gaining acceptance within the industry. In the furnace region where combustion occurs, the waterwall tubes are exposed to high heat inputs along with corrosive combustion gases and deposits. These conditions can cause rapid corrosion by a mixed sulfidation/oxidation mechanism. The corrosion rates increase further if low NOx combustion practices are used, since this causes a reducing atmosphere that forms more corrosive sulfur species such as H{sub 2}S gases and FeS deposits. The corrosion rates increase with tube metal temperatures, which are controlled by the local tube pressure as well as the operating practices (i.e. heat flux rates). In the highest pressure units that operate above the water triple point (supercritical plants) cracking can sometimes also be an issue. This cracking is caused by a corrosion fatigue mechanism and is

  4. 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.)

  5. 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

  6. 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.

  7. 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

  8. 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.

  9. 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)

  10. Irradiation Test of Dual Instrumented Fuel Rods by using an Instrumented Fuel Capsule(05F-01K) at HANARO

    International Nuclear Information System (INIS)

    Sohn, Jaemin; Park, Sungjae; Shin, Yoontaeg; Lee, Choongsung; Choo, Keenam; Cho, Mansoon; Oh, Jongmyung; Kim, Bonggoo; Kim, Harkrho

    2007-01-01

    The purpose of this paper is to verify the performance of dual instrumented fuel rods. The dual instrumented fuel rods, which allow for two characteristics to be measured simultaneously in one fuel rod, have been designed to enhance the efficiency of an irradiation test using an instrumented capsule for the nuclear fuel irradiation test(hereinafter referred to as 'instrumented fuel capsule') in HANARO(High-flux Advanced Neutron Application Reactor). Six types of dual instrumented fuel rods have been designed. The types of dual instrumented fuel rods are summarized as follows; 1) to measure the center temperature of the nuclear fuel and the internal pressure of the fuel rod, 2) to measure the center temperature of the nuclear fuel and the elongation of the fuel pellets, 3) to measure the surface temperature of the nuclear fuel and the internal pressure of the fuel rod, 4) to measure the surface temperature of the nuclear fuel and the elongation of the fuel pellets, 5) to measure the center and surface temperature of the nuclear fuel, and 6) to measure the center temperature of the nuclear fuel of the upper and lower part. And 05F-01K instrumented fuel capsule has been designed for an irradiation test of three dual instrumented fuel rods. This paper presents the manufacturing of the dual instrumented fuel rods and 05F-01K instrumented fuel capsule, and the results of the irradiation test

  11. Test Record of Flight Tests Using Alcohol-to-Jet/JP-8 Blended Fuel

    Science.gov (United States)

    2015-09-01

    SPECIAL REPORT RDMR-AE-15-02 TEST RECORD OF FLIGHT TESTS USING ALCOHOL -TO-JET/JP-8 BLENDED FUEL Franklin D Cox and George...2015 3. REPORT TYPE AND DATES COVERED Final 4. TITLE AND SUBTITLE Test Record of Flight Tests Using Alcohol -to-Jet/JP-8 Blended Fuel 5...be certified for use in these air vehicles. 14. SUBJECT TERMS UH-60, CH-47, alternative fuel , Alcohol -to-Jet, ATJ blend, flight test, longevity

  12. Italian progress on LWR fuel design, manufacturing, testing and management

    International Nuclear Information System (INIS)

    Arcelli, G.; Bozzoni, T.; Coluccio, G.; Gatti, S.; Gerevini, T.; Testa, G.

    1977-01-01

    This paper summarizes the main activities carried out in Italy from the last Geneva Conference (1971) in the field of LWR fuel design, manufacturing, testing and management. In fuel design, large effort has been devoted both to the assimilation of the information resulting from the licensing agreement with the US suppliers and to the development of an independent know-how through adequate R and D programs as well as to fuel behavior modeling activities. Among these it is worthwhile to mention the studies carried out by CNEN on the in-pile behavior of gadolinia fuel rods, the critical heat flux (CHF) and void fraction experiments, the overpower ramp tests on pre-irradiated fuel rods, the measurements of the pellet-cladding thermal conductance vs. burn-up and the investigation on fission gas release kinetics. The experience acquired with the fuel of the two LWR ENEL power stations, Garigliano BWR and Trino PWR, and some fuel management problems are reviewed. In this framework mention is made to the effort devoted to the acquisition of a wider knowledge on fuel behavior through post irradiation examination (PIE) of power reactor fuel elements as well as of experimental rods. The Italian fuel manufacturing capabilities are illustrated: they consist of the FN (Fabbricazioni Nucleari) plant at Bosco Marengo, owned by AGIP Nucleare and AMN, which has produced up-to-now a reload for Garigliano BWR and the first core for Caorso BWR and of COREN (COmbustibili REattori Nucleari) at Saluggia, owned by Breda, FIAT and Westinghouse, which has fabricated reloads for Trino PWR, including 8 UO 2 -PuO 2 fuel bundles, while fuel loading for the Otto Hahn nuclear ship and fuel mechanical components for some PWRs have been fabricated by FIAT [fr

  13. 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

  14. 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

  15. 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

  16. Reactivity feedbacks of a material test research reactor fueled with various low enriched uranium dispersion fuels

    International Nuclear Information System (INIS)

    Muhammad, Farhan; Majid, Asad

    2009-01-01

    The reactivity feedbacks of a material test research reactor using various low enriched uranium fuels, having same uranium density were calculated. For this purpose, the original aluminide fuel (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -Al) dispersion fuels having the same uranium density as of the original fuel. Calculations were carried out to find the fuel temperature reactivity feedback, moderator temperature reactivity feedback, moderator density reactivity feedback and moderator void reactivity feedback. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitudes all the respective reactivity feedbacks from 38 deg. C to 50 deg. C and 100 deg. C, at the beginning of life, of all the fuels were very close to each other. The fuel temperature reactivity feedback of the U 3 O 8 -Al was about 2% more than the original UAl x -Al fuel. The magnitudes of the moderator temperature, moderator density and moderator void reactivity feedbacks of all the fuels, showed very minor variations from the original aluminide fuel.

  17. 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

  18. In-core fuel motion detection for large scale tests

    International Nuclear Information System (INIS)

    Wright, S.A.; Dupree, S.A.

    1976-01-01

    An experimental program to test the feasibility of making and unfolding in-core measurements of fuel motion has been initiated at Sandia. The experiment calls for instrumenting a 37-pin fuel bundle and exposing it to neutron pulses in the Sandia Pulsed Reactor SPR-III. Fuel motion will be simulated by mechanically adjusting the fuel pins to create voids within the center of the bundle. A preliminary design of the experiment has been completed and analyses have been conducted on the application of 238 U and 235 U fission couple detectors to the experiment. Formulation of the unfolding problem has pointed out the necessity of carefully instrumenting the experiment to maximize the resolution of the mass measurement. Ultimately a measurement resolution of 100-200 g of fuel should be possible for this experiment. The experiment will provide a realistic test bed for obtaining practical experience in detector design and data reduction

  19. 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

  20. 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

  1. Simulation and Test of a Fuel Cell Hybrid Golf Cart

    Directory of Open Access Journals (Sweden)

    Jingming Liang

    2014-01-01

    Full Text Available This paper establishes the simulation model of fuel cell hybrid golf cart (FCHGC, which applies the non-GUI mode of the Advanced Vehicle Simulator (ADVISOR and the genetic algorithm (GA to optimize it. Simulation of the objective function is composed of fuel consumption and vehicle dynamic performance; the variables are the fuel cell stack power sizes and the battery numbers. By means of simulation, the optimal parameters of vehicle power unit, fuel cell stack, and battery pack are worked out. On this basis, GUI mode of ADVISOR is used to select the rated power of vehicle motor. In line with simulation parameters, an electrical golf cart is refitted by adding a 2 kW hydrogen air proton exchange membrane fuel cell (PEMFC stack system and test the FCHGC. The result shows that the simulation data is effective but it needs improving compared with that of the real cart test.

  2. Analysis and Testing of JP-5 Fuel Derived from Coal

    Science.gov (United States)

    1977-01-01

    Air propulsion T’,;,t Center (NAPTC) is investigating alternate sources ( coal , shale and tar sandt;) that could be used to ! roduce J1-5. This work is... coal derived JP-5 type fuel tested is a hydrotreated kerosene that had been fractionated from a Western Kentucky syncrude (Sample 0005). 5. Method of...IeeCko*wInbW) Evaluation Water Coalescers Soluble Copper Coal Derived JP-5 Fuel Storage/Thermal Stability S Engine Performiance Specification Tests

  3. Failed MTR Fuel Element Detect in a Sipping Tests

    International Nuclear Information System (INIS)

    Zeituni, C.A.; Terremoto, L.A.A.; Silva, J.E.R. da

    2004-01-01

    This work describes sipping tests performed on Material Testing Reactor (MTR) fuel elements of the IEA-R1 research reactor, in order to find out which one failed in the core during a routine operation. Radioactive iodine isotopes 131 I and 133 I, employed as failure monitors, were detected in samples corresponding to the failed fuel element. The specific activity of each sample, as well as the average leaking rate, were measured for 137 Cs. The nuclear fuels U 3 O 8 - Al dispersion and U - Al alloy were compared concerning their measured average leaking rates of 137 Cs

  4. Failed MTR Fuel Element Detect in a Sipping Tests

    Energy Technology Data Exchange (ETDEWEB)

    Zeituni, C.A.; Terremoto, L.A.A.; da Silva, J.E.R.

    2004-10-06

    This work describes sipping tests performed on Material Testing Reactor (MTR) fuel elements of the IEA-R1 research reactor, in order to find out which one failed in the core during a routine operation. Radioactive iodine isotopes {sup 131}I and {sup 133}I, employed as failure monitors, were detected in samples corresponding to the failed fuel element. The specific activity of each sample, as well as the average leaking rate, were measured for {sup 137}Cs. The nuclear fuels U{sub 3}O{sub 8} - Al dispersion and U - Al alloy were compared concerning their measured average leaking rates of {sup 137}Cs.

  5. Sipping tests on a failed irradiated MTR fuel element

    International Nuclear Information System (INIS)

    Zeituni, C. A.; Terremoto, L. A. A.; Da Silva, J. E. R.

    2004-01-01

    This work describes sipping tests performed on Material Testing Reactor (MTR) fuel elements of the IEA-R1 research reactor, in order to find out which one failed in the core during a routine operation. Radioactive iodine isotopes 131 I and 133 I, employed as failure monitors, were detected in samples corresponding to the failed fuel element. The specific activity of each sample, as well as the average leaking rate, were measured for 137 Cs. The nuclear fuels U 3 O 8 - Al dispersion and U - Al alloy were compared concerning their measured average leaking rates of 137 Cs. (authors)

  6. 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

  7. Results of industrial tests of carbonate additive to fuel oil

    Science.gov (United States)

    Zvereva, E. R.; Dmitriev, A. V.; Shageev, M. F.; Akhmetvalieva, G. R.

    2017-08-01

    Fuel oil plays an important role in the energy balance of our country. The quality of fuel oil significantly affects the conditions of its transport, storage, and combustion; release of contaminants to atmosphere; and the operation of main and auxiliary facilities of HPPs. According to the Energy Strategy of Russia for the Period until 2030, the oil-refining ratio gradually increases; as a result, the fraction of straight-run fuel oil in heavy fuel oils consistently decreases, which leads to the worsening of performance characteristics of fuel oil. Consequently, the problem of the increase in the quality of residual fuel oil is quite topical. In this paper, it is suggested to treat fuel oil by additives during its combustion, which would provide the improvement of ecological and economic indicators of oil-fired HPPs. Advantages of this method include simplicity of implementation, low energy and capital expenses, and the possibility to use production waste as additives. In the paper, the results are presented of industrial tests of the combustion of fuel oil with the additive of dewatered carbonate sludge, which is formed during coagulation and lime treatment of environmental waters on HPPs. The design of a volume delivery device is developed for the steady additive input to the boiler air duct. The values are given for the main parameters of the condition of a TGM-84B boiler plant. The mechanism of action of dewatered carbonate sludge on sulfur oxides, which are formed during fuel oil combustion, is considered. Results of industrial tests indicate the decrease in the mass fraction of discharged sulfur oxides by 36.5%. Evaluation of the prevented damage from sulfur oxide discharged into atmospheric air shows that the combustion of the fuel oil of 100 brand using carbonate sludge as an additive (0.1 wt %) saves nearly 6 million rubles a year during environmental actions at the consumption of fuel oil of 138240 t/year.

  8. 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.

  9. 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.

  10. 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

  11. 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)

  12. 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

  13. 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

  14. 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)

  15. Fabrication and performance testing of CANDU mixed-oxide fuel

    International Nuclear Information System (INIS)

    Dimayuga, F.C.; Floyd, M.R.; Cox, D.S.

    2000-01-01

    AECL's mixed-oxide fuel fabrication activities are performed in the Recycle Fuel Fabrication Laboratories (RFFL) at the Chalk River Laboratories. Since the start-up of the RFFL in the mid-1970s, several fabrication campaigns have been conducted in the facility, producing various types of mixed-oxide (MOX) fuel, which were used for both irradiation and physics testing. More recently, CANDU fuel bundles containing 0.5 w-t % plutonium in natural uranium, produced in the RFFL, were successfully irradiated in the NRU reactor at powers up to 65 kW/m and to burnups ranging from 13 to 23 MW·d/kg HE. Two of the bundles had power histories that bound the normal powers and burnups of natural UO 2 CANDU fuel ( 2 fuel. Significantly more grain growth was observed than that expected for UO 2 fuel; however, this increase in grain growth had no effect on the overall performance of the fuel. Two other bundles operated to extended burnups of 19 to 23 MW·d/kg HE. Burnup extension above 15 MW·d/kg HE only had a small effect on FGR. (author)

  16. Design verification testing for fuel element type CAREM

    International Nuclear Information System (INIS)

    Martin Ghiselli, A.; Bonifacio Pulido, K.; Villabrille, G.; Rozembaum, I.

    2013-01-01

    The hydraulic and hydrodynamic characterization tests are part of the design verification process of a nuclear fuel element prototype and its components. These tests are performed in a low pressure and temperature facility. The tests requires the definition of the simulation parameters for setting the test conditions, the results evaluation to feedback mathematical models, extrapolated the results to reactor conditions and finally to decide the acceptability of the tested prototype. (author)

  17. 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.

  18. Spent LWR fuel leach tests: Waste Isolation Safety Assessment program

    International Nuclear Information System (INIS)

    Katayama, Y.B.

    1979-04-01

    Spent light-water-reactor (LWR) fuels with burnups of 54.5, 28 and 9 MWd/kgU were leach-tested in deionized water at 25 0 C. Fuel burnup has no apparent effect on the calculated leach rates based upon the behavior of 137 Cs and 239+240 Pu. A leach test of 54.5 MWd/kgU spent fuel in synthetic sea brine showed that the cesium-based leach rate is lower in sea brine than in deionized water. A rise in the leach rate was observed after approximately 600 d of cumulative leaching. During the rise, the leach rate for all the measured radionuclides become nearly equal. Evidence suggests that exposure of new surfaces to the leachant may cause the increase. As a result, experimental work to study leaching mechanisms of spent fuel has been initiated. 22 figures

  19. Sipping test on a failed MTR fuel element

    International Nuclear Information System (INIS)

    Terremoto, Luis Antonio Albiac; Zeituni, Carlos Alberto; Silva, Antonio Teixeira e; Perrotta, Jose Augusto; Silva, Jose Eduardo Rosa da

    2002-01-01

    This work describes sipping tests performed on MTR fuel elements of the IEA-R1 research reactor, in order to determinate which one failed in the core during a routine operation of the reactor. radioactive iodine isotopes 131 I and 133 I, employed as failure indicators, were detected in samples corresponding to the fuel element IEA-156. The specific activity of each sample, as well as the average leaking rate, were measured for 137 Cs. The nuclear fuels U 3 O 8 - Al dispersion and U - Al alloy were compared concerning their measured average leaking rates of 137 Cs. (author)

  20. 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

  1. 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

  2. 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.

  3. Testing of a transport cask for research reactor spent fuel

    International Nuclear Information System (INIS)

    Mourao, Rogerio P.; Silva, Luiz Leite da; Miranda, Carlos A.; Mattar Neto, Miguel; Quintana, Jose F.A.; Saliba, Roberto O.; Novara, Oscar E.

    2011-01-01

    Since the beginning of the last decade three Latin American countries which operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the reactors operated in the region. As a step in this direction, a packaging for the transport of irradiated fuel from research reactors was designed by a tri-national team and a half-scale model for MTR fuel constructed in Argentina and tested in Brazil. Two test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. Although the specimen has not successfully performed the tests, its overall performance was considered very satisfactory, and improvements are being introduced to the design. A third test sequence is planned for 2011. (author)

  4. Modeling and Simulation of a Nuclear Fuel Element Test Section

    Science.gov (United States)

    Moran, Robert P.; Emrich, William

    2011-01-01

    "The Nuclear Thermal Rocket Element Environmental Simulator" test section closely simulates the internal operating conditions of a thermal nuclear rocket. The purpose of testing is to determine the ideal fuel rod characteristics for optimum thermal heat transfer to their hydrogen cooling/working fluid while still maintaining fuel rod structural integrity. Working fluid exhaust temperatures of up to 5,000 degrees Fahrenheit can be encountered. The exhaust gas is rendered inert and massively reduced in temperature for analysis using a combination of water cooling channels and cool N2 gas injectors in the H2-N2 mixer portion of the test section. An extensive thermal fluid analysis was performed in support of the engineering design of the H2-N2 mixer in order to determine the maximum "mass flow rate"-"operating temperature" curve of the fuel elements hydrogen exhaust gas based on the test facilities available cooling N2 mass flow rate as the limiting factor.

  5. Mixed oxide fuels testing in the advanced test reactor to support plutonium disposition

    International Nuclear Information System (INIS)

    Ryskamp, J.M.; Sterbentz, J.W.; Chang, G.S.

    1995-09-01

    An intense worldwide effort is now under way to find means of reducing the stockpile of weapons-grade plutonium. One of the most attractive solutions would be to use WGPu as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PUO 2 ) mixed with urania (UO 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. 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. 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

  6. 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 la...

  7. Design of the dual instrumented fuel rods to measure the nuclear fuel characteristics during Irradiation test at HANARO

    International Nuclear Information System (INIS)

    Sohn, Jae Min; Oh, Jong Myung; Cho, Man Soon; Choo, Ki nam; Choi, Myung Hwan; Lee, Dong Soo; Kim, Boong Goo; Kim, Young Jin

    2005-01-01

    The instrumented capsule for the nuclear fuel irradiation test (hereinafter referred to instrumented fuel capsule), which are crucial for the verification of a nuclear fuel performance and safety, have been developed at HANARO(High-flux Advanced Neutron Application Reactor). The irradiation test of the first instrumented fuel capsule(02F-11K) was carried out in March 2003 for 1,296 MWD(Mega Watt Day) and the irradiation test of the second instrumented fuel capsule(03F-05K) was carried out in April 2004 for 1,533MWD at HANARO. Through the irradiation tests of the two capsules, the design specifications and safety of the instrumented fuel capsule were verified successfully. In the 02F-11K instrumented fuel capsule, only the technologies for measuring the center temperature of the nuclear fuel and neutron flux were implemented. In the 03F-05K instrumented fuel capsule, the technologies for measuring the center temperature of the nuclear fuel, the internal pressure of the fuel rod, the elongation of the nuclear fuel and the neutron flux were implemented. The purpose of this paper is to develop the dual instrumented technology that enables two characteristics to be measured simultaneously in one fuel rod. Therefore, this paper presents the design of the dual instrumented fuel rods and the plan of the irradiation test for the newly designed fuel rods

  8. 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.

  9. 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.

  10. Mechanical Fatigue Testing of High Burnup Fuel for Transportation Applications

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-05-01

    This report describes testing designed to determine the ability of high burnup (HBU) (>45 GWd/MTU) spent fuel to maintain its integrity under normal conditions of transportation. An innovative system, Cyclic Integrated Reversible-bending Fatigue Tester (CIRFT), has been developed at Oak Ridge National Laboratory (ORNL) to test and evaluate the mechanical behavior of spent nuclear fuel (SNF) under conditions relevant to storage and transportation. The CIRFT system is composed of a U-frame equipped with load cells for imposing the pure bending loads on the SNF rod test specimen and measuring the in-situ curvature of the fuel rod during bending using a set up with three linear variable differential transformers (LVDTs).

  11. Thermal test and analysis of a spent fuel storage cask

    International Nuclear Information System (INIS)

    Yamakawa, H.; Gomi, Y.; Ozaki, S.; Kosaki, A.

    1993-01-01

    A thermal test simulated with full-scale cask model for the normal storage was performed to verify the storage skill of the spent fuels of the cask. The maximum temperature at each point in the test was lower than the allowable temperature. The integrity of the cask was maintained. It was observed that the safety of containment system was also kept according to the check of the seal before and after the thermal test. Therefore it was shown that using the present skill, it is possible to store spent fuels in the dry-type cask safely. Moreover, because of the good agreement between analysis and experimental results, it was shown that the analysis model was successfully established to estimate the temperature distribution of the fuel cladding and the seal portion. (J.P.N.)

  12. 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)

  13. 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. LWRS Fuels Pathway: Engineering Design and Fuels Pathway Initial Testing of the Hot Water Corrosion System

    Energy Technology Data Exchange (ETDEWEB)

    Dr. John Garnier; Dr. Kevin McHugh

    2012-09-01

    The Advanced LWR Nuclear Fuel Development R&D pathway performs strategic research focused on cladding designs leading to improved reactor core economics and safety margins. The research performed is to demonstrate the nuclear fuel technology advancements while satisfying safety and regulatory limits. These goals are met through rigorous testing and analysis. The nuclear fuel technology developed will assist in moving existing nuclear fuel technology to an improved level that would not be practical by industry acting independently. Strategic mission goals are to improve the scientific knowledge basis for understanding and predicting fundamental nuclear fuel and cladding performance in nuclear power plants, and to apply this information in the development of high-performance, high burn-up fuels. These will result in improved safety, cladding, integrity, and nuclear fuel cycle economics. To achieve these goals various methods for non-irradiated characterization testing of advanced cladding systems are needed. One such new test system is the Hot Water Corrosion System (HWCS) designed to develop new data for cladding performance assessment and material behavior under simulated off-normal reactor conditions. The HWCS is capable of exposing prototype rodlets to heated, high velocity water at elevated pressure for long periods of time (days, weeks, months). Water chemistry (dissolved oxygen, conductivity and pH) is continuously monitored. In addition, internal rodlet heaters inserted into cladding tubes are used to evaluate repeated thermal stressing and heat transfer characteristics of the prototype rodlets. In summary, the HWCS provides rapid ex-reactor evaluation of cladding designs in normal (flowing hot water) and off-normal (induced cladding stress), enabling engineering and manufacturing improvements to cladding designs before initiation of the more expensive and time consuming in-reactor irradiation testing.

  15. 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.

  16. Generator Set Durability Testing Using 25% ATJ Fuel Blend

    Science.gov (United States)

    2016-02-01

    additional check, the crankcase blow-by filter was replaced at the 16 hour mark, and checked for excessive oil at the end of test. No excess oil was found on...blended fuel in a 30kW AMMPS Generator utilizing a Cummins 3.3L QSB engine . In general, fuel compatibility within the bounds of the presented tests...seemed acceptable since there was no baseline data available to compare. The largest differences in engine operation may be attributed to the extreme

  17. Postirradiation examination of Peach Bottom fuel test element FTE-4

    International Nuclear Information System (INIS)

    Wallroth, C.F.; Holzgraf, J.F.; Jensen, D.D.; Zumwalt, L.R.

    1977-07-01

    The report presents the irradiation results and their evaluation for Peach Bottom fuel test element FTE-4. It describes in detail the efforts by General Atomic Company over the last two years to establish a system for extracting meaningful performance information from a fuel test element. This has been done with the goal of making direct comparisons between as-measured data and core design code predictions. Special emphasis has been placed on determining the 95% confidence limits on most of the preirradiation and postirradiation measurements in order to allow a better comparison with GAUGE, FEVER, and TREVER code calculations which are used in HTGR core thermal and mechanical design

  18. 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... Regulations for New Gasoline-Fueled and Diesel-Fueled Heavy-Duty Engines; Gaseous Exhaust Test Procedures § 86.335-79 Gasoline-fueled engine test cycle. (a) The following test sequence shall be followed in...

  19. Real life testing of a Hybrid PEM Fuel Cell Bus

    Science.gov (United States)

    Folkesson, Anders; Andersson, Christian; Alvfors, Per; Alaküla, Mats; Overgaard, Lars

    Fuel cells produce low quantities of local emissions, if any, and are therefore one of the most promising alternatives to internal combustion engines as the main power source in future vehicles. It is likely that urban buses will be among the first commercial applications for fuel cells in vehicles. This is due to the fact that urban buses are highly visible for the public, they contribute significantly to air pollution in urban areas, they have small limitations in weight and volume and fuelling is handled via a centralised infrastructure. Results and experiences from real life measurements of energy flows in a Scania Hybrid PEM Fuel Cell Concept Bus are presented in this paper. The tests consist of measurements during several standard duty cycles. The efficiency of the fuel cell system and of the complete vehicle are presented and discussed. The net efficiency of the fuel cell system was approximately 40% and the fuel consumption of the concept bus is between 42 and 48% lower compared to a standard Scania bus. Energy recovery by regenerative braking saves up 28% energy. Bus subsystems such as the pneumatic system for door opening, suspension and brakes, the hydraulic power steering, the 24 V grid, the water pump and the cooling fans consume approximately 7% of the energy in the fuel input or 17% of the net power output from the fuel cell system. The bus was built by a number of companies in a project partly financed by the European Commission's Joule programme. The comprehensive testing is partly financed by the Swedish programme "Den Gröna Bilen" (The Green Car). A 50 kW el fuel cell system is the power source and a high voltage battery pack works as an energy buffer and power booster. The fuel, compressed hydrogen, is stored in two high-pressure stainless steel vessels mounted on the roof of the bus. The bus has a series hybrid electric driveline with wheel hub motors with a maximum power of 100 kW. Hybrid Fuel Cell Buses have a big potential, but there are

  20. Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, C.; Wachs, D.; Carmack, J.; Woolstenhulme, N.

    2017-01-01

    The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, and salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.

  1. 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)

  2. 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.)

  3. 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

  4. 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

  5. Modelling of LOCA Tests with the BISON Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-01

    BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculations are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.

  6. Integrity confirmation tests and post-irradiation test plan of the HTTR first-loading fuel

    International Nuclear Information System (INIS)

    Sawa, Kazuhiro; Sumita, Junya; Ueta, Shouhei; Suzuki, Shuichi; Tobita, Tsutomu; Saito, Takashi; Minato, Kazuo; Koya, Toshio; Sekino, Hajime

    2001-01-01

    Since the first-loading fuel of the High Temperature Engineering Test Reactor (HTTR) is the first mass-production High Temperature Gas-cooled Reactor (HTGR) fuel in Japan, their quality should be carefully inspected. For the quality control related to the fabrication process, Japan Atomic Energy Research Institute (JAERI) carried out the tests to certify the fuel integrity during operation. The tests comprise (1) as-fabricated SiC failure fraction measurement, (2) high-temperature heatup test of irradiated fuel and (3) accelerated irradiation test. For (1), the SiC failure fraction was measured independently in JAERI in addition to the measurement in the fabrication process. The measure failure fractions agreed within 95% confidence limit. In order to confirm the integrity of the SiC layer with respect to the 1,600degC criterion, the high-temperature heatup test of irradiated fuel compact was carried out. The results showed that no failed particle was present in the fuel compact after hating. The diffusion coefficient of metallic fission products in SiC layer was also examined in a series of post-irradiation heating tests. The measure diffusion coefficient of 137 Cs showed a good holding ability as those obtained for research and development fuel specimen. The measured fission gas release rate in accelerated irradiation test showed no additional failure up to 60 GWd/t which was about two times higher than 3 GWd/t of the maximum burnup in the HTTR core. Through the tests, integrity of as-fabricated first-loading fuel of the HTTR was finally confirmed. The future post-irradiation test plan, which will be carried out to confirm the fuel irradiation performance and to obtain the data on its irradiation characteristics in the core, is also described. (author)

  7. Induction Heating Model of Cermet Fuel Element Environmental Test (CFEET)

    Science.gov (United States)

    Gomez, Carlos F.; Bradley, D. E.; Cavender, D. P.; Mireles, O. R.; Hickman, R. R.; Trent, D.; Stewart, E.

    2013-01-01

    Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames. Nuclear Thermal Rockets (NTR) are capable of producing a high specific impulse by employing heat produced by a fission reactor to heat and therefore accelerate hydrogen through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements due to large thermal gradients; therefore, high-melting-point ceramics-metallic matrix composites (cermets) are one of the fuels under consideration as part of the Nuclear Cryogenic Propulsion Stage (NCPS) Advance Exploration System (AES) technology project at the Marshall Space Flight Center. The purpose of testing and analytical modeling is to determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures and obtain data to assess the properties of the non-nuclear support materials. The fission process and the resulting heating performance are well known and do not require that active fissile material to be integrated in this testing. A small-scale test bed; Compact Fuel Element Environmental Tester (CFEET), designed to heat fuel element samples via induction heating and expose samples to hydrogen is being developed at MSFC to assist in optimal material and manufacturing process selection without utilizing fissile material. This paper details the analytical approach to help design and optimize the test bed using COMSOL Multiphysics for predicting thermal gradients induced by electromagnetic heating (Induction heating) and Thermal Desktop for radiation calculations.

  8. 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

  9. 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

  10. 40 CFR 1065.701 - General requirements for test fuels.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false General requirements for test fuels. 1065.701 Section 1065.701 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR... procedure 1 Light distillate and light blends with residual ASTM D975-07b. Diesel Middle distillate ASTM...

  11. TEST RESULTS FOR FUEL CELL OPERATION ON ANAEROBIC DIGESTER GAS

    Science.gov (United States)

    EPA, in conjunction with ONSI Corp., embarked on a project to define, design, test, and assess a fuel cell energy recovery system for application at anaerobic digester waste water (sewage) treatment plants. Anaerobic digester gas (ADG) is produced at these plants during the proce...

  12. 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...

  13. 40 CFR 1060.521 - How do I test fuel caps for permeation emissions?

    Science.gov (United States)

    2010-07-01

    ... tank and fuel cap meet emission standards by certifying them separately or by combining the separate... 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...

  14. 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

  15. Calibration Tests of Fuel Assembly Simulators of APR+ Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kih Wan; Chu, In Cheol; Euh, Dong Jin; Kwon, Tae Soon [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    A Reactor flow distribution is regarded to be major importance in improving the design margin of a flow distribution. The prediction of APR+ core fluid flow phenomena has been in demand, since 257 fuel assemblies are adapted in the APR+, unlike in the APR1400. The APR+ reactor flow test facility, the ACOP (APR+ Core Flow and Pressure Test Facility), was constructed to analyze the hydraulic characteristics. For the ACOP facility, the core simulator was designed with a scale analysis to simulate the real HIPER fuel assembly of an APR+. In this study, for all 257 core simulators, several calibration tests were conducted to verify their design performance before applying them to the ACOP facility. The inlet flow rate and the total pressure drop of the simulators were measured by varying flow rates to evaluate its compatibility. The discharge coefficients were also calculated from the experimental data to produce a statistical database for a further ACOP facility test.

  16. Testing of reactor fuel materials using nuclear techniques

    International Nuclear Information System (INIS)

    Khouri, M.T.F.C.

    1978-01-01

    The tests presented here apply to: the quantitative determination of uranium in the core of fuel element plates by the detection of the number of neutrons produced in photo induced reactions in uranium; the determination of 235 U proportion in uranium dioxide samples, in the form of uranyl nitrate, by the technique of the detection of tracks produced by fission fragments and in pellet samples by passive gamma spectrometry and the checking of uranium homogenization distribution in fuel plates and uranium dioxide pellets. (Author) [pt

  17. Used fuel rail shock and vibration testing options analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ross, Steven B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Best, Ralph E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Klymyshyn, Nicholas A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jensen, Philip J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maheras, Steven J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2014-09-25

    The objective of the rail shock and vibration tests is to complete the framework needed to quantify loads of fuel assembly components that are necessary to guide materials research and establish a technical basis for review organizations such as the U.S. Nuclear Regulatory Commission (NRC). A significant body of experimental and numerical modeling data exists to quantify loads and failure limits applicable to normal conditions of transport (NCT) rail transport, but the data are based on assumptions that can only be verified through experimental testing. The test options presented in this report represent possible paths for acquiring the data that are needed to confirm the assumptions of previous work, validate modeling methods that will be needed for evaluating transported fuel on a case-by-case basis, and inform material test campaigns on the anticipated range of fuel loading. The ultimate goal of this testing is to close all of the existing knowledge gaps related to the loading of used fuel under NCT conditions and inform the experiments and analysis program on specific endpoints for their research. The options include tests that would use an actual railcar, surrogate assemblies, and real or simulated rail transportation casks. The railcar carrying the cradle, cask, and surrogate fuel assembly payload would be moved in a train operating over rail track modified or selected to impart shock and vibration forces that occur during normal rail transportation. Computer modeling would be used to help design surrogates that may be needed for a rail cask, a cask’s internal basket, and a transport cradle. The objective of the design of surrogate components would be to provide a test platform that effectively simulates responses to rail shock and vibration loads that would be exhibited by state-of-the-art rail cask, basket, and/or cradle structures. The computer models would also be used to help determine the placement of instrumentation (accelerometers and strain gauges

  18. 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.

  19. 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)

  20. 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

  1. 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.

  2. Thermomechanical modeling of the Spent Fuel Test-Climax

    Energy Technology Data Exchange (ETDEWEB)

    Butkovich, T.R.; Patrick, W.C.

    1986-02-01

    The Spent Fuel Test-Climax (SFT-C) was conducted to evaluate the feasibility of retrievable deep geologic storage of commercially generated spent nuclear-reactor fuel assemblies. One of the primary aspects of the test was to measure the thermomechanical response of the rock mass to the extensive heating of a large volume of rock. Instrumentation was emplaced to measure stress changes, relative motion of the rock mass, and tunnel closures during three years of heating from thermally decaying heat sources, followed by a six-month cooldown period. The calculations reported here were performed using the best available input parameters, thermal and mechanical properties, and power levels which were directly measured or inferred from measurements made during the test. This report documents the results of these calculations and compares the results with selected measurements made during heating and cooling of the SFT-C.

  3. Thermomechanical modeling of the Spent Fuel Test-Climax

    International Nuclear Information System (INIS)

    Butkovich, T.R.; Patrick, W.C.

    1986-02-01

    The Spent Fuel Test-Climax (SFT-C) was conducted to evaluate the feasibility of retrievable deep geologic storage of commercially generated spent nuclear-reactor fuel assemblies. One of the primary aspects of the test was to measure the thermomechanical response of the rock mass to the extensive heating of a large volume of rock. Instrumentation was emplaced to measure stress changes, relative motion of the rock mass, and tunnel closures during three years of heating from thermally decaying heat sources, followed by a six-month cooldown period. The calculations reported here were performed using the best available input parameters, thermal and mechanical properties, and power levels which were directly measured or inferred from measurements made during the test. This report documents the results of these calculations and compares the results with selected measurements made during heating and cooling of the SFT-C

  4. A new method for estimating heat flux in superheater and reheater tubes

    Energy Technology Data Exchange (ETDEWEB)

    Purbolaksono, J. [Department of Mechanical Engineering, Universiti Tenaga Nasional, km 7 Jalan Kajang-Puchong, Kajang 43009, Selangor (Malaysia)], E-mail: judha@uniten.edu.my; Khinani, A.; Rashid, A.Z.; Ali, A.A. [Department of Mechanical Engineering, Universiti Tenaga Nasional, km 7 Jalan Kajang-Puchong, Kajang 43009, Selangor (Malaysia); Ahmad, J. [Kapar Energy Ventures Sdn Bhd, Jalan Tok Muda, Kapar 42200, Selangor (Malaysia); Nordin, N.F. [TNB Research Sdn Bhd, No. 1 Lorong Air Hitam, Kajang 43000, Selangor (Malaysia)

    2009-10-15

    In this paper a procedure on how to estimate the heat flux in superheater and reheater tubes utilizing the empirical formula and the finite element modeling is proposed. An iterative procedure consisting of empirical formulae and numerical simulation is used to determine heat flux as both temperature and scale thickness increase over period of time. Estimation results of the heat flux over period of time for two different design temperatures of the steam and different heat transfer parameters are presented.

  5. Structure and properties of high-temperature austenitic steels for superheater tubes

    Science.gov (United States)

    Blinov, V. M.

    2009-12-01

    The structure and properties of high-temperature austenitic steels intended for superheater tubes are analyzed. Widely used Kh18N10T (AISI 304) and Kh16N13M3 (AISI 316) steels are found not to ensure a stable austenitic structure and stable properties during long-term thermal holding under stresses. The hardening of austenitic steels by fine particles of vanadium and niobium carbides and nitrides and γ'-phase and Fe2W and Fe2Mo Laves phase intermetallics is considered. The role of Cr23C6 chromium carbides, the σ phase, and coarse precipitates of an M 3B2 phase and a boron-containing eutectic in decreasing the time to failure and the stress-rupture strength of austenitic steels is established. The mechanism of increasing the stress-rupture strength of steels by boron additions is described. The chemical compositions, mechanical properties, stress-rupture strength, and creep characteristics of Russian and foreign austenitic steels used or designed for superheater tubes intended for operation under stress conditions at temperatures above 600°C are presented. The conditions are found for increasing the strength, plasticity, and thermodeformation stability of austenite in steels intended for superheater tubes operating at 700°C under high stresses for a long time.

  6. 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.

  7. Improved Accelerated Stress Tests Based on Fuel Cell Vehicle Data

    Energy Technology Data Exchange (ETDEWEB)

    Patterson, Timothy [Research Engineer; Motupally, Sathya [Research Engineer

    2012-06-01

    UTC will led a top-tier team of industry and national laboratory participants to update and improve DOE’s Accelerated Stress Tests (AST’s) for hydrogen fuel cells. This in-depth investigation will focused on critical fuel cell components (e.g. membrane electrode assemblies - MEA) whose durability represented barriers for widespread commercialization of hydrogen fuel cell technology. UTC had access to MEA materials that had accrued significant load time under real-world conditions in PureMotion® 120 power plant used in transit buses. These materials are referred to as end-of-life (EOL) components in the rest of this document. Advanced characterization techniques were used to evaluate degradation mode progress using these critical cell components extracted from both bus power plants and corresponding materials tested using the DOE AST’s. These techniques were applied to samples at beginning-of-life (BOL) to serve as a baseline. These comparisons advised the progress of the various failure modes that these critical components were subjected to, such as membrane degradation, catalyst support corrosion, platinum group metal dissolution, and others. Gaps in the existing ASTs predicted the degradation observed in the field in terms of these modes were outlined. Using the gaps, new AST’s were recommended and tested to better reflect the degradation modes seen in field operation. Also, BOL components were degraded in a test vehicle at UTC designed to accelerate the bus field operation.

  8. 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...

  9. Foucault current testing of ferritic steel fuel cans

    International Nuclear Information System (INIS)

    Stossel, A.

    1984-10-01

    The analysis of impedance involved by a Foucault current test of ferritic steel tubes, is quite different from the classical analysis which refers to non-magnetic tubes; more particularly, volume defects are considered as magnetic anomalies. Contrarily to current instructions which recommend to test the product in a satured magnetic state, it is very interesting to work with a continuous energizing field, comparatively low, corresponding to a sequenced magnetization, of which value is obtained according to the magnetic structure of the product. This analysis is useful when testing fast reactor fuel cans [fr

  10. 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

  11. 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.

  12. Melcor benchmarking against integral severe fuel damage tests

    Energy Technology Data Exchange (ETDEWEB)

    Madni, I.K. [Brookhaven National Lab., Upton, NY (United States)

    1995-09-01

    MELCOR is a fully integrated computer code that models all phases of the progression of severe accidents in light water reactor nuclear power plants, and is being developed for the U.S. Nuclear Regulatory Commission (NRC) by Sandia National Laboratories (SNL). Brookhaven National Laboratory (BNL) has a program with the NRC to provide independent assessment of MELCOR, and a very important part of this program is to benchmark MELCOR against experimental data from integral severe fuel damage tests and predictions of that data from more mechanistic codes such as SCDAP or SCDAP/RELAP5. Benchmarking analyses with MELCOR have been carried out at BNL for five integral severe fuel damage tests, namely, PBF SFD 1-1, SFD 14, and NRU FLHT-2, analyses, and their role in identifying areas of modeling strengths and weaknesses in MELCOR.

  13. ORR irradiation experiment OF-1: accelerated testing of HTGR fuel

    International Nuclear Information System (INIS)

    Tiegs, T.N.; Long, E.L. Jr.; Kania, M.J.; Thoms, K.R.; Allen, E.J.

    1977-08-01

    The OF-1 capsule, the first in a series of High-Temperature Gas-Cooled Reactor fuel irradiations in the Oak Ridge Research Reactor, was irradiated for more than 9300 hr at full reactor power (30 MW). Peak fluences of 1.08 x 10 22 neutrons/cm 2 (> 0.18 MeV) were achieved. General Atomic Company's magazine P13Q occupied the upper two-thirds of the test space and the ORNL magazine OF-1 the lower one-third. The ORNL portion tested various HTGR recycle particles and fuel bonding matrices at accelerated flux levels under reference HTGR irradiation conditions of temperature, temperature gradient, and fast fluence exposure

  14. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Gregson, Michael Warren; Brockmann, John E.; Nolte, O. (Fraunhofer institut fur toxikologie und experimentelle Medizin, Germany); Loiseau, O. (Institut de radioprotection et de Surete Nucleaire, France); Koch, W. (Fraunhofer institut fur toxikologie und experimentelle Medizin, Germany); Molecke, Martin Alan; Autrusson, Bruno (Institut de radioprotection et de Surete Nucleaire, France); Pretzsch, Gunter Guido (Gesellschaft fur anlagen- und Reaktorsicherheit, Germany); Billone, M. C. (Argonne National Laboratory, USA); Lucero, Daniel A.; Burtseva, T. (Argonne National Laboratory, USA); Brucher, W (Gesellschaft fur anlagen- und Reaktorsicherheit, Germany); Steyskal, Michele D.

    2006-10-01

    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{sub 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

  16. Conceptional design of test loop for FIV in fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Sim, W. G.; Yang, J. S.; Kim, S. W. [Hannam Univ., Taejeon (Korea)

    2001-01-01

    It is urgent to develop the analytical model for the structural/mechanical integrity of fuel rod. In general, it is not easy to develop a pure analytical model. Occasionally, experimental results have been utilized for the model. Because of this reason, it is required to design proper test loop. Using the optimized test loop, with the optimized test loop, the dynamic behaviour of the rod will be evaluated and the critical flow velocity, which the rod loses the stability in, will be measured for the design of the rod. To verify the integrity of the fuel rod, it is required to evaluate the dynamic behaviour and the critical flow velocity with the test loop. The test results will be utilized to the design of the rod. Generally, the rod has a ground vibration due to turbulence in wide range of flow velocity and the amplitude of vibration becomes larger by the resonance, in a range of the velocity where occurs vortex. The rod loses stability in critical flow velocity caused by fluid-elastic instability. For the purpose of the present work to perform the conceptional design of the test loop, it is necessary (1) to understand the mechanism of the flow-induced vibration and the related experimental coefficients, (2) to evaluate the existing test loops for improving the loop with design parameters and (3) to decide the design specifications of the major equipments of the loop. 35 refs., 23 figs., 2 tabs. (Author)

  17. The J-2X Fuel Turbopump - Design, Development, and Test

    Science.gov (United States)

    Tellier, James G.; Hawkins, Lakiesha V.; Shinguchi, Brian H.; Marsh, Matthew W.

    2011-01-01

    Pratt and Whitney Rocketdyne (PWR), a NASA subcontractor, is executing the design, development, test, and evaluation (DDT&E) of a liquid oxygen, liquid hydrogen two hundred ninety four thousand pound thrust rocket engine initially intended for the Upper Stage (US) and Earth Departure Stage (EDS) of the Constellation Program Ares-I Crew Launch Vehicle (CLV). A key element of the design approach was to base the new J-2X engine on the heritage J-2S engine with the intent of uprating the engine and incorporating SSME and RS-68 lessons learned. The J-2S engine was a design upgrade of the flight proven J-2 configuration used to put American astronauts on the moon. The J-2S Fuel Turbopump (FTP) was the first Rocketdyne-designed liquid hydrogen centrifugal pump and provided many of the early lessons learned for the Space Shuttle Main Engine High Pressure Fuel Turbopumps. This paper will discuss the design trades and analyses performed for the current J-2X FTP to increase turbine life; increase structural margins, facilitate component fabrication; expedite turbopump assembly; and increase rotordynamic stability margins. Risk mitigation tests including inducer water tests, whirligig turbine blade tests, turbine air rig tests, and workhorse gas generator tests characterized operating environments, drove design modifications, or identified performance impact. Engineering design, fabrication, analysis, and assembly activities support FTP readiness for the first J-2X engine test scheduled for July 2011.

  18. The Construction Work Completion of the Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, Kook Nam; Lee, Chung Young; Chi, Dae Young; Park, Su Ki; Shim, Bong Sik; Ahn, Sung Ho; Kim, Hark Rho [KAERI, Daejeon (Korea, Republic of); Lee, Jong Min [Global Nuclear Engineering Company, Daejeon (Korea, Republic of)

    2007-07-01

    FTL(Fuel Test Loop) is a facility that confirms performance of nuclear fuel at a similar irradiation condition with that of nuclear power plant. FTL consists of In-Pile Test Section (IPS) and Out-Pile System (OPS). FTL construction work began on August, 2006 and ended on March, 2007. During Construction, ensuring the worker's safety was the top priority and installation of the FTL without hampering the integrity of the HANARO was the next one. Task Force Team was organized to do a construction systematically and the communication between members of the task force team was done through the CoP(community of Practice) notice board provided by the Institute. The installation works were done successfully overcoming the difficulties such as on the limited space, on the radiation hazard inside the reactor pool, and finally on the shortening of the shut down period of the HANARO. Without a sweet of the workers of the participating company of HEC(Hyundae Engineering Co, Ltd), HDEC(HyunDai Engineering and Construction Co. Ltd), equipment manufacturer, and the task force team, it is not possible to install the FTL facility within the planned shutdown period. The Commissioning of the FTL is on due to check the function and the performance of the equipment and the overall system as well. The FTL shall start operation with high burn up test fuels in early 2008 if the commissioning and licensing progress on schedule.

  19. Accident situations tests HTR fuel with the device Kufa

    International Nuclear Information System (INIS)

    Kellerbauer, A. I.; Freis, D.

    2010-01-01

    The ceramic and ceramic-like coating materials in modern high-temperature reactor fuel are designed to ensure mechanical stability and retention of fission products under normal and transient conditions, regardless of the radiation damage sustained in-pile. In hypothetical depressurization and loss-of-forced-circulation (D LOFC) accidents, fuel elements of modular high-temperate reactors are exposed to temperatures several hundred degrees higher than during normal operation, causing increased thermo-mechanical stress on the coating layers. At the Institute for Transuranium Elements of the European Commission, a vigorous experimental program is being pursued with the aim of characterizing the performance of irradiated HTR fuel under such accident conditions. A cold finger device (Kufa), operational in ITUs hot cells since 2006, has been used to perform heating experiments on eight irradiated HTR fuel pebbles from the AVR experimental reactor and from dedicated irradiation campaigns at the High-Flux Reactor in Petten, the Netherlands. Gaseous fission products are collected in a cryogenic charcoal trap, while volatiles,are plated out on a water-cooled condensate plate. A quantitative measurement of the release is obtained by gamma spectroscopy. We highlight experimental results from the Kufa testing as well as the on-going development of new experimental facilities. (Author) 9 refs.

  20. ROTARY FUEL INJECTION PUMP WEAR TESTING USING A 30 %/ 70% ATJ/F-24 FUEL BLEND

    Science.gov (United States)

    2017-09-30

    Table Page Table 1 . 30% ATJ Blend & Neat F-24 Chemical & Physical Properties ..................................... 3 Table 2 . Bulk Modulus of...has included basic chemical and physical property investigation to identify surrogate fuel sources with similar properties as traditional petroleum... chemical and physical properties of the neat F-24 (AF-9623) and the tested 30% ATJ blend (AF-9625). Table 2 shows the speed of sound and bulk modulus

  1. Storage, inspection and sip testing of spent nuclear fuel from the HIFAR materials test reactor

    International Nuclear Information System (INIS)

    Selwyn, H.; Finlay, R.; Bull, P.; Irwin, A.

    2002-01-01

    Aluminum clad U-Al fuel used within the HIFAR MTR has been stored both in dry (underground) and wet (pond) storage facilities at the Lucas Heights site since the 1960's. As part of ANSTO's current program to send this fuel for long term storage or reprocessing, a significant level of visual inspection and water sip testing has been performed. This data has been used to demonstrate the integrity and suitability of the fuel for transport and receipt at the re processors interim storage ponds. This paper presents the key technical background-history of HIFAR fuel and its storage at Lucas Heights, presents the data obtained to date regarding its condition and discusses some observations regarding visual corrosion indicators and actual sip test results. (author)

  2. The safety concept and classification for HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1997-10-01

    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 principle subjects of this study is to determine the safety concept and classification of HANARO FTL for the design of systems. HANARO FTL should be approved by KINS for installation and operation. And these irradiation for facilities will be designed and installed based upon the safety concept and classification results. (author). 14 refs., 12 tabs., 10 figs.

  3. Shaft shock absorber tests for a spent fuel canister

    International Nuclear Information System (INIS)

    Kukkola, T.; Toermaelae, V.P.

    2005-06-01

    The disposal canister for spent nuclear fuel will be transferred by a lift to the repository, which is 500 m deep in the bedrock. Model tests were carried out with the objective to estimate weather feasible shock absorber can be developed against the design accident case where the canister should survive a free fall to the lift shaft. If the velocity of the canister is not controlled by air drag or by any other deceleration means, the impact velocity may reach ultimate speed of 100m/s. The canister would retain its integrity in impact on water when the bottom pit of the lift well is filled with groundwater. However, the canister would hit the pit bottom with high velocity since the water hardly slows down the canister. The impact to the bottom of the pit should be dampened mechanically. The tests demonstrated that 20 m high filling to the bottom pit of the lift well by Light Expanded Clay Aggregate (LECA), gives fair impact absorption to protect the fuel canister. Presence of ground water is not harmful for impact absorption system provided that the ceramic gravel is not floating too high from the pit bottom. Almost ideal impact absorption conditions are met if the water high level does not exceed two thirds of the height of the gravel. Shaping of the bottom head of the cylindrical canister does not give meaningful advantages to the impact absorption system. The flat nose bottom head of the fuel canister gives adequate deceleration properties. (orig.)

  4. Underwater Coatings Testing for INEEL Fuel Basin Applications

    Energy Technology Data Exchange (ETDEWEB)

    Julia L. Tripp

    2004-01-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) is deactivating several fuel storage basins. Airborne contamination is a concern when the sides of the basins are exposed and allowed to dry during water removal. One way of controlling this airborne contamination is to fix the contamination in place while the pool walls are still submerged. There are many underwater coatings available on the market that are used in marine, naval and other applications. A series of tests were run to determine whether the candidate underwater fixatives are easily applied and adhere well to the substrates (pool wall materials) found in INEEL fuel pools. The four pools considered included (1) Test Area North (TAN-607) with epoxy painted concrete walls; (2) Idaho Nuclear Technology and Engineering Center (INTEC) (CPP-603) with bare concrete walls; (3) Materials Test Reactor (MTR) Canal with stainless steel lined concrete walls; and (4) Power Burst Facility (PBF-620) with stainless steel lined concrete walls on the bottom and epoxy painted carbon steel lined walls on the upper portions. Therefore, the four materials chosen for testing included bare concrete, epoxy painted concrete, epoxy painted carbon steel, and stainless steel. The typical water temperature of the pools varies from 55 F to 80 F dependent on the pool and the season. These tests were done at room temperature.

  5. Tests of prototype salt stripper system for IFR fuel cycle

    International Nuclear Information System (INIS)

    Carls, E.L.; Blaskovitz, R.J.; Johnson, T.R.; Ogata, T.

    1993-01-01

    One of the waste treatment steps for the on-site reprocessing of spent fuel from the Integral Fast Reactor fuel cycles is stripping of the electrolyte salt used in the electrorefining process. This involves the chemical reduction of the actinides and rare earth chlorides forming metals which then dissolve in a cadmium pool. To develop the equipment for this step, a prototype salt stripper system has been installed in an engineering scale argon-filled glovebox. Pumping trails were successful in transferring 90 kg of LiCl-KCl salt containing uranium and rare earth metal chlorides at 500 degree C from an electrorefiner to the stripper vessel at a pumping rate of about 5 L/min. The freeze seal solder connectors which were used to join sections of the pump and transfer line performed well. Stripping tests have commenced employing an inverted cup charging device to introduce a Cd-15 wt % Li alloy reductant to the stripper vessel

  6. THERMAL VACUUM TEST OF ORBITAL STATIC MOISTURE-REMOVAL FUEL CELL.

    Science.gov (United States)

    The report presents the results of a thermal vacuum chamber test of an orbital fuel cell of advanced design. The fuel cell package used a static moisture-removal system. The fuel cell , tested in the thermal vacuum chamber at Wright-Patterson AFB, gave satisfactory results. This test constituted the second and final ground qualification of this orbital fuel cell prior to orbital test. (Author)

  7. Non destructive testing of irradiated fuel assemblies at the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Jose Eduardo Rosa da; Terremoto, Luis Antonio Albiac; Castanheira, Myrthes; Teodoro, Celso Antonio; Silva, Antonio Teixeira e; Damy, Margaret de Almeida; Lucki, Georgi [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mails: jersilva@ipen.br; laaterre@ipen.br; myrthes@ipen.br; cteodoro@ipen.br; teixeira@ipen.br; madamy@ipen.br; glucki@ipen.br

    2007-07-01

    Fuel performance and nuclear fuel qualification require a post-irradiation analysis. Non-destructive methods are utilised both in irradiated fuel storage pools and in hot-cells laboratories. As Brazil does not have hot-cells facilities for post-irradiation analysis, a qualification program for the Material Testing Reactor (MTR) fuel elements made at IPEN/CNEN-SP was adopted, based on non-destructive tests. The IPEN Fuel Engineering Group - CENC developed basic facilities for fuels post-irradiated analysis inside the reactor pool, which gives indications of: general state, by visual inspection; the integrity of the irradiated fuel cladding, by sipping tests; thickness measurements of the fuel miniplates during the irradiation time, for swelling evaluation; and, local burn-up evaluation by gamma spectrometry along the active area of the fuel element. This work describes that facilities, equipment and examples of some irradiated fuels analysis performed. (author)

  8. Non destructive testing of irradiated fuel assemblies at the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Silva, Jose Eduardo Rosa da; Terremoto, Luis Antonio Albiac; Castanheira, Myrthes; Teodoro, Celso Antonio; Silva, Antonio Teixeira e; Damy, Margaret de Almeida; Lucki, Georgi

    2007-01-01

    Fuel performance and nuclear fuel qualification require a post-irradiation analysis. Non-destructive methods are utilised both in irradiated fuel storage pools and in hot-cells laboratories. As Brazil does not have hot-cells facilities for post-irradiation analysis, a qualification program for the Material Testing Reactor (MTR) fuel elements made at IPEN/CNEN-SP was adopted, based on non-destructive tests. The IPEN Fuel Engineering Group - CENC developed basic facilities for fuels post-irradiated analysis inside the reactor pool, which gives indications of: general state, by visual inspection; the integrity of the irradiated fuel cladding, by sipping tests; thickness measurements of the fuel miniplates during the irradiation time, for swelling evaluation; and, local burn-up evaluation by gamma spectrometry along the active area of the fuel element. This work describes that facilities, equipment and examples of some irradiated fuels analysis performed. (author)

  9. Focused ultrasonic wave testing, in immersion of spent fuel cans

    International Nuclear Information System (INIS)

    Poinboeuf, P.; Furlan, J.

    1984-10-01

    To detect weak and very weak damage of the fuel can, ultrasonic testing has been used. For that, a simple mechanical device, allowing to maintain an optimal ultrasonic focussing on irradiated cans, is presented. Its aim is to correct the variation of the incidence angle due to the possible ovalization of pins. After a description of the device, the results obtained with tests carried out on non-irradiated cans, including artificial ovalized regions, standard defects, are presented. After the description of the adaptation of this mechanism on a test bench which allows an helicoidal exploration of pins, some results obtained in hot cell during examinations experimental pins and previously tested by Foucault current [fr

  10. Certification testing of the MOX Fresh Fuel Package (MFFP)

    International Nuclear Information System (INIS)

    Nichols, J.C. III; Yapuncich, F.L.

    2004-01-01

    Packaging Technology, Inc. (PacTec) is designing the MFFP as part of the Duke, COGEMA, Stone and Webster (DCS) consortium. DCS is tasked with providing the Department of Energy (DOE) with domestic MOX fuel fabrication and reactor irradiation services for the purpose of disposing of surplus weapons usable plutonium. This paper will discuss the development of the MFFP certification test program. The MFFP was subjected to a total of eleven free and puncture drops of the course of the certification testing. Because of the plutonium content, the design must be a Type BF, which among other things requires a containment boundary with a tested leakage rate of 1 x 10 -7 cm 3 /s air at 1 atm absolute and 25 C, or less. Both economics (desire for maximized payload) and operational (conveyance mode restricts size and weight) constraints lead to a highly optimized design. The optimized package design led to a significant test program which needed to address the containment boundary stability, puncture resistance of the package and lid end impact limiter, structural performance of the light weight lid structure, and stability of the internal structures. The test program efficiently balanced the test objectives while minimizing the number of costly hardware items used during this destructive testing. This balance achieved by strategic replacement of mock and prototypic payloads, impact limiters, and by careful test order considerations. The paper will conclude with a selected summary of the testing and an assessment of the test programs thoroughness

  11. Test Operations Procedure (TOP) 02-2-603A Vehicle Fuel Consumption

    Science.gov (United States)

    2012-05-10

    Final 3. DATES COVERED (From - To) 4. TITLE AND SUBTITLE Test Operations Procedure (TOP) 02-2-603A Vehicle Fuel Consumption 5a. CONTRACT...test methods used to measure and present the fuel consumption characteristics for wheeled and tracked vehicles. Specific facilities, instrumentation...test controls, and analysis techniques are presented. 15. SUBJECT TERMS fuel consumption traction battery hybrid

  12. Development of the instrumented capsule for nuclear fuel irradiation test at HANARO

    International Nuclear Information System (INIS)

    Sohn, Jae Min; Oh, Jong Myung; Park, Sung Jae; Shin, Yoon Taeg; Kim, Bong Goo; Kang, Young Hwan; Kim, Hark Rho; Kim, Young Jin

    2006-01-01

    The instrumented capsule for the nuclear fuel irradiation test (hereinafter referred to as 'instrumented fuel capsule'), which is crucial for the verification of a nuclear fuel performance and safety, has been developed at HANARO (High-flux Advanced Neutron Application Reactor). The irradiation test of the first instrumented fuel capsule was carried out in March 2003 and the irradiation test of the second instrumented fuel capsule was carried out in April 2004 at HANARO. Through the irradiation tests of the two capsules, the design specifications and safety of the instrumented fuel capsule were successfully verified. In the first instrumented fuel capsule, only the technologies for measuring the center temperature of the nuclear fuel and neutron flux were implemented. In the second instrumented fuel capsule, the technologies for measuring the center temperature of the nuclear fuel, the internal pressure of the fuel rod, the elongation of the nuclear fuel and the neutron flux were implemented. Currently the dual instrumented technologies that alloy for two characteristics to be measured simultaneously in one fuel rod, is being developed. The duel instrumented fuel rods have been successfully designed as a part of the technology enhancement program for the instrumented fuel capsule. The instrumented fuel capsule will be utilized for the development of nuclear fuel. The instrumentation technologies for measuring the nuclear fuel characteristics will be applied to the 3-pin FTL (Fuel Test Loop) facility which is currently being developed. And, the duel instrumented technologies will contribute to enhancing the efficiency of the irradiation test using an instrumented fuel capsule at HANARO. (author)

  13. 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

  14. Fuel bundle examination techniques for the Phebus fission product test

    International Nuclear Information System (INIS)

    Blanc, J.Y.; Clement, B.; Hardt, P. von der

    1996-01-01

    The paper develops the non-destructive examinations, with a special emphasis on transmission tomography, performed in the Phebus facility, using a linear accelerator associated with a line scan camera based on PCD components. This particular technique enabled the high level of penetration to be obtained, necessary for this high density application. Spatial resolution is not far from the theoretical limit and the density resolution is often adequate. This technique permitted: 1) to define beforehand the cuts on a precise basis, avoiding a long step-by-step choice as in previous in-pile tests; 2) to determine, at an early stage, mass balance, material relocations (in association with axial gamma spectrometry), and FP distribution, as an input into re-calculations of the bundle events. However, classical cuttings, periscopic visual examinations, macrographies, micrographies and EPMA analyses remain essential to give oxidation levels (in the less degraded zones), phase aspect and composition, to distinguish between materials of identical density, and, if possible, to estimate temperatures. Oxidation resistance of sensors (thermocouples or ultrasonic thermometers) is also traced. The EPMA gives access to the molten material chemical analyses, especially in the molten fuel blockage area. The first results show that an important part of the fuel bundle melted (which was one of the objectives of this test) and that the degradation level is close to TIMI-2 with a molten plug under a cavity surrounded by an uranium-rich crust. In lower and upper areas fuel rods are less damaged. Complementaries between these examination techniques and between international teams involved will be major advantages in the Phebus FPT0 test comprehension. 3 refs, 9 figs

  15. Full scale torch tests on spent fuel cask shipping system

    International Nuclear Information System (INIS)

    Vigil, M.G.; Trujillo, A.A.; Yoshimura, H.R.; Joseph, B.J.; Eggers, P.E.; Crawford, H.L.

    1982-01-01

    Full scale experimental measurements, including the instrumentation designed to obtain the data, are presented on the thermal effects of torch fires on a large, spent nuclear fuel shipping cask. The measured temperature data in the various materials of the multilayered cask are unique, since no torch tests have been previously performed on a cask: These data were obtained during a series of four torch tests which simulate a situation in which the relief valve of a liquefied gas tank railcar has been opened and and the contents are vented and ignited so that the resultant torch impinges on the cask. The modified cask instrumentation geometry and materials are discussed. Temperature data throughout the cask are compared for two cask on the corrugated outer jacket surface, within the neutron shield, on the carbon steel shell, on the inner stainless steel shell and near the cask head closure seals are presented for the four torch tests

  16. Effect of chemical composition and superheat on macrostructure of high Cr white iron castings

    Energy Technology Data Exchange (ETDEWEB)

    Dogan, Omer N.

    2005-08-01

    White cast irons are frequently used in applications requiring high wear resistance. High Cr white cast irons have a composite microstructure composed of hard (Fe,Cr)7C3 carbides in a steel matrix. Previous research has indicated that the equiaxed region of these high Cr white iron castings is much more wear resistant under high stress abrasive conditions than the columnar region, when the carbides are oriented perpendicular to the wear surface. In the present study, the effect of both the chemical composition, particularly carbon content, and the pouring superheat of the melt on the macrostructure of high Cr white iron castings is investigated.

  17. 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...

  18. 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)

  19. Liquefied Gaseous Fuels Spill Test Facility: Overview of STF capabilities

    International Nuclear Information System (INIS)

    Gray, H.E.

    1993-01-01

    The Liquefied Gaseous Fuels Spill Test Facility (STF) constructed at the Department of Energy's Nevada Test Site is a basic research tool for studying the dynamics of accidental releases of various hazardous liquids. This Facility is designed to (1) discharge, at a controlled rate, a measured volume of hazardous test liquid on a prepared surface of a dry lake bed (Frenchman Lake); (2) monitor and record process operating data, close-in and downwind meteorological data, and downwind gaseous concentration levels; and (3) provide a means to control and monitor these functions from a remote location. The STF will accommodate large and small-scale testing of hazardous test fluid release rates up to 28,000 gallons per minute. Spill volumes up to 52,800 gallons are achievable. Generic categories of fluids that can be tested are cryogenics, isothermals, aerosol-forming materials, and chemically reactive. The phenomena that can be studied include source definition, dispersion, and pool fire/vapor burning. Other capabilities available at the STF include large-scale wind tunnel testing, a small test cell for exposing personnel protective clothing, and an area for developing mitigation techniques

  20. Proton Exchange Membrane Fuel Cell Engineering Model Powerplant. Test Report: Benchmark Tests in Three Spatial Orientations

    Science.gov (United States)

    Loyselle, Patricia; Prokopius, Kevin

    2011-01-01

    Proton exchange membrane (PEM) fuel cell technology is the leading candidate to replace the aging alkaline fuel cell technology, currently used on the Shuttle, for future space missions. This test effort marks the final phase of a 5-yr development program that began under the Second Generation Reusable Launch Vehicle (RLV) Program, transitioned into the Next Generation Launch Technologies (NGLT) Program, and continued under Constellation Systems in the Exploration Technology Development Program. Initially, the engineering model (EM) powerplant was evaluated with respect to its performance as compared to acceptance tests carried out at the manufacturer. This was to determine the sensitivity of the powerplant performance to changes in test environment. In addition, a series of tests were performed with the powerplant in the original standard orientation. This report details the continuing EM benchmark test results in three spatial orientations as well as extended duration testing in the mission profile test. The results from these tests verify the applicability of PEM fuel cells for future NASA missions. The specifics of these different tests are described in the following sections.

  1. Fabrication experience and integrity confirmation tests of the first-loading-fuel of the HTTR

    International Nuclear Information System (INIS)

    Sawa, Kazuhiro; Tobita, Tsutomu; Sumita, Junya; Mogi, Haruyoshi; Yoshimuta, Shigeharu; Suzuki, Shuuichi; Deushi, Kouzaburou; Kato, Shigeru

    1999-01-01

    The first-loading core of the High Temperature Engineering Test Reactor (HTTR) consists of 150 fuel assemblies. An HTTR fuel assembly is so-called a pin-in-block type of hexagonal graphite block. A fuel rod consists of a graphite sleeve and of 14 fuel compacts. In a fuel compact, about 13,000 TRISO coated fuel panicles are dispersed densely. The coated fuel panicle is TRISO (Tri-isotropic) type with four coating layers. The fuel kernel is low-enriched (average 6wt%) UO 2 . The fabrication of the first-loading fuel started from June 1995. A total of 4,770 fuel rods were successfully produced and transferred to the reactor building of the HTTR. Finally, in the reactor building, the fuel rods were inserted to the graphite blocks to form fuel assemblies. On December 1997, 150 fuel assemblies were completely formed and were stored in new fuel storage cells. The cells were filled with helium gas to keep the fuel blocks in dry condition. Fabrication technology of the HTTR fuel was established through a lot of R and D activities and fabrication experiences of irradiation examination samples spread over about 30 years. High quality and production efficiency of fuel were achieved by the development of the fuel kernel process using the vibration dropping technology, the continuous 4-layer coating process and optimization of the compaction conditions. In the safety design of HTGR fuel, it is important to retain fission products within the coated fuel panicles so that their release to the primary coolant may not exceed an acceptable level. From this point of view, as-fabricated failure fraction is important. In the specification, SiC-failure and exposed uranium fractions were determined to be less than 1.5x10 -3 and l.5x10 -4 , respectively. The quality of the first loading fuel fully satisfied the design specifications for the fuel. The fuel compacts contained almost no through-coatings failed particles and few SiC-defective particles. Average through-coatings and Si

  2. 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)

  3. Fracture mapping at the Spent Fuel Test-Climax

    International Nuclear Information System (INIS)

    Wilder, D.G.; Yow, J.L. Jr.

    1981-05-01

    Mapping of geologic discontinuities has been done in several phases at the Spent Fuel Test-Climax (SFT-C) in the granitic Climax stock at the Nevada Test Site. Mapping was carried out in the tail drift, access drift, canister drift, heater drifts, instrumentation alcove, and receiving room. The fractures mapped as intersecting a horizontal datum in the canister and heater drifts are shown on one figure. Fracture sketch maps have been compiled as additional figures. Geologic mapping efforts were scheduled around and significantly impacted by the excavation and construction schedules. Several people were involved in the mapping, and over 2500 geologic discontinuities were mapped, including joints, shears, and faults. Some variance between individuals' mapping efforts was noticed, and the effects of various magnetic influences upon a compass were examined. The examination of compass errors improved the credibility of the data. The compass analysis work is explained in Appendix A. Analysis of the fracture data will be presented in a future report

  4. 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.

  5. 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)

  6. Results of a diesel multiple unit fuel tank blunt impact test

    Science.gov (United States)

    2017-04-04

    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 types of dyn...

  7. 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...

  8. Mineralogic and petrologic investigation of post-test core samples from the Spent Fuel Test - Climax

    International Nuclear Information System (INIS)

    Ryerson, F.J.; Beiriger, J.

    1985-02-01

    We have characterized a suite of samples taken subsequent to the end of the Spent Fuel Test - Climax by petrographic and microanalytical techniques and determined their mineral assemblage, modal properties, and mineral chemistry. The samples were obtained immediately adjacent to the canister borehole at a variety of depths and positions within the canister drift, as well as radially outward from each canister hole. This method of sampling allows variations in post-test mineralogic properties to be evaluated on the basis of (1) depth along a particular canister hole and (2) position within the canister drift, with respect to the heat and radiation sources, and with respect to the pre - test samples. In no case did we find any significant correlation between the mineralogical properties and variables listed above. In short, the Spent Fuel Test - Climax has produced no identifiable mineralogical response in the Climax quartz monzonite. 12 refs., 11 figs., 5 tabs

  9. Design and Fabrication of the Double Cladding Instrumented Fuel Rods and the Instrumented Fuel Capsule(07F-06K) for the Irradiation Test at HANARO

    International Nuclear Information System (INIS)

    Sohn, Jae Min; Oh, Jong Myung; Oh, Soo Yeol; Park, Sung Jae; Sho, Man Soon; Kim, Bong Goo; Choo, Kee Nam; Kim, Young Ki

    2009-01-01

    An instrumented capsule for a nuclear fuel irradiation test (hereinafter referred to as 'instrumented fuel capsule'), which is crucial for the verification of a nuclear fuel performance and safety, has been developed to measure the fuel characteristics, such as the centerline and surface temperatures of the nuclear fuel, the internal pressure of a fuel rod, the elongation of the fuel pellet and the neutron fluxes during an irradiation test at HANARO(High-flux Advanced Neutron Application Reactor). The irradiation test of the first instrumented fuel capsule(02F-11K) was carried out for verification test at HANARO in March 2003. Through the irradiation tests of the some capsules, the design specifications and safety of the instrumented fuel capsule were verified successfully. And the dual instrumented fuel rods, which allow for two characteristics to be measured simultaneously in one fuel rod, have been developed to enhance the efficiency of the irradiation test using the instrumented fuel capsule. In this paper, we designed and fabricated a double cladding fuel rod to control the high temperature of nuclear fuels during an irradiation test at HANARO. And we design an instrumented fuel capsule(07F-06K) for an irradiation test of the double cladding fuel rods. We have designed and fabricated the double cladding fuel rod mockups and performed the out-pile tests using these mockups. The purposes of the out-pile tests were to analyze an effect of a gap size(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. Through the results of the out-pile tests, we have obtained the effects of a gap size and a gas mixture ratio on the temperature of nuclear fuels. Therefore an double cladding fuel rod and the 07F- 06K instrumented fuel capsule were designed on the base of the results of the out-pile tests using the mockups

  10. 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

  11. Fuel cell drive system with hydrogen generation in test

    Science.gov (United States)

    Emonts, B.; Bøgild Hansen, J.; Schmidt, H.; Grube, T.; Höhlein, B.; Peters, R.; Tschauder, A.

    In the future, drive systems for vehicles with polymer electrolyte membrane fuel cells (PEMFC) may be the environmentally more acceptable alternative to conventional drives with internal combustion engines. The energy carrier may not be gasoline or diesel, as in combustion engines today, but methanol, which is converted on-board into a hydrogen-rich synthesis gas in a reforming reaction with water. After removal of carbon monoxide in a gas-cleaning step, the conditioned synthesis gas is converted into electricity in a fuel cell using air as the oxidant. The electric energy thus generated serves to supply a vehicle's electric drive system. Based on the process design for a test drive system, a test facility was prepared and assembled at Forschungszentrum Jülich (FZJ). Final function tests with the PEMFC and the integrated compact methanol reformer (CMR) were carried out to determine the performance and the dynamic behaviour. With regard to the 50-kW(H 2)-compact methanol reformer, a special design of catalytic burner was constructed. The burner units, with a total power output of 16 kW, were built and tested under different states of constant and alternating load. If selecting a specific catalyst loading of 40 g Pt/m 2, the burner emissions are below the super ultra low emission vehicle (SULEV) standard. The stationary performance test of the CMR shows a specific hydrogen production of 6.7 m N3/(kg cat h) for a methanol conversion rate of 95% at 280°C. Measurements of the transient behaviour of the CMR clearly show a response time of about 20 s, reaching 99% of the hydrogen flow demand due to the limited performance of the test facility control system. Simulations have been carried out in order to develop a control strategy for hydrogen production by the CMR during the New European Driving Cycle (NEDC). Based on the NEDC, an optimized energy management for the total drive system was evaluated and the characteristic data for different peak load storage systems are

  12. 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)

  13. Fuel dynamics loss-of-flow test L3. Final report

    International Nuclear Information System (INIS)

    Fischer, A.K.; Lo, R.K.; Barts, E.W.

    1976-06-01

    The behavior of FTR-type, mixed-oxide, preirradiated, ''intermediate-power-structure'' fuel during a simulation of an FTR loss-of-flow accident was studied in the Mark-IIA integral TREAT loop. Analysis of the data reported here leads to a postulated scenario (sequence and timing) of events in the test. This scenario is presented, together with the calculated timing of events obtained by use of the SAS code. The initial fuel motion, starting during the preheat phase, consisted of coherent motion of the entire intact fuel bundle toward the pump. Incoherence developed as temperature rose. The fuel motion was mostly upward, and the greatest was in the top third of the fuel column. Fuel fragments formed against the pump side of the fluted tube near the original fuel midplane. A penetration of fluted tube occurred. A sudden voiding of the central region of the fuel column occurred at 29.75 s and was largely completed within 150 ms. The lower steel blockage of the fuel elements occurred in the vicinity of the lower insulator pellets. The upper steel blockage just above the tops of the original fuel pins appeared to have channels through it. Cladding and spacer wires melted away in the fuel section. Fuel pellets were only evident at and above the top and at the bottom of the original fuel column, where a large mass of melted fuel was present. Over the length of the fuel column, most of the fluted tube had melted away

  14. Study on the requirement for the fuel test loop performance in the HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Chul; Kim, Hark Rho

    2000-06-01

    The requirement of the FTL (Fuel Test Loop) performance were investigated to reaffirm the technical feasibility of the FTL facility which is in consideration to install at the HANARO. LH hole in the reflector region and OR3 in the outer core region are considered as candidate sites for the IPS (In-Pile Section) for the analysis purpose. The achievable linear power at test fuel pin(s) and neutron flux levels at the cladding are analyzed in the IPS, which accommodates CANDU or PWR test fuel. The enrichment of test fuels is assumed as natural uranium for CANDU and 3.5% or 5% for PWR. The test fuel configuration is bundle or 7-pin in LH hole but 1-pin in OR site. For the CANDU test fuel, the target linear power of 60kW/m can not be achieved for all cases. For the PWR test fuel, the target linear power of 40kW/m is obtained at fuels located in the direction of the core for only the case of 5% bundle irradiation. From a sensitivity study, the linear power at test fuel is expected to increase at least 30% of the present results if the core bumup effect, optimization of the ratio of fuel-to-moderator number density, etc., are considered in the detail design. Thus, the linear power for PWR fuel is expected to reach the target value, and that for CANDU fuel will reach the target value if enriched fuel is used. The fast neutron flux at the test fuel cladding is estimated for most cases to be lower than one-third of the target value of 10{sup 1}4 n/cm{sup 2}-sec and expected not to reach the target value.

  15. Spent-fuel dry-storage testing at E-MAD (March 1978-March 1982)

    International Nuclear Information System (INIS)

    Unterzuber, R.; Milnes, R.D.; Marinkovich, B.A.; Kubancsek, G.M.

    1982-09-01

    From March 1978 through March 1982, spent fuel dry storage tests were conducted at the Engine Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site to confirm that commercial reactor spent fuel could be encapsulated and passively stored in one or more interim dry storage cell concepts. These tests were: electrically heated drywell, isolated and adjacent drywell, concrete silo, fuel assembly internal temperature measurement, and air-cooled vault. This document presents the test data and results as well as results from supporting test operations (spent fuel calorimetry and canister gas sampling)

  16. 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)

  17. Imminent: Irradiation Testing of (Th,Pu)O2 Fuel - 13560

    International Nuclear Information System (INIS)

    Kelly, Julian F.; Franceschini, Fausto

    2013-01-01

    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)

  18. Test design description Volume 2, Part 1. IFR-1 metal fuel irradiation test (AK-181) element as-built data

    Energy Technology Data Exchange (ETDEWEB)

    Dodds, N. E.

    1986-06-01

    The IFR-1 Test, designated as the AK-181 Test Assembly, will be the first irradiation test of wire wrapped, sodium-bonded metallic fuel elements in the Fast Flux Test Facility (FFTF). The test is part of the Integral Fast Reactor (IFR) fuels program conducted by Argonne National Laboratory (ANL) in support of the Innovative Reactor Concepts Program sponsored by the US Department of Energy (DOE). One subassembly, containing 169 fuel elements, will be irradiated for 600 full power days to achieve 10 at.% burnup. Three metal fuel alloys (U-10Zr, U-8Pu-10Zr) will be irradiated in D9 cladding tubes. The metal fuel elements have a fuel-smeared density of 75% and each contains five slugs. The enriched zone contains three slugs and is 36-in. long. One 6.5-in. long depleted uranium axial blanket slug (DU-10Zr) was loaded at each end of the enriched zone. the fuel elements were fabricated at ANL-W and delivered to Westinghouse-Hanford for wirewrapping and assembly into the test article. This Test Design Description contains relevant data on compositions, densities, dimensions and weights for the cast fuel slugs and completed fuel elements. The elements conform to the requirements in MG-22, "Users` Guide for the Irradiation of Experiments in the FTR."

  19. Test experiences with the DaimlerChrysler: Fuel cell electric vehicle NECAR

    Directory of Open Access Journals (Sweden)

    Friedlmeier Gerardo

    2002-01-01

    Full Text Available The DalmlerChrysler fuel cell electric vehicle NECAR 4, a hydrogen-fueled zero-emission compact car based on the A-Class of Mercedes-Benz, is described. Test results obtained on the road and on the dynamometer are presented. These and other results show the high technological maturity reliability and durability already achieved with fuel cell technology.

  20. 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

  1. 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)

  2. Fuel Cell Stack Testing and Durability in Support of Ion Tiger UAV

    Science.gov (United States)

    2010-06-02

    This report covers efforts by the Hawaii Natural Energy Institute (HNEI) of the University of Hawaii under the ONR-funded Ion Tiger UAV award that included testing of Ion Tiger fuel cell stacks in HNEI’s Hawaii Fuel Cell Test Facility located in Honolulu, Hawaii. Work was focused on steady-state stack characteristics of Protonex fuel cell stacks under various operating conditions. In addition, Hardware-in-the-Loop testing was performed to characterize dynamic

  3. Subcritical Measurements Research Program for Fresh and Spent Materials Test Reactor Fuels

    International Nuclear Information System (INIS)

    Blanchard, A.

    1999-01-01

    'A series of subcritical noise measurements were performed on fresh and spent University of Missouri Research Reactor fuel assemblies. These experimental measurements were performed for the purposes of providing benchmark quality data for validating transport theory computer codes and nuclear cross-section data used to perform criticality safety analyses for highly enriched, uranium-aluminum Material Test Reactor fuel assemblies. A mechanical test rig was designed and built to hold up to four fuel assemblies and neutron detectors in a subcritical array. The rig provided researchers with the ability to evaluate the reactivity effects of variable fuel/detector spacing, fuel rotation, and insertion of metal reflector plates into the lattice.'

  4. Simulation, State Estimation and Control of Nonlinear Superheater Attemporator using Neural Networks

    DEFF Research Database (Denmark)

    Bendtsen, Jan Dimon; Sørensen, O.

    2000-01-01

    This paper considers the use of neural networks for nonlinear state estimation, system identification and control. As a case study we use data taken from a nonlinear injection valve for a superheater attemporator at a power plant. One neural network is trained as a nonlinear simulation model...... of the process, then another network is trained to act as a combined state and parameter estimator for the process. The observer network incorporates smoothing of the parameter estimates in the form of regularization. A pole placement controller is designed which takes advantage of the sample......-by-sample linearizations and state estimates provided by the observer network. Simulation studies show that the nonlinear observer-based control loop performs better than a similar control loop based on a linear observer....

  5. Simulation, State Estimation and Control of Nonlinear Superheater Attemporator using Neural Networks

    DEFF Research Database (Denmark)

    Bendtsen, Jan Dimon; Sørensen, O.

    1999-01-01

    This paper considers the use of neural networks for nonlinear state estimation, system identification and control. As a case study we use data taken from a nonlinear injection valve for a superheater attemporator at a power plant. One neural network is trained as a nonlinear simulation model...... of the process, then another network is trained to act as a combined state and parameter estimator for the process. The observer network incorporates smoothing of the parameter estimates in the form of regularization. A pole placement controller is designed which takes advantage of the sample......-by-sample linearizations and state estimates provided by the observer network. Simulation studies show that the nonlinear observer-based control loop performs better than a similar control loop based on a linear observer....

  6. 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.

  7. Active approach for holographic nondestructive testing of satellite fuel tanks

    Science.gov (United States)

    Merz, Torsten; Elandaloussi, Frank; Paulus, Dietrich W.; Osten, Wolfgang

    1999-09-01

    In computer vision several views exist how to solve vision problems. The first general methodology was introduced by Marr; he proposed a data-driven and straightforward analysis strategy. Nowadays the concept of active vision introduced by Aloimonos et al. becomes more and more important. In contrast to Marr's philosophy, active vision implies a feedback loop which consists of sensors and active components. In this paper we present a system for the identification of material faults under the surface of a test object. For that purpose the specimen is elastically deformed, then the deformation is made visible using holographic interferometry, and finally flaw parameters are estimated using a model-based approach to analyze interferograms. This is an underconstrained computer vision problem which is regularized using a priori knowledge and an active modification of the experimental setup. More mathematically, this vision task can be seen in the context of inverse problem theory. In this contribution we describe the system and point out how it is related to the methodologies named above. To illustrate the functionality of the system, results are shown from nondestructive testing of satellite fuel tanks.

  8. LMFBR fuel-design environment for endurance testing, primarily of oxide fuel elements with local faults

    International Nuclear Information System (INIS)

    Warinner, D.K.

    1980-01-01

    The US Department of Energy LMFBR Lines-of-Assurance are briefly stated and local faults are given perspective with an historical review and definition to help define the constraints of LMFBR fuel-element designs. Local-fault-propagation (fuel-element failure-propagation and blockage propagation) perceptions are reviewed. Fuel pin designs and major LMFBR parameters affecting pin performance are summarized. The interpretation of failed-fuel data is aided by a discussion of the effects of nonprototypicalities. The fuel-pin endurance expected in the US, USSR, France, UK, Japan, and West Germany is outlined. Finally, fuel-failure detection and location by delayed-neutron and gaseous-fission-product monitors are briefly discussed to better realize the operational limits

  9. Test plan for reactions between spent fuel and J-13 well water under unsaturated conditions

    International Nuclear Information System (INIS)

    Finn, P.A.; Wronkiewicz, D.J.; Hoh, J.C.; Emery, J.W.; Hafenrichter, L.D.; Bates, J.K.

    1993-01-01

    The Yucca Mountain Site Characterization Project is evaluating the long-term performance of a high-level nuclear waste form, spent fuel from commercial reactors. Permanent disposal of the spent fuel is possible in a potential repository to be located in the volcanic tuff beds near Yucca Mountain, Nevada. During the post-containment period the spent fuel could be exposed to water condensation since of the cladding is assumed to fail during this time. Spent fuel leach (SFL) tests are designed to simulate and monitor the release of radionuclides from the spent fuel under this condition. This Test Plan addresses the anticipated conditions whereby spent fuel is contacted by small amounts of water that trickle through the spent fuel container. Two complentary test plans are presented, one to examine the reaction of spent fuel and J-13 well water under unsaturated conditions and the second to examine the reaction of unirradiated UO 2 pellets and J-13 well water under unsaturated conditions. The former test plan examines the importance of the water content, the oxygen content as affected by radiolysis, the fuel burnup, fuel surface area, and temperature. The latter test plant examines the effect of the non-presence of Teflon in the test vessel

  10. TESTING OF GAS REACTOR MATERIALS AND FUEL IN THE ADVANCED TEST REACTOR

    International Nuclear Information System (INIS)

    Grover, S.B.

    2004-01-01

    The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

  11. 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.

  12. On0Line Fuel Failure Monitor for Fuel Testing and Monitoring of Gas Cooled Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ayman I. Hawari; Mohamed A. Bourham

    2010-04-22

    IVery 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.

  13. Proceedings of the international meeting on development, fabrication and application of reduced enrichment fuels for research and test reactors

    International Nuclear Information System (INIS)

    1983-08-01

    Separate abstracts were prepared for each of the papers presented in the following areas: (1) Reduced Enrichment Fuels for Research and Test Reactors (RERTR) Program Status; (2) Fuel Development; (3) Fuel Demonstrations; (4) General Topics; and (5) Specific Reactor Applications

  14. Conceptual design of experimental LFR fuel element for testing in TRIGA reactor, ACPR zone

    International Nuclear Information System (INIS)

    Nastase, D.; Olteanu, G.; Ioan, M.; Pauna, E.

    2013-01-01

    In the pulsed area of the TRIGA reactor (ACPR zone), the irradiation tests called ''rapid insertions of reactivity on different types of nuclear fuel elements'' are usually realized. During these tests, in the fuel element high powers for a relatively short period of time (about few milliseconds) are generated. The generated heat in fuel pellets raise their central temperature to values over 100 deg C. The conceptual design of an experimental fuel element proposed to be developed and presented in this paper must fulfill a couple of requirements, as follows: to ensure full compatibility with irradiation device sample holder (compatibility is achieved through reduced length of the fuel stack pellets - this way assures a flow flattening on the entire length of the fuel element); to be compatible with the project of irradiated fuel bundle in Lead cooled Fast Reactors (LFR). (authors)

  15. Dynamical Capillary Rise Photonic Sensor for Testing of Diesel and Biodiesel Fuel

    Directory of Open Access Journals (Sweden)

    Michal BORECKI

    2015-10-01

    Full Text Available There are many fuel quality standards introduced by national organizations and fuel producers. Usual techniques for measuring fuel parameters like cetane number, cetane index, fraction composition, viscosity, density, and flash point, require relatively complex and expensive laboratory equipment. On the fuel user side, fast and low cost sensing of useful state of biodiesel fuel is important. The main parameters of diesel fuel compatibility are: density, viscosity and surface tension. These three parameters define indirectly the quality of the fuel atomization process and the injected portion of energy that affect the quality of the fuel. In the presented paper the purposefulness of fuel testing using measurements of separable parameters is discussed. On this base, a sensor which enables the examination of relation of the mentioned parameters in one arrangement is proposed, analyzed and tested. The sensor uses the dynamic capillary rise method with photonic multichannel data reading in an inclined capillary. The principle of the sensor’s operation, the construction of the sensor head, and the experimental results are presented. The capillary is a disposable element. The sensor testing was performed with freshly prepared biodiesel fuels, and fuels stored for 2 years. We conclude that the proposed construction may be in future the base of low cost commercially marketable instruments for basic fuel classification: fit for use or not. That classification includes initial fuel composition and fuel parameters change during storage. Therefore, the proposed sensor is intended to use in fuel buying/selling point rather than used as part of a diesel engine automated system.

  16. Survey of European LWR fuel irradiation test facilities

    International Nuclear Information System (INIS)

    Hardt, P. von der

    1983-01-01

    The first European commercial nuclear power plants (1956) featured gas-cooled thermal reactors. Although there is now a general orientation towards light water cooled plants (with a slight preference for the PWR) a large fraction of the 1982 nuclear generating capacity is still invested in gas-cooled reactors. R and D also continues for the HTGR with its long-term development potential. This paper, however, is limited to a general survey of experimental programmes and facilities for light water reactor fuel testing in Western Europe, particularly inside the European Communities. As it turns out, over a dozen major installations are available, all connected to research reactors in government-funded R and D centres. Their equipment is briefly reviewed. Some 50% of the experimental programmes are carried out in large international collaboration, involving up to 20 organizations per project. Techniques and results are rapidly communicated through frequent meetings and conferences. It is anticipated that a part of the present research reactor-based work will gradually shift to power reactor pool side inspection facilities. (author)

  17. Results from Nevada Nuclear Waste Storage Investigations (NNWSI) Series 3 spent fuel dissolution tests

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1990-06-01

    The dissolution and radionuclide release behavior of spent fuel in groundwater is being studied by the Yucca Mountain Project (YMP), formerly the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. Specimens prepared from pressurized water reactor fuel rod segments were tested in sealed stainless steel vessels in Nevada Test Site J-13 well water at 85 degree C and 25 degree C. The test matrix included three specimens of bare-fuel particles plus cladding hulls, two fuel rod segments with artificially defected cladding and water-tight end fittings, and an undefected fuel rod section with watertight end fittings. Periodic solution samples were taken during test cycles with the sample volumes replenished with fresh J-13 water. Test cycles were periodically terminated and the specimens restarted in fresh J-13 water. The specimens were run for three cycles for a total test duration of 15 months. 22 refs., 32 figs., 26 tabs

  18. Post-test investigation result on the WWER-1000 fuel tested under severe accident conditions

    International Nuclear Information System (INIS)

    Goryachev, A.; Shtuckert, Yu.; Zwir, E.; Stupina, L.

    1996-01-01

    The model bundle of WWER-type were tested under SFD condition in the out-of-pile CORA installation. The objective of the test was to provide an information on the WWER-type fuel bundles behaviour under severe fuel damage accident conditions. Also it was assumed to compare the WWER-type bundle damage mechanisms with these experienced in the PWR-type bundle tests with aim to confirm a possibility to use the various code systems, worked our for PWR as applied to WWER. In order to ensure the possibility of the comparison of the calculated core degradation parameters with the real state of the tested bundle, some parameters have been measured on the bundle cross-sections under examination. Quantitative parameters of the bundle degradation have been evaluated by digital image processing of the bundle cross-sections. The obtained results are shown together with corresponding results obtained by the other participants of this investigation. (author). 3 refs, 13 figs

  19. Calculation of fuel element temperature TRIGA 2000 reactor in sipping test tubes using CFD

    International Nuclear Information System (INIS)

    Sudjatmi KA

    2013-01-01

    It has been calculated the fuel element temperature in the sipping test of Bandung TRIGA 2000 reactor. The calculation needs to be done to ascertain that the fuel element temperatures are below or at the limit of the allowable temperature fuel elements during reactor operation. ensuring that the implementation of the test by using this device, the temperature is still within safety limits. The calculation is done by making a model sipping test tubes containing a fuel element surrounded by 9 fuel elements. according to the position sipping test tubes in the reactor core. by using Gambit. Dimensional model adapted to the dimensions of the tube and the fuel element in the reactor core of Bandung TRIGA 2000 reactor. Sipping test Operation for each fuel element performed for 30 minutes at 300 kW power. Calculations were performed using CFD software and as input adjusted parameters of TRIGA 2000 reactor. Simulations carried out on the operation of the 30, 60, 90, 120, 150, 180 and 210 minutes. The calculation result shows that the temperature of the fuel in tubes sipping test of 236.06 °C, while the temperature of the wall is 87.58 °C. The maximum temperature in the fuel center of TRIGA 2000 reactor in normal operation is 650 °C. and the boiling is not allowed in the reactor. So it can be concluded that the operation of the sipping test device are is very safe because the fuel center temperature is below the temperature limits the allowable fuel under normal operating conditions as well as the fuel element wall temperature is below the boiling temperature of water. (author)

  20. Improving the AGR Fuel Testing Power Density Profile Versus Irradiation-Time in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-05-01

    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 235U 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 235U is an important design parameter. One of the more important AGR fuel testing requirements is to maintain the peak fuel compact temperature close to 1250°C 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 235U in

  1. Acute aquatic toxicity of heavy fuel oils. Summary of relevant test data

    International Nuclear Information System (INIS)

    Comber, M.I.H.; Den Haan, K.; Djemel, N.; Eadsforth, C.V.; King, D.; Parkerton, T.; Paumen, M.L.; Dmytrasz, B.

    2011-12-01

    This report describes the experimental procedures and results obtained in acute ecotoxicity tests on several heavy fuel oil (HFO) samples. Water accommodated fractions (WAFs) of these samples were tested for toxicity to the rainbow trout (Oncorhynchus mykiss), the crustacean zooplankter (Daphnia magna) and green algae (Selenastrum capricornutum). These results assist in determining the environmental hazard from heavy fuel oil.

  2. Acute aquatic toxicity of heavy fuel oils. Summary of relevant test data

    Energy Technology Data Exchange (ETDEWEB)

    Comber, M.I.H.; Den Haan, K.; Djemel, N.; Eadsforth, C.V.; King, D.; Parkerton, T.; Paumen, M.L.; Dmytrasz, B.

    2011-12-15

    This report describes the experimental procedures and results obtained in acute ecotoxicity tests on several heavy fuel oil (HFO) samples. Water accommodated fractions (WAFs) of these samples were tested for toxicity to the rainbow trout (Oncorhynchus mykiss), the crustacean zooplankter (Daphnia magna) and green algae (Selenastrum capricornutum). These results assist in determining the environmental hazard from heavy fuel oil.

  3. 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)

  4. U-Mo Monolithic Fuel for Nuclear Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Prabhakaran, Ramprashad

    2017-11-02

    The metallic fuel selected to replace the current HEU fuels in the research and test reactors is the LEU-10 weight % Mo alloy in the form of a thin sheet or foil encapsulated in AA6061 aluminum alloy with a zirconium interlayer. In order to effectively lead this pursuit, new developments in processing and fabrication of the fuel elements have been initiated, along with a better understanding of material behavior before and after irradiation as a result of these new developments. This editorial note gives an introduction about research and test reactors, need for HEU to LEU conversion, fuel requirements, high uranium density monolithic fuel development and an overview of the four articles published in the December 2017 issue of JOM under a special topic titled “U-Mo Monolithic Fuel for Nuclear Research and Test Reactors”.

  5. 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

  6. Assessment of Startup Fuel Options for a Test or Demonstration Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, Jon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hayes, Steven [Idaho National Lab. (INL), Idaho Falls, ID (United States); Walters, L. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    This document explores startup fuel options for a proposed test/demonstration fast reactor. The fuel options considered are the metallic fuels U-Zr and U-Pu-Zr and the ceramic fuels UO2 and UO2-PuO2 (MOX). Attributes of the candidate fuel choices considered were feedstock availability, fabrication feasibility, rough order of magnitude cost and schedule, and the existing irradiation performance database. The reactor-grade plutonium bearing fuels (U-Pu-Zr and MOX) were eliminated from consideration as the initial startup fuels because the availability and isotopics of domestic plutonium feedstock is uncertain. There are international sources of reactor grade plutonium feedstock but isotopics and availability are also uncertain. Weapons grade plutonium is the only possible source of Pu feedstock in sufficient quantities needed to fuel a startup core. Currently, the available U.S. source of (excess) weapons-grade plutonium is designated for irradiation in commercial light water reactors (LWR) to a level that would preclude diversion. Weapons-grade plutonium also contains a significant concentration of gallium. Gallium presents a potential issue for both the fabrication of MOX fuel as well as possible performance issues for metallic fuel. Also, the construction of a fuel fabrication line for plutonium fuels, with or without a line to remove gallium, is expected to be considerably more expensive than for uranium fuels. In the case of U-Pu-Zr, a relatively small number of fuel pins have been irradiated to high burnup, and in no case has a full assembly been irradiated to high burnup without disassembly and re-constitution. For MOX fuel, the irradiation database from the Fast Flux Test Facility (FFTF) is extensive. If a significant source of either weapons-grade or reactor-grade Pu became available (i.e., from an international source), a startup core based on Pu could be reconsidered.

  7. 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

  8. Fabrication and Testing of CERMET Fuel Materials for Nuclear Thermal Propulsion

    Science.gov (United States)

    Hickman, Robert; Broadway, Jeramie; Mireles, Omar

    2012-01-01

    A first generation Nuclear Cryogenic Propulsion Stage (NCPS) based on Nuclear Thermal Propulsion (NTP) is currently being developed for Advanced Space Exploration Systems. The overall goal of the project is to address critical NTP technology challenges and programmatic issues to establish confidence in the affordability and viability of NTP systems. The current technology roadmap for NTP identifies the development of a robust fuel form as a critical near term need. The lack of a qualified nuclear fuel is a significant technical risk that will require a considerable fraction of program resources to mitigate. Due to these risks and the cost for qualification, the development and selection of a primary fuel must begin prior to Authority to Proceed (ATP) for a specific mission. The fuel development is a progressive approach to incrementally reduce risk, converge the fuel materials, and mature the design and fabrication process of the fuel element. A key objective of the current project is to advance the maturity of CERMET fuels. The work includes fuel processing development and characterization, fuel specimen hot hydrogen screening, and prototypic fuel element testing. Early fuel materials development is critical to help validate requirements and fuel performance. The purpose of this paper is to provide an overview and status of the work at Marshall Space Flight Center (MSFC).

  9. National Fuel Cell Bus Program : Accelerated Testing Report, AC Transit

    Science.gov (United States)

    2009-01-01

    This is an evaluation of hydrogen fuel cell transit buses operating at AC Transit in revenue service since March 20, 2006 compared to similar diesel buses operating from the same depot. This evaluation report includes results from November 2007 throu...

  10. Technical concept for test of geologic storage of spent reactor fuel in the Climax granite, Nevada Test Site

    International Nuclear Information System (INIS)

    Ramspott, L.D.; Ballou, L.B.; Carlson, R.C.; Montan, D.N.; Butkovich, T.R.; Duncan, J.E.; Patrick, W.C.; Wilder, D.G.; Brough, W.G.; Mayr, M.C.

    1979-01-01

    The Spent Fuel Test in the Climax granite at the Nevada Test Site is a generic test in which spent fuel assemblies from an operating commercial nuclear reactor are emplaced at, and retrieved from, a plausible waste repository depth in a typical granite. Eleven canisters of spent fuel are emplaced in a storage drift 420 m below the surface along with six electrical simulator canisters. Two adjacent drifts contain electrical heaters which are operated so as to simulate the initial five years of the temperature-stress-displacement fields of a large repository. The site is described, and the pre-operational measurement program and characteristics of the spent fuel are given. Both thermal and mechanical response calculations are summarized. The field instrumentation and data acquisition systems are described, as well as the system for handling the spent fuel

  11. Evaluation of burnup characteristics and energy deposition during NSRR pulse irradiation tests on irradiated BWR fuels

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio

    2000-11-01

    Pulse irradiation tests of irradiated fuel are performed in the Nuclear Safety Research Reactor (NSRR) to investigate the fuel behavior under Reactivity Initiated Accident Conditions (RIA). The severity of the RIA is represented by energy deposition or peak fuel enthalpy during the power excursion. In case of the irradiated fuel tests, the energy deposition varies depending both on the amounts and distribution of residual fissile and neutron absorbing fission products generated during the base irradiation. Thus, proper fuel burnup characterization, especially for low enriched commercial fuels, is important, because plutonium (Pu) takes a large part of fissile and its generation depends on the neutron spectrum during the base irradiation. Fuel burnup calculations were conducted with ORIGEN2, RODBURN and SWAT codes for the BWR fuels tested in the NSRR. The calculation results were compared with the measured isotope concentrations and used for the NSRR neutron calculations to evaluate energy depositions of the test fuel. The comparison of the code calculations and the measurements revealed that the neutron spectrum change due to difference in void fraction altered Pu generation and energy deposition in the NSRR tests considerably. With the properly evaluated neutron spectrum, the combined burnup and NSRR neutron calculation gave reasonably good evaluation of the energy deposition. The calculations provided radial distributions of the fission product accumulation during the base irradiation and power distribution during the NSRR pulse irradiation, which were important for the evaluation of both burnup characteristics and fission gas release behavior. (author)

  12. Fabrication and Testing of Full-Length Single-Cell Externally Fueled Converters for Thermionic Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Schock, Alfred

    1995-08-01

    Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes the fabrication and testing of full-length prototypcial converters, both unfueled and fueled, and presents parametric results of electrically heated tests.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-04-01

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

  14. FFTF [Fast Flux Test Facility]/IEM [Interim Examination and Maintenance] Cell Fuel Pin Weighing System

    International Nuclear Information System (INIS)

    Gibbons, P.W.

    1987-09-01

    A Fuel Pin Weighing Machine has been developed for use in the Fast Flux Test Facility (FFTF) Interim Examination and Maintenance (IEM) Cell to assist in identifying an individual breached fuel pin from its fuel assembly pin bundle. A weighing machine, originally purchased for use in the Fuels and Materials Examination Facility (FMEF) at Hanford, was used as the basis for the IEM Cell system. Design modifications to the original equipment were centered around: 1) adapting the FMEF machine for use in the IEM Cell and 2) correcting operational deficiencies discovered during functional testing in the IEM Cell Mockup

  15. Post-irradiation examination of Oconee 1 fuel: end-of-cycle 2 nondestructive test phase

    International Nuclear Information System (INIS)

    1979-11-01

    Standard B and W Mark B (15 x 15) pressurized water reactor fuel assemblies were nondestructively examined at the end of the second cycle of Oconee 1 reactor operation. Burnups of the 16 fuel assemblies examined ranged from 13,100 to 20,000 MWd/mtU. The examinations were conducted in the Oconee 1 and 2 spent fuel storage pool using the installed underwater test equipment. Data obtained included fuel rod and fuel assembly dimensions, water channel spacings, holddown spring forces, fuel rod crud characteristics, and fuel column axial gap and stack lengths. Visual examinations revealed no evidence of significant rod bowing, cladding deformation, cocked grids, or rod defects. The results, summarized in this report, indicate that the assemblies performed well through two cycles of reactor operation

  16. FFTF [fast flux tests facility] metal fuel pin sodium bond quality verification

    International Nuclear Information System (INIS)

    Pitner, A.L.

    1989-01-01

    The Fast Flux Tests Facility (FFTF) Series III driver fuel design consists of U-10 Zr fuel slugs contained in a ferritic alloy cladding. A liquid-metal, sodium bond between the fuel and cladding is required to prevent unacceptable temperatures during operation. Excessive voiding or porosity in the sodium thermal bond could result in localized fuel melting during irradiation. It is therefore imperative that bond quality be verified during fabrication of these metal fuel pins prior to irradiation. Cladding stripping of two spare metal fuel pins was successfully performed to inspect the quality of the sodium thermal bond and to compare observed defects in the bond with eddy current measurements. Bond quality was found to be generally good, thereby confirming the fabrication process used to bond the fuel and cladding with liquid-metal sodium to be acceptable. Observed porosity distribution in the sodium bond correlated well with eddy current indications

  17. Recent irradiation tests of uranium-plutonium-zirconium metal fuel elements

    International Nuclear Information System (INIS)

    Pahl, R.G.; Lahm, C.E.; Villarreal, R.; Hofman, G.L.; Beck, W.N.

    1986-09-01

    Uranium-Plutonium-Zirconium metal fuel irradiation tests to support the ANL Integral Fast Reactor concept are discussed. Satisfactory performance has been demonstrated to 2.9 at.% peak burnup in three alloys having 0, 8, and 19 wt % plutonium. Fuel swelling measurements at low burnup in alloys to 26 wt % plutonium show that fuel deformation is primarily radial in direction. Increasing the plutonium content in the fuel diminishes the rate of fuel-cladding gap closure and axial fuel column growth. Chemical redistribution occurs by 2.1 at.% peak burnup and generally involves the inward migration of zirconium and outward migration of uranium. Fission gas release to the plenum ranges from 46% to 56% in the alloys irradiated to 2.9 at.% peak burnup. No evidence of deleterious fuel-cladding chemical or mechanical interaction was observed

  18. Design verification of the CANFLEX fuel bundle - quality assurance requirements for mechanical flow testing

    International Nuclear Information System (INIS)

    Alavi, P.; Oldaker, I.E.; Chung, C.H.; Suk, H.C.

    1997-01-01

    As part of the design verification program for the new fuel bundle, a series of out-reactor tests was conducted on the CANFLEX 43-element fuel bundle design. These tests simulated current CANDU 6 reactor normal operating conditions of flow, temperature and pressure. This paper describes the Quality Assurance (QA) Program implemented for the tests that were run at the testing laboratories of Atomic Energy of Canada Limited (AECL) and Korea Atomic energy Research Institute (KAERI). (author)

  19. Fuel Testing for Sylvatex: Cooperative Research and Development Final Report, CRADA Number CRD-16-636

    Energy Technology Data Exchange (ETDEWEB)

    Burton, Jonathan L. [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2018-02-07

    Sylvatex is a green nano-chemistry company that has developed a platform technology utilizing renewable, non-toxic inputs to create a stable nanoparticle that can be used in multiple applications. Their mission is to increase the use of renewables globally, to empower a cleaner and healthier future. The main application is a fuel technology product - MicroX - that utilizes proprietary knowledge to scale low-cost, cleaner-burning renewable diesel fuel and additives by using a co-location commercial model. The aspects of this project will include testing of two Sylvatex MicroX fuels on an engine dynamometer platform. Industry standard ultra-low sulfur diesel (ULSD) B3 fuel and a ULSD B20 will both be used for comparison of the Sylvatex fuels (U.S. standard diesel fuel at the pump contains an average of approximately 3% biodiesel; this is why B3 would be used as a baseline comparison). Sylvatex is currently using a prototype formulation (MicroX 1) that applies a high cost surfactant. An experimental formulation (MicroX 2) that uses lower cost materials is under development. The MicroX 1 will be blended at a 10% level into the B3 ULSD fuel and the MicroX 2 will be blended at a 10% level into both the B3 and the B20 ULSD fuels for study on the engine dynamometer test platform. All fuel blends will be tested over the FTP transient engine test cycle and a steady state ramped modal engine test cycle. Each test cycle will be performed a minimum of 3 times for each fuel. Tailpipe and/or engine out gaseous exhaust emissions (CO2, CO, NOx, THC, O2,), engine out PM emissions, and brake-specific fuel consumption rates will be evaluated for all test cycles.

  20. Field testing at the Climax Stock on the Nevada Test Site: spent fuel test and radionuclide migration experiments

    International Nuclear Information System (INIS)

    Ballou, L.B.; Isherwood, D.J.; Patrick, W.C.

    1982-01-01

    Two field tests in the Climax Stock are being conducted. The Climax Stock, a granitic instrusive, has been administratively excluded from consideration as a full-scale repository site. However, it provides a readily available facility for field testing with high-level radioactive materials at a depth (420 m) approaching that of a repository. The major test activity in the 1980 fiscal year has been initiation of the Spent Fuel Test-Climax (SFT-C). This test, which was authorized in June 1978, is designed to evaluate the generic feasibility of geologic storage and retrievability of commercial power reactor spent fuel assemblies in a granitic medium. In addition, the test is configured and instrumented to provide thermal and thermomechanical response data that will be relevant to the design of a repository in hard crystalline rock. The other field activity in the Climax Stock is a radionuclide migration test. It combines a series of field and laboratory migration experiments with the use of existing hydrologic models for pretest predictions and data interpretation. Goals of this project are to develop: (1) field measurement techniques for radionuclide migration studies in a hydrologic regime where the controlling mechanism is fracture permeability; (2) field test data on radionuclide migration; and (3) a comparison of laboratory- and field-measured retardation factors. This radionuclide migration test, which was authorized in the middle of the 1980 fiscal year, is in the preliminary design phase. The detailed program plan was prepared and subjected to formal peer review in August. In September/October researchers conducted preliminary flow tests with water in selected near-vertical fractures intersected by small horizontal boreholes. These tests were needed to establish the range of pressures, flow rates, and other operating parameters to be used in conducting the nuclide migration tests. 21 references, 14 figures, 1 table

  1. 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.

  2. Heat and mass release for some transient fuel source fires: A test report

    International Nuclear Information System (INIS)

    Nowlen, S.P.

    1986-10-01

    Nine fire tests using five different trash fuel source packages were conducted by Sandia National Laboratories. This report presents the findings of these tests. Data reported includes heat and mass release rates, total heat and mass release, plume temperatures, and average fuel heat of combustion. These tests were conducted as a part of the US Nuclear Regulatory Commission sponsored fire safety research program. Data from these tests were intended for use in nuclear power plant probabilistic risk assessment fire analyses. The results were also used as input to a fire test program at Sandia investigating the vulnerability of electrical control cabinets to fire. The fuel packages tested were chosen to be representative of small to moderately sized transient trash fuel sources of the type that would be found in a nuclear power plant. The highest fire intensity encountered during these tests was 145 kW. Plume temperatures did not exceed 820 0 C

  3. ROSA-IV Large Scale Test Facility (LSTF) system description for second simulated fuel assembly

    International Nuclear Information System (INIS)

    1990-10-01

    The ROSA-IV Program's Large Scale Test Facility (LSTF) is a test facility for integral simulation of thermal-hydraulic response of a pressurized water reactor (PWR) during small break loss-of-coolant accidents (LOCAs) and transients. In this facility, the PWR core nuclear fuel rods are simulated using electric heater rods. The simulated fuel assembly which was installed during the facility construction was replaced with a new one in 1988. The first test with this second simulated fuel assembly was conducted in December 1988. This report describes the facility configuration and characteristics as of this date (December 1988) including the new simulated fuel assembly design and the facility changes which were made during the testing with the first assembly as well as during the renewal of the simulated fuel assembly. (author)

  4. Performance testing of refractory alloy-clad fuel elements for space reactors

    International Nuclear Information System (INIS)

    Dutt, D.S.; Cox, C.M.; Karnesky, R.A.; Millhollen, M.K.

    1985-01-01

    Two fast reactor irradiation tests, SP-1 and SP-2, provide a unique and self-consistent data set with which to evaluate the technical feasibility of potential fuel systems for the SP-100 space reactor. Fuel pins fabricated with leading cladding candidates (Nb-1Zr, PWC-11, and Mo-13Re) and fuel forms (UN and UO 2 ) are operated at temperatures typical of those expected in the SP-100 design. The first US fast reactor irradiated, refractory alloy clad fuel pins, from the SP-1 test, reached 1 at. % burnup in EBR-II in March 1985. At that time selected pins were discharged for interim examination. These examinations confirmed the excellent performance of the Nb-1Zr clad uranium oxide and uranium nitride fuel elements, which are the baseline fuel systems for two SP-100 reactor concepts

  5. 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)

  6. Analysis of Ignition Testing on K-West Basin Fuel

    Energy Technology Data Exchange (ETDEWEB)

    J. Abrefah; F.H. Huang; W.M. Gerry; W.J. Gray; S.C. Marschman; T.A. Thornton

    1999-08-10

    Approximately 2100 metric tons of spent nuclear fuel (SNF) discharged from the N-Reactor have been stored underwater at the K-Basins in the 100 Area of the Hanford Site. The spent fuel has been stored in the K-East Basin since 1975 and in the K-West Basin since 1981. Some of the SNF elements in these basins have corroded because of various breaches in the Zircaloy cladding that occurred during fuel discharge operations and/or subsequent handling and storage in the basins. Consequently, radioactive material in the fuel has been released into the basin water, and water has leaked from the K-East Basin into the soil below. To protect the Columbia River, which is only 380 m from the basins, the SNF is scheduled to be removed and transported for interim dry storage in the 200 East Area, in the central portion of the Site. However, before being shipped, the corroded fuel elements will be loaded into Multi-Canister OverPacks and conditioned. The conditioning process will be selected based on the Integrated Process Strategy (IPS) (WHC 1995), which was prepared on the basis of the dry storage concept developed by the Independent Technical Assessment (ITA) team (ITA 1994).

  7. Reactor tests and post-reactor examination of HTGR fuel elements and coated particles

    International Nuclear Information System (INIS)

    Degaltsev, Yu.G.; Khrulev, A.A.; Mosevitskii, I.S.; Ponomarev-Stepnoi, N.N.; Tikhonov, N.I.; Yakovlev, V.V.

    1990-01-01

    Tests and examinations of HTGR fuel elements performed recently at I.V.Kurchatov AEI have been directed to solve the following problems: determination of correspondence of the operability characteristics of fuel elements and coated particles from different lots to requirements to be imposed on VGM and VG-400 designs being developed; comparison of the characteristics of fuel elements manufactured using two different technologies of matrix graphite (KPD and GSP) as well as coated particles of various designs and material composition; investigation of fuel element operability under severe accident conditions; investigation of processes occurring in fuel element and coated particle, resulting in formation of defects and release of GFP; construction of physical models of fuel element behaviour in the reactor on this basis. Tests and investigations were performed for fuel elements manufactured using the KPD and GSP technologies. For solving the above problems the experimental basis and methods described herein were used. The present paper is a short review of works covering these problems, with main emphasis being made on description and analysis of fuel element and coated particle tests under non-nominal accident and severe accident conditions. Such tests are of specific interest. While tests for confirming operability under nominal conditions are necessary for justification of the design characteristics and must be supported by sufficient statistics, tests for severe conditions have at least two objectives: 1) estimation of the limits of fuel element operability depending on the basing affecting factors (temperatures burnup, etc.); 2) getting the data for studying processes resulting in loss of operability; on the basis of these data physical models of fuel element operability and accident sequence in the reactor unit may be constructed and verified. The experiments presented herein are the first stage of such investigations on determination of influence of the

  8. Development of OTM Syngas Process and Testing of Syngas Derived Ultra-clean Fuels in Diesel Engines and Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    E.T. Robinson; John Sirman; Prasad Apte; Xingun Gui; Tytus R. Bulicz; Dan Corgard; John Hemmings

    2005-05-01

    This final report summarizes work accomplished in the Program from January 1, 2001 through December 31, 2004. Most of the key technical objectives for this program were achieved. A breakthrough material system has lead to the development of an OTM (oxygen transport membrane) compact planar reactor design capable of producing either syngas or hydrogen. The planar reactor shows significant advantages in thermal efficiency and a step change reduction in costs compared to either autothermal reforming or steam methane reforming with CO{sub 2} recovery. Syngas derived ultra-clean transportation fuels were tested in the Nuvera fuel cell modular pressurized reactor and in International Truck and Engine single cylinder test engines. The studies compared emission and engine performance of conventional base fuels to various formulations of ultra-clean gasoline or diesel fuels. A proprietary BP oxygenate showed significant advantage in both applications for reducing emissions with minimal impact on performance. In addition, a study to evaluate new fuel formulations for an HCCI engine was completed.

  9. Optical superheat of a surface as applied to investigation of the onset of boiling of cryogenic liquids

    International Nuclear Information System (INIS)

    Kopytov, G.F.; Novikov, B.V.; Shklyaev, A.G.

    1992-01-01

    The cryogenic liquid boiling onset indication is of interest in special cryotechnique applications, in particular, in preventive treatment of energency situation in the tanks for cryogenic liquid storage. Acoustic indication of the liquid boiling onset in case of its application for cryogenic liquids, which is widely used during the last decades, encounters technical difficulties related to the minor value of an acoustic signal, generated by the detectors used, low information content of high-frequency part of an acoustic spectrum due to signal blocking by gas bubbles just at the beginning of boiling etc. On the other hand, there is an opinion concerning the impossibility of using thermocouples under liquid boiling. A one-thermocouple method of cryogenic liquid boiling onset indication is proposed, assuming that local variations of the surface superheat temperature at the beginning of boiling, inevitable under boiling, make up a negligeable share of the mean superheat value

  10. Final Report - Spent Nuclear Fuel Retrieval System Manipulator System Cold Validation Testing

    Energy Technology Data Exchange (ETDEWEB)

    D.R. Jackson; G.R. Kiebel

    1999-08-24

    Manipulator system cold validation testing (CVT) 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. The FRS is composed of three major subsystems. The Manipulator Subsystem provides remote handling of fuel, scrap, and debris; the In-Pool Equipment subsystem performs cleaning of fuel and provides a work surface for handling materials; and the Remote Viewing Subsystem provides for remote viewing of the work area by operators. There are two complete and identical FRS systems, one to be installed in the K-West basin and one to be installed in the K-East basin. Another partial system will be installed in a cold test facility to provide for operator training.

  11. Final Report - Spent Nuclear Fuel Retrieval System Manipulator System Cold Validation Testing

    International Nuclear Information System (INIS)

    D.R. Jackson; G.R. Kiebel

    1999-01-01

    Manipulator system cold validation testing (CVT) 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. The FRS is composed of three major subsystems. The Manipulator Subsystem provides remote handling of fuel, scrap, and debris; the In-Pool Equipment subsystem performs cleaning of fuel and provides a work surface for handling materials; and the Remote Viewing Subsystem provides for remote viewing of the work area by operators. There are two complete and identical FRS systems, one to be installed in the K-West basin and one to be installed in the K-East basin. Another partial system will be installed in a cold test facility to provide for operator training

  12. Affordable Development and Optimization of CERMET Fuels for NTP Ground Testing

    Science.gov (United States)

    Hickman, Robert R.; Broadway, Jeramie W.; Mireles, Omar R.

    2014-01-01

    CERMET fuel materials for Nuclear Thermal Propulsion (NTP) are currently being developed at NASA's Marshall Space Flight Center. The work is part of NASA's Advanced Space Exploration Systems Nuclear Cryogenic Propulsion Stage (NCPS) Project. The goal of the FY12-14 project is to address critical NTP technology challenges and programmatic issues to establish confidence in the affordability and viability of an NTP system. A key enabling technology for an NCPS system is the fabrication of a stable high temperature nuclear fuel form. Although much of the technology was demonstrated during previous programs, there are currently no qualified fuel materials or processes. The work at MSFC is focused on developing critical materials and process technologies for manufacturing robust, full-scale CERMET fuels. Prototypical samples are being fabricated and tested in flowing hot hydrogen to understand processing and performance relationships. As part of this initial demonstration task, a final full scale element test will be performed to validate robust designs. The next phase of the project will focus on continued development and optimization of the fuel materials to enable future ground testing. The purpose of this paper is to provide a detailed overview of the CERMET fuel materials development plan. The overall CERMET fuel development path is shown in Figure 2. The activities begin prior to ATP for a ground reactor or engine system test and include materials and process optimization, hot hydrogen screening, material property testing, and irradiation testing. The goal of the development is to increase the maturity of the fuel form and reduce risk. One of the main accomplishmens of the current AES FY12-14 project was to develop dedicated laboratories at MSFC for the fabrication and testing of full length fuel elements. This capability will enable affordable, near term development and optimization of the CERMET fuels for future ground testing. Figure 2 provides a timeline of the

  13. Post-test thermal calculations and data analyses for the Spent Fuel Test, Climax

    International Nuclear Information System (INIS)

    Montan, D.N.; Patrick, W.C.

    1986-06-01

    After the Spent Fuel Test - Climax (SFT-C) was completed, additional calculations were performed using the best available (directly measured or inferred from measurements made during the test) input parameters, thermal properties, and power levels. This report documents those calculations and compares the results with measurements made during the three-year heating phase and six-month posttest cooling phase of the SFT-C. Three basic types of heat-transfer calculations include a combined two-dimensional/three-dimensional, infinite-length, finite-difference model; a fully three-dimensional, finite-length, finite-difference model; and a fully three-dimensional, finite-length, analytical solution. The finite-length model much more accurately reflects heat flow near the ends of the array and produces cooler temperatures everywhere than does its infinite-length counterpart. 14 refs., 144 figs., 4 tabs

  14. Re-fabrication and re-instrumentation of irradiated LWR fuel rods for irradiation testing at the HFR Petten

    International Nuclear Information System (INIS)

    Fischer, B.; Markgraf, J.F.W.; Puschek, P.; Duijves, K.A.; Haan, K.W. de

    1996-01-01

    LWR fuel testing at the High Flux Reactor (HFR) and the Hot Cells at Petten has been successfully performed with pre-irradiated fuel rod segments. The testing methods have been extended with hot cell techniques for re-fabrication of test fuel rods from full length fuel rods from power reactors; re-instrumentation of pre-irradiated fuel rod segments with pressure sensors; and instrumentation of re-fabricated fuel rods or fuel rod segments with central thermocouple and/or pressure sensors. 5 refs, 9 figs, 5 tabs

  15. Design criteria -- Modification of fuel element test facilities. 1706-KER Project CGI-839

    Energy Technology Data Exchange (ETDEWEB)

    Rudock, E.R.

    1959-08-27

    The following criteria outlines the basis, objectives, and fundamental methods that shall govern the preparation of final design for ``Project CGI-839, Modification to Fuel Element Test Facilities -- 1706 KER.`` These modifications will provide the equipment to test NPR size fuel elements in the KER recirculating loops. The 1706-KER Recirculation Test Facility of KE Reactor is operated to obtain experimental data regarding high temperature reactor coolant technology and high temperature fuel element testing. The facility consists of four in-pile recirculating loops. These loops will permit testing of fuel elements with the existing process tubes of 2.1 inches I.D. To provide adequate in-reactor fuel element test facilities to support the development of NPR fuel, two KER loops, {number_sign}3 and {number_sign}4 will be converted to provide a process tube of 2.7 inches ID that will be operated at typical NPR irradiation conditions. The remaining loops No. 1 and 2, will be modified to provide additional flow and heat transfer capacity for greater flexibility in the testing of high temperature fuel elements smaller than the NPR size. New pumps, heat exchangers, and minor piping modifications will be required in all loops.

  16. Accelerated testing of fuel cell components in 2 x 2 inch fuel cells

    International Nuclear Information System (INIS)

    Coleman, A.J.; Adams, A.A.; Joebstl, J.A.; Walker, G.W.

    1981-01-01

    A description is presented of diagnostic procedures which can be used to predict failure modes and assess the effects of these failures on fuel cell performance. Some straightforward diagnostic techniques have been used to evaluate fuel cells assembled with a variety of matrix and electrode combinations. These techniques included accelerated on-off cycling, thermal cycling with H2/CO mixtures, and automatic polarization measurements. Information has been obtained concerning the effects of electrolyte management and catalyst poisoning on performance and lifetime characteristics of 2 x 2 in. single cells. The use of on-off cycling has shown that short-term fuel cell performance is generally unaffected by load changes and cycle sequence in 2 x 2 in. cells when electrolyte management is adequate. Dynamic polarization curves can be used instead of point by point steady-state plots without any loss in accuracy

  17. Microscale phase change of fuels for MEMS power applications

    Science.gov (United States)

    Haendler, Brenda

    As portable electronics technology advances, systems are becoming smaller and more energy intensive. While batteries are currently the only commercial power source for these applications, work is being done to create liquid fuel based portable power packs. These systems would leverage the higher energy density (W-hr/l) and specific energy (W-hr/kg) of liquid hydrocarbon fuels over available battery chemistries. For micro engines and small fuels cells there are advantages to preheating and vaporizing the fuel in a microchannel. The work presented in this dissertation focuses on understanding and characterizing the temperature and pressure signatures that result from microscale boiling of fuels in etched silicon channels approximately 100 mm in diameter. Building on previous microscale boiling work which used water as the working fluid for electronics cooling applications, the studies presented in this dissertation use both water and fuels including methanol, ethanol, and octane. Results are presented in the form of pressure and temperature measurements for a range of working fluids, volumetric flow rates, superheat temperatures and channel geometries. From a Fourier transform analysis of the pressure signatures, it was found that the frequency of the pressure fluctuations increases with superheat for ethanol as the working fluid while for methanol the frequency increases with volumetric flow rate. Tests were also conducted with sudden expansion geometries, which reduce the amplitude of the pressure fluctuations and create a localized cooling in the working fluid. Results are compared using fluid properties, including surface tension and heat of vaporization, and non-dimensional numbers including the Weber and the Jakob number. This study presents a significant contribution to the body of knowledge on microscale boiling. One application of microscale boiling for portable power technologies is also presented. Fuel cracking, breaking apart of long hydrocarbon chain

  18. Failure Analysis and Magnetic Evaluation of Tertiary Superheater Tube Used in Gas-Fired Boiler

    Science.gov (United States)

    Mohapatra, J. N.; Patil, Sujay; Sah, Rameshwar; Krishna, P. C.; Eswarappa, B.

    2018-01-01

    Failure analysis was carried out on a prematurely failed tertiary superheater tube used in gas-fired boiler. The analysis includes a comparative study of visual examination, chemical composition, hardness and microstructure at failed region, adjacent and far to failure as well as on fresh tube. The chemistry was found matching to the standard specification, whereas the hardness was low in failed tube compared to the fish mouth opening region and the fresh tube. Microscopic examination of failed sample revealed the presence of spheroidal carbides of Cr and Mo predominantly along the grain boundaries. The primary cause of failure is found to be localized heating. Magnetic hysteresis loop (MHL) measurements were carried out to correlate the magnetic parameters with microstructure and mechanical properties to establish a possible non-destructive evaluation (NDE) for health monitoring of the tubes. The coercivity of the MHL showed a very good correlation with microstructure and mechanical properties deterioration enabling a possible NDE technique for the health monitoring of the tubes.

  19. Failure Analysis and Magnetic Evaluation of Tertiary Superheater Tube Used in Gas-Fired Boiler

    Science.gov (United States)

    Mohapatra, J. N.; Patil, Sujay; Sah, Rameshwar; Krishna, P. C.; Eswarappa, B.

    2018-02-01

    Failure analysis was carried out on a prematurely failed tertiary superheater tube used in gas-fired boiler. The analysis includes a comparative study of visual examination, chemical composition, hardness and microstructure at failed region, adjacent and far to failure as well as on fresh tube. The chemistry was found matching to the standard specification, whereas the hardness was low in failed tube compared to the fish mouth opening region and the fresh tube. Microscopic examination of failed sample revealed the presence of spheroidal carbides of Cr and Mo predominantly along the grain boundaries. The primary cause of failure is found to be localized heating. Magnetic hysteresis loop (MHL) measurements were carried out to correlate the magnetic parameters with microstructure and mechanical properties to establish a possible non-destructive evaluation (NDE) for health monitoring of the tubes. The coercivity of the MHL showed a very good correlation with microstructure and mechanical properties deterioration enabling a possible NDE technique for the health monitoring of the tubes.

  20. A computational approach for thermomechanical fatigue life prediction of dissimilarly welded superheater tubes

    Energy Technology Data Exchange (ETDEWEB)

    Krishnasamy, Ram-Kumar; Seifert, Thomas; Siegele, Dieter [Fraunhofer-Institut fuer Werkstoffmechanik (IWM), Freiburg im Breisgau (Germany)

    2010-07-01

    In this paper a computational approach for fatigue life prediction of dissimilarly welded superheater tubes is presented and applied to a dissimilar weld between tubes made of the nickel base alloy Alloy617 tube and the 12% chromium steel VM12. The approach comprises the calculation of the residual stresses in the welded tubes with a multi-pass dissimilar welding simulation, the relaxation of the residual stresses in a post weld heat treatment (PWHT) simulation and the fatigue life prediction using the remaining residual stresses as initial condition. A cyclic fiscoplasticity model is used to calculate the transient stresses and strains under thermocyclic service loadings. The fatigue life is predicted with a damage parameter which is based on fracture mechanics. The adjustable parameters of the model are determined based on LCF and TMF experiments. The simulations show, that the residual stresses that remain after PWHT further relax in the first loading cycles. The predicted fatigue lives depend on the residual stresses and, thus, on the choice of the loading cycle in which the damage parameter is evaluated. It the first loading cycle, where residual stresses are still present, is considered, lower fatigue lives are predicted compared to predictions considering loading cycles with relaxed residual stresses. (orig.)

  1. Spent Fuel Handling and Packaging Program Demonstration at the Nevada Test Site

    International Nuclear Information System (INIS)

    Bolmgren, C.R.

    1980-02-01

    The Spent Fuel Handling and Packaging Program Demonstration was initiated in 1977 to develop and test the capability to satisfactorily encapsulate typical spent fuel assemblies from commercial light water nuclear power plants and to establish the suitability of surface (Sealed Storage Cask) and near surface (drywell) concepts for the interim dry storage of the encapsulated spent fuel assemblies. The E-MAD facility at the Nevada Test Site was chosen as the location for this demonstration. This document describes the Sealed Storage Cask and drywell storage configuration, the processes and equipment utilized to encapsulate the spent fuel assemblies and place them into the storage configurations, and the thermal tests performed to establish the suitability of the storage configurations. Also presented is a criticality safety evaluation for the spent fuel handling operations and the storage configurations

  2. The Defect Inspection on the Irradiated Fuel Rod by Eddy Current Test

    International Nuclear Information System (INIS)

    Koo, D. S.; Park, Y. K.; Kim, E. K.

    1996-01-01

    The eddy current test(ECT) probe of differential encircling coil type was designed and fabricated, and the optimum condition of ECT was derived for the examination of the irradiated fuel rod. The correlation between ECT test frequency and phase and amplitude was derived by performing the test of the standard rig that includes inner notches, outer notches and through-holes. The defect of through-hole was predicted by ECT at the G33-N2 fuel rod irradiated in the Kori-1 nuclear power reactor. The metallographic examination on the G33-N2 fuel rod was Performed at the defect location predicted by ECT. The result of metallographic examination for the G33-N2 fuel rod was in good agreement with that of ECT. This proves that the evaluation for integrity of irradiated fuel rod by ECT is reliable

  3. Fuel containment and damage tolerance for large composite primary aircraft structures. Phase 1: Testing

    Science.gov (United States)

    Sandifer, J. P.

    1983-01-01

    Technical problems associated with fuel containment and damage tolerance of composite material wings for transport aircraft were identified. The major tasks are the following: (1) the preliminary design of damage tolerant wing surface using composite materials; (2) the evaluation of fuel sealing and lightning protection methods for a composite material wing; and (3) an experimental investigation of the damage tolerant characteristics of toughened resin graphite/epoxy materials. The test results, the test techniques, and the test data are presented.

  4. Achilles tests finally nail PWR fuel clad ballooning fears

    International Nuclear Information System (INIS)

    Dore, P.; McMinn, K.

    1992-01-01

    A conclusive series of experiments carried out by AEA Reactor Services at its Achilles rig in the UK has finally allayed fears that fuel clad ballooning is a major safety problem for Sizewell B, Britain's first Pressurized Water Reactor. The experiments are described in this article. (author)

  5. Development and Implementation of the National Test Facility (NaTeF) for Fuels and Propulsion

    Science.gov (United States)

    2013-10-01

    durability studies with new fuels and fuels at both elevated and low temperatures. • Development of streamlined testing processes that will enable...funding delay) Access to TGA training Never received Procurement of cured polymer slab stock 3 months Second shop air failure 12 months 31

  6. Durability Testing of Biomass Based Oxygenated Fuel Components in a Compression Ignition Engine

    Energy Technology Data Exchange (ETDEWEB)

    Ratcliff, Matthew A [National Renewable Energy Laboratory (NREL), Golden, CO (United States); McCormick, Robert L [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Baumgardner, Marc E. [Gonzaga University; Lakshminarayanan, Arunachalam [Colorado State University; Olsen, Daniel B. [Colorado State University; Marchese, Anthony J. [Colorado State University

    2017-10-18

    Blending cellulosic biofuels with traditional petroleum-derived fuels results in transportation fuels with reduced carbon footprints. Many cellulosic fuels rely on processing methods that produce mixtures of oxygenates which must be upgraded before blending with traditional fuels. Complete oxygenate removal is energy-intensive and it is likely that such biofuel blends will necessarily contain some oxygen content to be economically viable. Previous work by our group indicated that diesel fuel blends with low levels (<4%-vol) of oxygenates resulted in minimal negative effects on short-term engine performance and emissions. However, little is known about the long-term effects of these compounds on engine durability issues such as the impact on fuel injection, in-cylinder carbon buildup, and engine oil degradation. In this study, four of the oxygenated components previously tested were blended at 4%-vol in diesel fuel and tested with a durability protocol devised for this work consisting of 200 hrs of testing in a stationary, single-cylinder, Yanmar diesel engine operating at constant load. Oil samples, injector spray patterns, and carbon buildup from the injector and cylinder surfaces were analyzed. It was found that, at the levels tested, these fuels had minimal impact on the overall engine operation, which is consistent with our previous findings.

  7. Preparation results for lifetime test of conversion LEU fuel in plutonium production reactors

    International Nuclear Information System (INIS)

    Vatulin, A.; Stetskiy, Yu.; Kukharkin, N.; Kalougin, A.; Gavrilov, P.; Ivanov, A.

    1999-01-01

    The program of converting Russian production reactors for the purpose to stop their plutonium fabrication is currently in progress. The program also provides for operation of these reactors under the conversion mode with using of low-enriched fuel (LEU). LEU fuel elements were developed and activities related to their preparation for reactor tests were carried out. (author)

  8. Single-Cylinder Diesel Engine Tests with Unstabilized Water-in-Fuel Emulsions

    Science.gov (United States)

    1978-08-01

    A single-cylinder, four-stroke cycle diesel engine was operated on unstabilized water-in-fuel emulsions. Two prototype devices were used to produce the emulsions on-line with the engine. More than 350 test points were run with baseline diesel fuel an...

  9. Realization of an Electronic Load for Testing Low Power PEM Fuel Cells

    Directory of Open Access Journals (Sweden)

    Djordje Šaponjić

    2011-06-01

    Full Text Available A realized electronic load system intended for testing and characterization of hydrogen fuel sells is described. The system is based on microcontroller PIC16F877 by applying the concept of virtual instrumentation. The accomplished accuracy of the developed electronic system allows performing efficiently investigations of the electro-chemical phenomena involved in the process of designing hydrogen fuel cells.

  10. Evaluation and Testing of the Suitability of a Coal-Based Jet Fuel

    Science.gov (United States)

    2008-06-01

    with a total wattage of 7980 watts. Each oven section has two K type thermocouples per zone with Inconel sheathed spring loaded bayonet type mounts...also exceeded the thermal stability goals (525°F bulk and 625 °F WWT) for the JP-8+225 fuel program. Tests were conducted on a JP-8 fuel to compare

  11. Performance Degradation Tests of Phosphoric Acid Doped PBI Membrane Based High Temperature PEM Fuel Cells

    DEFF Research Database (Denmark)

    Zhou, Fan; Araya, Samuel Simon; Grigoras, Ionela

    2014-01-01

    Degradation tests of two phosphoric acid (PA) doped PBI membrane based HT-PEM fuel cells were reported in this paper to investigate the effects of start/stop and the presence of methanol in the fuel to the performance degradation. Continuous tests with H2 and simulated reformate which was composed...... to the redistribution of PA between the membrane and electrodes. EIS measurement of first fuel cell during the start/stop test showed that the mass transfer resistance and ohmic resistance increased which can be attributed to the corrosion of carbon support in the catalyst layer and degradation of the PBI membrane...

  12. 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

  13. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lanning, D.D.; Lombardo, N.J.; Hensley, W.K.; Fitzsimmons, D.E.; Panisko, F.E. [Pacific Northwest Lab., Richland, WA (United States); Hartwell, J.K. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    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 ({approximately}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.

  15. Investigation of the ramp testing behaviour of fuel pins with different diameters

    International Nuclear Information System (INIS)

    Pott, G.; Herren, M.; Wigger, B.

    1979-09-01

    The aim of these experiments was the investigation of the influence of different fuel pin diameter on the ramp testing behaviour. Fuel elements with diameter between 10,75 and 15,6 mm and different cladding thickness had been ramptested in the HBWR (Halden Boiling Water Reactor) after preirradiated in the same facility. Fuel pins with the smallest diameter of 10,75 mm failed. This was indicated by fission gas release measurement. Metallographic examination showed these failure were caused by hydride blisters. A systematic influence of fuel pin diameter and cladding thickness on the ramptesting behaviour was not observed. (orig.) [de

  16. E85 and fuel efficiency: An empirical analysis of 2007 EPA test data

    International Nuclear Information System (INIS)

    Roberts, Matthew C.

    2008-01-01

    It is well known that ethanol has less energy per unit volume than gasoline. Differences in engine design and fuel characteristics affect the efficiency with which the chemical energy in gasoline and ethanol is converted into mechanical energy, so that the change in fuel economy may not be a linear function of energy content. This study analyzes the fuel economy tests performed by the US Environmental Protection Agency (EPA) on 2007 model year E85-compliant vehicles and finds that the difference in average fuel economy is not statistically different from the differential in energy content

  17. Hydraulic test for non-instrumented capsule of advanced PWR fuel pellet

    International Nuclear Information System (INIS)

    Jun, Hyung Gil; Yoon, Y. J.; Chun, S. Y.; Kim, D. H.; Lee, C. B.; Ryu, J.

    2001-04-01

    This report presents the results of pressure drop test, vibration test and endurance test for Non-instrumented Capsule of Advanced PWR Fuel Pellet which were designed fabricated by KAERI. From the pressure drop test results, it is noted that the flow rate across the Non-instrumented Capsule of Advanced PWR Fuel Pellet corresponding to the pressure drop of 200 kPa is measured to be about 7.45 kg/sec. Vibration frequency for the Non-instrumented Capsule of Advanced PWR Fuel Pellet ranges from 13.0 to 32.3 Hz. RMS(Root Mean Square) displacement for the fuel rig is less than 11.6 μm, and the maximum displacement is less than 30.5 μm. The endurance test was carried out for 103 days and 17 hours

  18. Development And Testing Of Biogas-Petrol Blend As An Alternative Fuel For Spark Ignition Engine

    Directory of Open Access Journals (Sweden)

    Awogbemi

    2015-08-01

    Full Text Available Abstract This research is on the development and testing of a biogas-petrol blend to run a spark ignition engine. A2080 ratio biogaspetrol blend was developed as an alternative fuel for spark ignition engine test bed. Petrol and biogas-petrol blend were comparatively tested on the test bed to determine the effectiveness of the fuels. The results of the tests showed that biogas petrol blend generated higher torque brake power indicated power brake thermal efficiency and brake mean effective pressure but lower fuel consumption and exhaust temperature than petrol. The research concluded that a spark ignition engine powered by biogas-petrol blend was found to be economical consumed less fuel and contributes to sanitation and production of fertilizer.

  19. Technical concept for a test of geologic storage of spent reactor fuel in the climax granite, Nevada Test Site

    International Nuclear Information System (INIS)

    Ramspott, L.D.; Ballou, L.B.; Carlson, R.C.; Montan, D.N.; Butkovich, T.R.; Duncan, J.E.; Patrick, W.C.; Wilder, D.G.; Brough, W.G.; Mayr, M.C.

    1979-01-01

    We plan to emplace spent fuel assemblies from an operating commercial nuclear reactor in the Climax granite at the US Department of Energy's Nevada Test Site. In this generic test, 11 canisters of spent fuel will be emplaced with 6 electrical simulator canisters in a storage drift 420 m below in surface and their effects compared. Two adjacent drifts will contain electrical heaters, operated to simulate the temperature-stress-displacement fields of a large repository. We describe the test objectives, the technical issues, the site, the preoperational measurement program, thermal and mechanical response calculations, the characteristics of the spent fuel, the field instrumentation and data-acquisition systems, and the system for handling the spent fuel

  20. Status of irradiation testing and PIE of MOX (Pu-containing) fuel

    International Nuclear Information System (INIS)

    Dimayuga, F.C.; Zhou, Y.N.; Ryz, M.A.

    1995-01-01

    This paper describes AECL's mixed oxide (MOX) fuel-irradiation and post-irradiation examination (PIE) program. Post-irradiation examination results of two major irradiation experiments involving several (U, Pu)O 2 fuel bundles are highlighted. One experiment involved bundles irradiated to burnups ranging fro 400 to 1200 MWh/kgHe in the Nuclear Power Demonstration (NPD) reactor. The other experiment consisted of several (U, Pu)O 2 bundles irradiated to burnups of up to 500 Mwh/kgHe in the National Research Universal (NRU) reactor. Results of these experiments demonstrate the excellent performance of CANDU MOX fuel. This paper also outlines the status of current MOX fuel irradiation tests, including the irradiation of various (U, Pu)O 2 bundles. The strategic importance of MOX fuel to CANDU fuel-cycle flexibility is discussed. (author)

  1. Emission testing of jatropha and pongamia mixed bio diesel fuel in a diesel engine

    International Nuclear Information System (INIS)

    Ali, M.; Shaikh, A.A.

    2012-01-01

    The present investigation is based on the emission characteristics of mixed bio diesel fuel in a four stroke single cylinder compression ignition engine at constant speed. Refined oils of jatropha and pongamia are converted into bio diesel by acid catalyzed esterification and base catalyzed transesterification reactions. The jatropha and pongamia bio diesel were mixed in equal proportions with conventional mineral diesel fuel. Four samples of fuel were tested namely, diesel fuel, B10, B20 and B40. The emission analysis showed B20 mixed bio diesel fuel blend having better results as compared to other samples. There is 60% and 35% lower emission of carbon monoxide and in sulphur dioxide observed while consuming B20 blended fuel respectively. The test result showed NOx emissions were 10% higher from bio diesel fuel, as compared to conventional diesel fuel. However, these emissions may be reduced by EGR (Exhaust Gas Recirculation) technology. Present research also revealed that that B20 mixed bio diesel fuel can be used, without any modification in a CI engine. (author)

  2. Thermal analysis of lithium cooled natural circulation loop module for fuel rod testing in the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Eyler, L.L.; Kim, D.; Stover, R.L.; Beaver, T.R.

    1987-01-01

    Maximum heat removal capability of a lithium cooled natural circulation fuel rod test module design is determined. Loop geometry is optimized within limitations of design specifications for nominal operation temperatures, materials, and test module environment. Results provide test module operation limits and range of potential uncertainties. 3 refs., 12 figs

  3. Chemical Characterization and Reactivity Testing of Fuel-Oxidizer Reaction Product (Test Report)

    Science.gov (United States)

    1996-01-01

    The product of incomplete reaction of monomethylhydrazine (MMH) and nitrogen tetroxide (NTO) propellants, or fuel-oxidizer reaction product (FORP), has been hypothesized as a contributory cause of an anomaly which occurred in the chamber pressure (PC) transducer tube on the Reaction Control Subsystem (RCS) aft thruster 467 on flight STS-51. A small hole was found in the titanium-alloy PC tube at the first bend below the pressure transducer. It was surmised that the hole may have been caused by heat and pressure resulting from ignition of FORP. The NASA Johnson Space Center (JSC) White Sands Test Facility (WSTF) was requested to define the chemical characteristics of FORP, characterize its reactivity, and simulate the events in a controlled environment which may have lead to the Pc-tube failure. Samples of FORP were obtained from the gas-phase reaction of MMH with NTO under laboratory conditions, the pulsed firings of RCS thrusters with modified PC tubes using varied oxidizer or fuel lead times, and the nominal RCS thruster firings at WSTF and Kaiser-Marquardt. Fourier transform infrared spectroscopy (FTIR), differential scanning calorimetry (DSC), accelerating rate calorimetry (ARC), ion chromatography (IC), inductively coupled plasma (ICP) spectrometry, thermogravimetric analysis (TGA) coupled to FTIR (TGA/FTIR), and mechanical impact testing were used to qualitatively and quantitatively characterize the chemical, thermal, and ignition properties of FORP. These studies showed that the composition of FORP is variable but falls within a limited range of compositions that depends on the fuel loxidizer ratio at the time of formation, composition of the post-formation atmosphere (reducing or oxidizing), and reaction or postreaction temperature. A typical composition contains methylhydrazinium nitrate (MMHN), ammonium nitrate (AN), methylammonium nitrate (MAN), and trace amounts of hydrazinium nitrate and 1,1-dimethylhydrazinium nitrate. The thermal decomposition

  4. Results from Cycles 1 and 2 of NNWSI Series 2 spent fuel dissolution tests

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1987-05-01

    PWR spent fuel rod segments from the H.B. Robinson Unit 2 and Turkey Point Unit 3 reactors were leach tested in Nevada Nuclear Waste Storage Investigations (NNWSI) reference J-13 water under ambient hot cell conditions. The test matrix included bare fuel plus the cladding, rod segments with artificially induced cladding defects, and undefected rod segments. Radionuclide release results are presented and discussed. The actinides Pu, Am, Cm and Np appear to have been released congruently as the UO 2 oxide fuel matrix dissolved. Preferential U release measured in certain tests may be related to dissolution of oxidized UO/sub 2+x/ from the fuel surface, and/or greater solubility (and mobility) of U relative to the other actinides within defected cladding specimens. Uranium solubility measured in the J-13 water was much greater then that measured in deionized water in previous tests. All of the principal fission products analyzed ( 137 Cs, 129 I, 99 Tc and 90 Sr) were released preferentially relative to the actinides. Preferential release of activation product 14 C was also measured, with a portion of the 14 C release appearing to originate from the cladding exterior surface. Much greater fractional fuel dissolution appeared to have occurred with bare fuel particles than from fuel contained in defected cladding. Actinide release from test specimens containing small (∼200 μm) laser-drilled holes through the cladding was not significantly greater than from undefected specimens

  5. Integrity assessment of research reactor fuel cladding and material testing using eddy current inspection

    International Nuclear Information System (INIS)

    Alencar, Donizete Anderson de

    2004-01-01

    A methodology to perform the integrity assessment of research reactors nuclear fuels cladding, such as those installed in IPR-Rl (TRIGA) and IEA-R1 (MTR), using nondestructive electromagnetic inspection (eddy current) is presented. This methodology is constituted by: the development of calibration reference standards, specific for each type of fuel; the development of special test probes; the recommendations for the inspection equipment calibration; the construction of voltage based evaluation curves and the inspection procedures developed for the characterization of detected flaws. The test probes development, specially those designed for the inspection of MTR fuels cladding, which present access difficulties due to the narrow gap between fuel plates (2,89 mm for IEAR-R1), constituted a challenge that demanded the introduction of unusual materials and constructive techniques. The operational performance of the developed resources, as well as the special operative characteristics of the test probes, such as their immunity to adjacent fuel plates interference and electrical resistivity changes of the fuels meat are experimentally demonstrated. The practical applicability of the developed methodology is verified in non radioactive environment, using a dummy MTR fuel element model, similar to an IEA-R1 reactor fuel element, produced and installed in IPEN, Sao Paulo. The efficacy of the proposed methodology was verified by the achieved results. (author)

  6. 77 FR 16868 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

    Science.gov (United States)

    2012-03-22

    ... Fuel Elements for Use in Research and Test Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test Reactors.'' This guide describes... plate-type uranium-aluminum fuel elements used in research and test reactors (RTRs). DATES: Submit...

  7. An evaluation of propane as a fuel for testing fire-resistant oil spill containment booms

    International Nuclear Information System (INIS)

    Walton, W. D.; Twilley, W. H.

    1997-01-01

    A series of experiments have been conducted to measure and compare the thermal exposure to a fire-resistant boom from liquid hydrocarbon fuel and propane fires. The objective was to test the potential of propane fueled fires as a fire source for testing fire-resistant oil spill containment booms.Thermal exposure from propane fires have been measured with and without waves. Results indicated that although propane diffusion flames on water look like liquid hydrocarbon fuel flames and produce very little smoke, the heat flux at the boom location from propane fires is about 60 per cent of that from liquid hydrocarbon fuel fires. Despite the attractive features in terms of ease of application, control and smoke emissions, it was concluded that the low heat flux would preclude the application of propane as a fuel for evaluating fire resistant containment booms. 2 refs., 7 figs

  8. Used nuclear fuel separations process simulation and testing

    International Nuclear Information System (INIS)

    Pereira, C.; Krebs, J.F.; Copple, J.M.; Frey, K.E.; Maggos, L.E.; Figueroa, J.; Willit, J.L.; Papadias, D.D.

    2013-01-01

    Recent efforts in separations process simulation at Argonne have expanded from the traditional focus on solvent extraction flowsheet design in order to capture process dynamics and to simulate other components, processing and systems of a used nuclear fuel reprocessing plant. For example, the Argonne Model for Universal Solvent Extraction (AMUSE) code has been enhanced to make it both more portable and more readily extensible. Moving away from a spreadsheet environment makes the addition of new species and processes simpler for the expert user, which should enable more rapid implementation of chemical models that simulate evolving processes. The dyAMUSE (dynamic AMUSE) version allows the simulation of transient behavior across an extractor. Electrochemical separations have now been modeled using spreadsheet codes that simulate the electrochemical recycle of fast reactor fuel. The user can follow the evolution of the salt, products, and waste compositions in the electro-refiner, cathode processors, and drawdown as a function of fuel batches treated. To further expand capabilities in integrating multiple unit operations, a platform for linking mathematical models representing the different operations that comprise a reprocessing facility was adapted to enable systems-level analysis and optimization of facility functions. (authors)

  9. Visualization test facility of nuclear fuel rod emergency cooling system

    International Nuclear Information System (INIS)

    Candido, Marcos Antonio; Mesquita, Amir Zacarias; Rezende, Hugo Cesar; Santos, Andre Augusto Campagnole

    2013-01-01

    The nuclear reactors safety is determined according to their protection against the consequences that may result from postulated accidents. The Loss of Coolant Accident (LOCA) is one the most important design basis accidents (DBA). The failure may be due to rupture of the primary loop piping. Another accident postulated is due to lack of power in the pump motors in the primary circuit. In both cases the reactor shut down automatically due to the decrease of reactivity to maintain the fissions, and to the drop of control rods. In the event of an accident it is necessary to maintain the coolant flow to remove the fuel elements residual heat, which remains after shut down. This heat is a significant amount of the maximum thermal power generated in normal operation (about 7%). Recently this event has been quite prominent in the press due to the reactor accident in Fukushima nuclear power station. This paper presents the experimental facility under rebuilding at the Thermal Hydraulic Laboratory of the Nuclear Technology Development Center (CDTN) that has the objective of monitoring and visualization of the process of emergency cooling of a nuclear fuel rod simulator, heated by Joule effect. The system will help the comprehension of the heat transfer process during reflooding after a loss of coolant accident in the fuel of light water reactor core. (author)

  10. Used nuclear fuel separations process simulation and testing

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, C.; Krebs, J.F.; Copple, J.M.; Frey, K.E.; Maggos, L.E.; Figueroa, J.; Willit, J.L.; Papadias, D.D. [Argonne National Laboratory: 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2013-07-01

    Recent efforts in separations process simulation at Argonne have expanded from the traditional focus on solvent extraction flowsheet design in order to capture process dynamics and to simulate other components, processing and systems of a used nuclear fuel reprocessing plant. For example, the Argonne Model for Universal Solvent Extraction (AMUSE) code has been enhanced to make it both more portable and more readily extensible. Moving away from a spreadsheet environment makes the addition of new species and processes simpler for the expert user, which should enable more rapid implementation of chemical models that simulate evolving processes. The dyAMUSE (dynamic AMUSE) version allows the simulation of transient behavior across an extractor. Electrochemical separations have now been modeled using spreadsheet codes that simulate the electrochemical recycle of fast reactor fuel. The user can follow the evolution of the salt, products, and waste compositions in the electro-refiner, cathode processors, and drawdown as a function of fuel batches treated. To further expand capabilities in integrating multiple unit operations, a platform for linking mathematical models representing the different operations that comprise a reprocessing facility was adapted to enable systems-level analysis and optimization of facility functions. (authors)

  11. 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

  12. Fuel containment and damage tolerance in large composite primary aircraft structures. Phase 2: Testing

    Science.gov (United States)

    Sandifer, J. P.; Denny, A.; Wood, M. A.

    1985-01-01

    Technical issues associated with fuel containment and damage tolerance of composite wing structures for transport aircraft were investigated. Material evaluation tests were conducted on two toughened resin composites: Celion/HX1504 and Celion/5245. These consisted of impact, tension, compression, edge delamination, and double cantilever beam tests. Another test series was conducted on graphite/epoxy box beams simulating a wing cover to spar cap joint configuration of a pressurized fuel tank. These tests evaluated the effectiveness of sealing methods with various fastener types and spacings under fatigue loading and with pressurized fuel. Another test series evaluated the ability of the selected coatings, film, and materials to prevent fuel leakage through 32-ply AS4/2220-1 laminates at various impact energy levels. To verify the structural integrity of the technology demonstration article structural details, tests were conducted on blade stiffened panels and sections. Compression tests were performed on undamaged and impacted stiffened AS4/2220-1 panels and smaller element tests to evaluate stiffener pull-off, side load and failsafe properties. Compression tests were also performed on panels subjected to Zone 2 lightning strikes. All of these data were integrated into a demonstration article representing a moderately loaded area of a transport wing. This test combined lightning strike, pressurized fuel, impact, impact repair, fatigue and residual strength.

  13. Formulation and Testing of Paraffin-Based Solid Fuels Containing Energetic Additives for Hybrid Rockets

    Science.gov (United States)

    Larson, Daniel B.; Boyer, Eric; Wachs,Trevor; Kuo, Kenneth K.; Story, George

    2012-01-01

    Many approaches have been considered in an effort to improve the regression rate of solid fuels for hybrid rocket applications. One promising method is to use a fuel with a fast burning rate such as paraffin wax; however, additional performance increases to the fuel regression rate are necessary to make the fuel a viable candidate to replace current launch propulsion systems. The addition of energetic and/or nano-sized particles is one way to increase mass-burning rates of the solid fuels and increase the overall performance of the hybrid rocket motor.1,2 Several paraffin-based fuel grains with various energetic additives (e.g., lithium aluminum hydride (LiAlH4) have been cast in an attempt to improve regression rates. There are two major advantages to introducing LiAlH4 additive into the solid fuel matrix: 1) the increased characteristic velocity, 2) decreased dependency of Isp on oxidizer-to-fuel ratio. The testing and characterization of these solid-fuel grains have shown that continued work is necessary to eliminate unburned/unreacted fuel in downstream sections of the test apparatus.3 Changes to the fuel matrix include higher melting point wax and smaller energetic additive particles. The reduction in particle size through various methods can result in more homogeneous grain structure. The higher melting point wax can serve to reduce the melt-layer thickness, allowing the LiAlH4 particles to react closer to the burning surface, thus increasing the heat feedback rate and fuel regression rate. In addition to the formulation of LiAlH4 and paraffin wax solid-fuel grains, liquid additives of triethylaluminum and diisobutylaluminum hydride will be included in this study. Another promising fuel formulation consideration is to incorporate a small percentage of RDX as an additive to paraffin. A novel casting technique will be used by dissolving RDX in a solvent to crystallize the energetic additive. After dissolving the RDX in a solvent chosen for its compatibility

  14. Spent fuel sabotage aerosol ratio program : FY 2004 test and data summary

    International Nuclear Information System (INIS)

    Brucher, Wenzel; Koch, Wolfgang; Pretzsch, Gunter Guido; Loiseau, Olivier; Mo, Tin; Billone, Michael C.; Autrusson, Bruno A.; Young, F. I.; Coats, Richard Lee; Burtseva, Tatiana; Luna, Robert Earl; Dickey, Roy R.; Sorenson, Ken Bryce; Nolte, Oliver; Thompson, Nancy Slater; Hibbs, Russell S.; Gregson, Michael Warren; Lange, Florentin; Molecke, Martin Alan; Tsai, Han-Chung

    2005-01-01

    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 data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. The program also provides significant technical and political benefits in international cooperation. We are quantifying the Spent Fuel Ratio (SFR), the ratio of the aerosol particles released from HEDD-impacted actual spent fuel to the aerosol particles produced from surrogate materials, measured under closely matched test conditions, in a contained test chamber. In addition, we are measuring the amounts, nuclide content, size distribution of the released aerosol materials, and enhanced sorption of volatile fission product nuclides onto specific aerosol particle size fractions. These data are the input for follow-on modeling studies to quantify respirable hazards, associated radiological risk assessments, vulnerability assessments, and potential cask physical protection design modifications. This document includes an updated description of the test program and test components for all work and plans made, or revised, during FY 2004. It also serves as a program status report as of the end of FY 2004. All available test results, observations, and aerosol analyses plus interpretations--primarily for surrogate material Phase 2 tests, series 2/5A through 2/9B, using cerium oxide sintered ceramic pellets are included. Advanced plans and progress are described for upcoming tests with unirradiated, depleted uranium oxide and actual spent fuel test rodlets. This spent fuel sabotage--aerosol test program is coordinated with the international Working Group for Sabotage Concerns of

  15. Spent fuel sabotage aerosol ratio program : FY 2004 test and data summary.

    Energy Technology Data Exchange (ETDEWEB)

    Brucher, Wenzel (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Koch, Wolfgang (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Pretzsch, Gunter Guido (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Loiseau, Olivier (Institut de Radioprotection et de Surete Nucleaire, France); Mo, Tin (U.S. Nuclear Regulatory Commission, Washington, DC); Billone, Michael C. (Argonne National Laboratory, Argonne, IL); Autrusson, Bruno A. (Institut de Radioprotection et de Surete Nucleaire, France); Young, F. I. (U.S. Nuclear Regulatory Commission, Washington, DC); Coats, Richard Lee; Burtseva, Tatiana (Argonne National Laboratory, Argonne, IL); Luna, Robert Earl; Dickey, Roy R.; Sorenson, Ken Bryce; Nolte, Oliver (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Thompson, Nancy Slater (U.S. Department of Energy, Washington, DC); Hibbs, Russell S. (U.S. Department of Energy, Washington, DC); Gregson, Michael Warren; Lange, Florentin (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Molecke, Martin Alan; Tsai, Han-Chung (Argonne National Laboratory, Argonne, IL)

    2005-07-01

    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 data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. The program also provides significant technical and political benefits in international cooperation. We are quantifying the Spent Fuel Ratio (SFR), the ratio of the aerosol particles released from HEDD-impacted actual spent fuel to the aerosol particles produced from surrogate materials, measured under closely matched test conditions, in a contained test chamber. In addition, we are measuring the amounts, nuclide content, size distribution of the released aerosol materials, and enhanced sorption of volatile fission product nuclides onto specific aerosol particle size fractions. These data are the input for follow-on modeling studies to quantify respirable hazards, associated radiological risk assessments, vulnerability assessments, and potential cask physical protection design modifications. This document includes an updated description of the test program and test components for all work and plans made, or revised, during FY 2004. It also serves as a program status report as of the end of FY 2004. All available test results, observations, and aerosol analyses plus interpretations--primarily for surrogate material Phase 2 tests, series 2/5A through 2/9B, using cerium oxide sintered ceramic pellets are included. Advanced plans and progress are described for upcoming tests with unirradiated, depleted uranium oxide and actual spent fuel test rodlets. This spent fuel sabotage--aerosol test program is coordinated with the international Working Group for Sabotage Concerns of

  16. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  17. Sipping test on a failed MTR fuel element; Teste de sipping em um elemento combustivel tipo placa falhado

    Energy Technology Data Exchange (ETDEWEB)

    Terremoto, Luis Antonio Albiac; Zeituni, Carlos Alberto; Silva, Antonio Teixeira e; Perrotta, Jose Augusto; Silva, Jose Eduardo Rosa da [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear

    2002-07-01

    This work describes sipping tests performed on MTR fuel elements of the IEA-R1 research reactor, in order to determinate which one failed in the core during a routine operation of the reactor. radioactive iodine isotopes {sup 131} I and {sup 133} I, employed as failure indicators, were detected in samples corresponding to the fuel element IEA-156. The specific activity of each sample, as well as the average leaking rate, were measured for {sup 137} Cs. The nuclear fuels U{sub 3} O{sub 8} - Al dispersion and U - Al alloy were compared concerning their measured average leaking rates of {sup 137} Cs. (author)

  18. Electromagnetic Acoustic Test of the Artificial Defects for a Plate-type Nuclear Fuel

    International Nuclear Information System (INIS)

    Jung, Hyun Kyu; Kim, Dong Min; Lee, Yoon Sang; Cheong, Yong Moo

    2011-01-01

    Most research and test reactors use the nuclear fuel plates which are consisted of a fuel meat in aluminum alloy. Last year, KAERI signed a deal with the Jordan Atomic Energy Commission to build the research reactor and have to supply the plate-type nuclear fuels. For the demands of world market, KAERI started the research and development of the plate-type fuel elements and endeavored to achieve a localization of the plate-type fuel fabrication. For the inspection of plate-type fuel elements to be used in Research Reactors, an immersion pulse-echo ultrasonic technique was applied. This inspection was done under immersion condition, so a nuclear fuel was immersed to be prone to corrosion and needed to have time and cost due to an additional process. The sample that will be examined is a non-ferromagnetic material such as aluminum with a good acousto-elastic property, which requires an effective inspection of a bond quality for a nuclear fuel under a manufacturing environment. The purpose of this study is to investigate the feasibility of an Electromagnetic Acoustic Transducer (EMAT) technology for an automated inspection of a nuclear fuel without water

  19. Status Report on the Fabrication of Fuel Cladding Chemical Interaction Test Articles for ATR Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-28

    FeCrAl alloys are a promising new class of alloys for light water reactor (LWR) applications due to their superior oxidation and corrosion resistance in high temperature environments. The current R&D efforts have focused on the alloy composition and processing routes to generate nuclear grade FeCrAl alloys with optimized properties for enhanced accident tolerance while maintaining properties needed for normal operation conditions. Therefore, the composition and processing routes must be optimized to maintain the high temperature steam oxidation (typically achieved by increasing the Cr and Al content) while still exhibiting properties conducive to normal operation in a LWR (such as radiation tolerance where reducing Cr content is favorable). Within this balancing act is the addition of understanding the influence on composition and processing routes on the FeCrAl alloys for fuel-cladding chemical interactions (FCCI). Currently, limited knowledge exists on FCCI for the FeCrAl-UO2 clad-fuel system. To overcome the knowledge gaps on the FCCI for the FeCrAl-UO2 clad-fuel system a series of fueled irradiation tests have been developed for irradiation in the Advanced Test Reactor (ATR) housed at the Idaho National Laboratory (INL). The first series of tests has already been reported. These tests used miniaturized 17x17 PWR fuel geometry rodlets of second-generation FeCrAl alloys fueled with industrial Westinghouse UO2 fuel. These rodlets were encapsulated within a stainless steel housing.To provide high fidelity experiments and more robust testing, a new series of rodlets have been developed deemed the Accident Tolerant Fuel Experiment #1 Oak Ridge National Laboratory FCCI test (ATF-1 ORNL FCCI). The main driving factor, which is discussed in detail, was to provide a radiation environment where prototypical fuel-clad interface temperatures are met while still maintaining constant contact between industrial fuel and the candidate cladding alloys

  20. Four-point Bend Testing of Irradiated Monolithic U-10Mo Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Rabin, B. H. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lloyd, W. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schulthess, J. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wright, J. K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lind, R. P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Scott, L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wachs, K. M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    This paper presents results of recently completed studies aimed at characterizing the mechanical properties of irradiated U-10Mo fuel in support of monolithic base fuel qualification. Mechanical properties were evaluated in four-point bending. Specimens were taken from fuel plates irradiated in the RERTR-12 and AFIP-6 Mk. II irradiation campaigns, and tests were conducted in the Hot Fuel Examination Facility (HFEF) at Idaho National Laboratory (INL). The monolithic fuel plates consist of a U-10Mo fuel meat covered with a Zr diffusion barrier layer fabricated by co-rolling, clad in 6061 Al using a hot isostatic press (HIP) bonding process. Specimens exhibited nominal (fresh) fuel meat thickness ranging from 0.25 mm to 0.64 mm, and fuel plate average burnup ranged from approximately 0.4 x 1021 fissions/cm3 to 6.0 x 1021 fissions/cm3. After sectioning the fuel plates, the 6061 Al cladding was removed by dissolution in concentrated NaOH. Pre- and post-dissolution dimensional inspections were conducted on test specimens to facilitate accurate analysis of bend test results. Four-point bend testing was conducted on the HFEF Remote Load Frame at a crosshead speed of 0.1 mm/min using custom-designed test fixtures and calibrated load cells. All specimens exhibited substantially linear elastic behavior and failed in a brittle manner. The influence of burnup on the observed slope of the stress-strain curve and the calculated fracture strength is discussed.

  1. 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.

  2. Dissolution test for homogeneity of mixed oxide fuel pellets

    International Nuclear Information System (INIS)

    Lerch, R.E.

    1979-08-01

    Experiments were performed to determine the relationship between fuel pellet homogeneity and pellet dissolubility. Although, in general, the amount of pellet residue decreased with increased homogeneity, as measured by the pellet figure of merit, the relationship was not absolute. Thus, all pellets with high figure of merit (excellent homogeneity) do not necessarily dissolve completely and all samples that dissolve completely do not necessarily have excellent homogeneity. It was therefore concluded that pellet dissolubility measurements could not be substituted for figure of merit determinations as a measurement of pellet homogeneity. 8 figures, 3 tables

  3. Heat exchanger for fuel cell power plant reformer

    Science.gov (United States)

    Misage, Robert; Scheffler, Glenn W.; Setzer, Herbert J.; Margiott, Paul R.; Parenti, Jr., Edmund K.

    1988-01-01

    A heat exchanger uses the heat from processed fuel gas from a reformer for a fuel cell to superheat steam, to preheat raw fuel prior to entering the reformer and to heat a water-steam coolant mixture from the fuel cells. The processed fuel gas temperature is thus lowered to a level useful in the fuel cell reaction. The four temperature adjustments are accomplished in a single heat exchanger with only three heat transfer cores. The heat exchanger is preheated by circulating coolant and purge steam from the power section during startup of the latter.

  4. Biomass Fuel Characterization : Testing and Evaluating the Combustion Characteristics of Selected Biomass Fuels : Final Report May 1, 1988-July, 1989.

    Energy Technology Data Exchange (ETDEWEB)

    Bushnell, Dwight J.; Haluzok, Charles; Dadkhah-Nikoo, Abbas

    1990-04-01

    Results show that two very important measures of combustion efficiency (gas temperature and carbon dioxide based efficiency) varied by only 5.2 and 5.4 percent respectively. This indicates that all nine different wood fuel pellet types behave very similarly under the prescribed range of operating parameters. The overall mean efficiency for all tests was 82.1 percent and the overall mean temperature was 1420 1{degree}F. Particulate (fly ash) ad combustible (in fly ash) data should the greatest variability. There was evidence of a relationship between maximum values for both particulate and combustible and the percentages of ash and chlorine in the pellet fuel. The greater the percentage of ash and chlorine (salt), the greater was the fly ash problem, also, combustion efficiency was decreased by combustible losses (unburned hydrocarbons) in the fly ash. Carbon monoxide and Oxides of Nitrogen showed the next greatest variability, but neither had data values greater than 215.0 parts per million (215.0 ppm is a very small quantity, i.e. 1 ppm = .001 grams/liter = 6.2E-5 1bm/ft{sup 3}). Visual evidence indicates that pellets fuels produced from salt laden material are corrosive, produce the largest quantities of ash, and form the only slag or clinker formations of all nine fuels. The corrosion is directly attributable to salt content (or more specifically, chloride ions and compounds formed during combustion). 45 refs., 23 figs., 19 tabs.

  5. Testing and evaluation of doubly impacted simulant-fueled Milliwatt Generator heat sources

    International Nuclear Information System (INIS)

    Teaney, P.E.; Cartmill, W.B.; Wise, R.L.

    1982-01-01

    As part of the Milliwatt Generator (MWG) Program, 12 simulant-fueled heat sources were fabricated double impact tested, and evaluated at Mound. Ten assemblies were tested at approx. 80 m/sec, and two were tested at approx. 105 m/sec. None of the strength members were breached; therefore, no fuel would have been released as a result of double impacts at the velocities and orientations tested at 450 0 C. There was little difference in results for duplicate tests conducted approx. 80 and approx. 105 m/sec. Ten units contained liners that were embrittled prior to testing. This resulted in cracks in some of the liner that would not have occurred in normally fueled heat sources

  6. Testing and evaluation of doubly impacted simulant-fueled Milliwatt Generator heat sources

    Energy Technology Data Exchange (ETDEWEB)

    Teaney, P.E.; Cartmill, W.B.; Wise, R.L.

    1982-04-09

    As part of the Milliwatt Generator (MWG) Program, 12 simulant-fueled heat sources were fabricated double impact tested, and evaluated at Mound. Ten assemblies were tested at approx. 80 m/sec, and two were tested at approx. 105 m/sec. None of the strength members were breached; therefore, no fuel would have been released as a result of double impacts at the velocities and orientations tested at 450/sup 0/C. There was little difference in results for duplicate tests conducted approx. 80 and approx. 105 m/sec. Ten units contained liners that were embrittled prior to testing. This resulted in cracks in some of the liner that would not have occurred in normally fueled heat sources.

  7. Power ramp tests of MOX fuel rods. HBWR irradiation with the instrument rig, IFA-591

    International Nuclear Information System (INIS)

    Ozawa, Takayuki; Abe, Tomoyuki

    2006-03-01

    Plutonium-uranium mixed oxide (MOX) fuel rods of instrumental rig IFA-591 were ramped in HBWR to study the Advanced Thermal Reactor (ATR) MOX fuel behavior during transient operation and to determine a failure threshold of the MOX fuel rods. Eleven segments were base-irradiated in ATR 'FUGEN' up to 18.4 GWd/t. Zirconium liner claddings were adopted for four segments of them. As the results of non-destructive post irradiation examinations (PIEs) after the base-irradiation and before the ramp tests, no remarkable behavior affecting the integrity of fuel assembly and fuel rod was confirmed. All segments to be used for the ramp tests, which consisted of the multi-step ramp tests and the single-step ramp tests, had instrumentations for in-pile measurements of cladding elongation or plenum pressure, and heated up to the maximum linear power of 58.3-68.4 kW/m without failure. The major results of ramp tests are as follows: There is no difference in PCMI behaviors between two type rods of Zry-2 and Zirconium liner claddings from the in-pile measurements of cladding elongation and plenum pressure. The computations of cladding elongation and inner pressure gave slightly lower elongation and pressure than the in-pile measurements during the ramp-test. However, the cladding relaxation during the power hold was in good agreement, and the fission gas release behavior during cooling down could be evaluated by taking into account the relaxation of contact pressure between pellet and cladding. Although the final power during IFA-591 ramp tests reached the higher linear power than the failure threshold power of UO 2 fuel rods, no indication of fuel failure was observed during the ramp tests. The cladding relaxation due to the creep deformation of the MOX pellets at high temperature could be confirmed at the power steps during the multi-ramp test. The fission gas release due to the emancipation from PCMI stress was observed during the power decreasing. The burn-up dependence could be

  8. Spent Fuel Test - Climax: technical measurements. Interim report, fiscal year 1982

    International Nuclear Information System (INIS)

    Patrick, W.C.; Ballou, L.B.; Butkovich, T.R.

    1983-02-01

    The Spent Fuel Test - Climax (SFT-C) is located 420 m below surface in the Climax stock granite on the Nevada Test Site. The test is being conducted for the US Department of Energy (DOE) under the technical direction of the Lawrence Livermore National Laboratory (LLNL). Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized April to May 1980, thus initiating a test with a planned 3- to 5-year fuel storage phase. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. Three exchanges of spent fuel between the SFT-C and a surface storage facility furthered this demonstration. Technical objectives of the test led to development of a technical measurements program, which is the subject of this and two previous interim reports. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the first 2-1/2 years of the test on more than 900 channels. Data continue to be acquired from the test. Some data are now available for analysis and are presented here. Highlights of activities this year include analysis of fracture data obtained during site characterization, laboratory studies of radiation effects and drilling damage in Climax granite, improved calculations of near-field heat transfer and thermomechanical response, a ventilation effects study, and further development of the data acquisition and management systems

  9. Fabrication and testing of full-length single-cell externally fueled converters for thermionic reactors

    International Nuclear Information System (INIS)

    Schock, A.

    1994-01-01

    The preceding paper described designs and analyses of thermionic reactors employing full-core-length single-cell converters with their heated emitters located on the outside of their internally cooled collectors, and it presented results of detailed parametric analyses which illustrate the benefits of this unconventional design. The present paper describes the fabrication and testing of full-length prototypical converters, both unfueled and fueled, and presents parametric results of electrically heated tests. The unfueled converter tests demonstrated the practicality of operating such long converters without shorting across a 0.3-mm interelectrode gap. They produced a measured peak output of 751 watts(e) from a single diode and a peak efficiency of 15.4%. The fueled converter tests measured the parametric performance of prototypic UO 2 -fueled converters designed for subsequent in-pile testing. They employed revolver-shaped tungsten elements with a central emitter hole surrounded by six fuel chambers. The full-length converters were heated by a water-cooled RF-induction coil inside an ion-pumped vacuum chamber. This required development of high-vacuum coaxial RF feedthroughs. In-pile test rules required multiple containment of the UO 2 -fuel, which complicated the fabrication of the test article and required successful development of techniques for welding tungsten and other refractory components. The tests measured a peak power output of 530 watts(e) or 7.1 watts/cm 2 at an efficiency of 11.5%

  10. HEAPA Filter Bank In-Place Leak Test of Advanced Fuel Science Building

    International Nuclear Information System (INIS)

    Ji, C. G.; Bae, S. O.; Kim, C. H.

    2007-12-01

    To maintain the optimum condition of Advanced Fuel Science Building in KAERI, this report is described leak tests for HEPA Filter of HVAC in this facility. The main topics of this report are as follows for: - Procurement Specification - Visual Inspection - Airflow Capacity Test - HEPA Filter Bank In-Place Test

  11. HEAPA Filter Bank In-Place Leak Test of Advanced Fuel Science Building

    Energy Technology Data Exchange (ETDEWEB)

    Ji, C. G.; Bae, S. O.; Kim, C. H

    2007-12-15

    To maintain the optimum condition of Advanced Fuel Science Building in KAERI, this report is described leak tests for HEPA Filter of HVAC in this facility. The main topics of this report are as follows for: - Procurement Specification - Visual Inspection - Airflow Capacity Test - HEPA Filter Bank In-Place Test.

  12. Fueled viking generator S/N 106 acceptance vibration test report

    International Nuclear Information System (INIS)

    Anderson, C.; Brewer, C.O.; Abrahamson, S.G.

    1976-01-01

    The Viking Generator S/N 106 was vibrated to the Teledyne Isotope Flight Acceptance Schedule (Random Only) with no deviation from normal generator functional output. Radiographic analysis and power tests before and after the vibration test indicated no change in the condition of the generator. The work was conducted in the Alpha Fuels Environmental Test Facility at Mound Laboratory

  13. Evaluation of plate type fuel elements by eddy current test method

    International Nuclear Information System (INIS)

    Frade, Rangel Teixeira

    2015-01-01

    Plate type fuel elements are used in MTR research nuclear reactors. The fuel plates are manufactured by assembling a briquette containing the fissile material inserted in a frame, with metal plates in both sides of the set, to act as a cladding. This set is rolled under controlled conditions in order to obtain the fuel plate. In Brazil, this type of fuel is manufactured by IPEN and used in the IEA-R1 reactor. After fabrication of three batches of fuel plates, 24 plates, one of them is taken, in order to verify the thickness of the cladding. For this purpose, the plate is sectioned and the thickness measurements are carried out by using optical microscopy. This procedure implies in damage of the plate, with the consequent cost. Besides, the process of sample preparation for optical microscopy analysis is time consuming, it is necessary an infrastructure for handling radioactive materials and there is a generation of radioactive residues during the process. The objective of this study was verify the applicability of eddy current test method for nondestructive measurement of cladding thickness in plate type nuclear fuels, enabling the inspection of all manufactured fuel plates. For this purpose, reference standards, representative of the cladding of the fuel plates, were manufactured using thermomechanical processing conditions similar to those used for plates manufacturing. Due to no availability of fuel plates for performing the experiments, the presence of the plate’s core was simulated using materials with different electrical conductivities, fixed to the thickness reference standards. Probes of eddy current testing were designed and manufactured. They showed high sensitivity to thickness variations, being able to separate small thickness changes. The sensitivity was higher in tests performed on the reference standards and samples without the presence of the materials simulating the core. For examination of the cladding with influence of materials simulating the

  14. Spent fuel sabotage test program, characterization of aerosol dispersal : interim final report

    International Nuclear Information System (INIS)

    Gregson, Michael Warren; Brockmann, John E.; Loiseau, Olivier; Klennert, Lindsay A.; Nolte, Oliver; Molecke, Martin Alan; Autrusson, Bruno A.; Koch, Wolfgang; Pretzsch, Gunter Guido; Brucher, Wenzel; Steyskal, Michele D.

    2008-01-01

    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 provides source-term data that are relevant to plausible sabotage scenarios in relation to spent fuel transport and storage casks and associated risk assessments. We present details and significant results obtained from this program from 2001 through 2007. Measured aerosol results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; measurements of volatile fission product species enhanced sorption--enrichment factors onto respirable particles; and, status on determination of the spent fuel ratio, SFR, needed for scaling studies. Emphasis is provided on recent Phase 3 tests using depleted uranium oxide pellets plus non-radioactive fission product dopants in surrogate spent fuel test rodlets, plus the latest surrogate cerium oxide results and aerosol laboratory supporting calibration work. The DUO 2 , CeO 2 , plus fission product dopant aerosol particle results are compared with available historical data. We also provide a status review on continuing preparations for the final Phase 4 in this program, tests using individual short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. The source-term data, aerosol results, and program design have been tailored to support and guide follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments. This spent fuel sabotage, aerosol test program was performed primarily at Sandia National Laboratories, with support provided by both the U.S. Department of Energy and the Nuclear Regulatory Commission. This program has significant input from, and is cooperatively supported and

  15. A CAREM fuel assembly prototype construction in order to verify its mechanical design using hydrodynamic testing

    International Nuclear Information System (INIS)

    Aparicio, Gaspar; Di Marco, Agustin; Falcone, Jose M.; Giorgis, Miguel A.; Mathot, Sergio R.; Migliori, Julio; Orlando, Oscar S.; Restelli, Miguel A.; Ruggirello, Gabriel; Sapia, Gustavo C.; Zinzallari, Fausto; Bianchi, Daniel R.; Volpi, Ricardo M.

    2000-01-01

    The scope of this paper is to describe the activities of several Groups from three Atomic Centers (C. A. Bariloche, C. A. Ezeiza and C. A. Constituyentes), involved in the manufacturing of a CAREM fuel assembly prototype. The Design Group (UAIN-CAB) carried out the fuel assembly engineering. Cladding components were constructed by the Special Alloys Pilot Factory (UAMCN-CAE). Engineering Group (UACN-CAC) manufactured the parts to be processed, resorting to qualified suppliers. Elastic spacers were completely designed and constructed by this Group, and fuel rods, control rods, guide tubes and spacers were also welded here. Research Reactors Fuels Group (UACN-CAC) carried out the dimensional control of the elaborated parts, while Postirradiation Testing Group (UACN-CAC) performed the assembling of the fuel element. This paper also refers to the design and development of special equipment and devices, all of them required for the prototype construction. (author)

  16. Test validation of nuclear and fossil fuel control operators

    International Nuclear Information System (INIS)

    Moffie, D.J.

    1976-01-01

    To establish job relatedness, one must go through a procedure of concurrent and predictive validation. For concurrent validity a group of employees is tested and the test scores are related to performance concurrently or during the same time period. For predictive validity, individuals are tested but the results of these tests are not used at the time of employment. The tests are sealed and scored at a later date, and then related to job performance. Job performance data include ratings by supervisors, actual job performance indices, turnover, absenteeism, progress in training, etc. The testing guidelines also stipulate that content and construct validity can be used

  17. Spent Fuel Drying System Test Results (Dry-Run in Preparation for Run 8)

    International Nuclear Information System (INIS)

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

    1999-01-01

    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 (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)(a)on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of a test ''dry-run'' conducted prior to the eighth and last of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister6513U. The system used for the dry-run test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. The experimental results are provided in Section 4.0 and discussed Section 5.0

  18. Global nuclear energy partnership fuels transient testing at the Sandia National Laboratories nuclear facilities : planning and facility infrastructure options

    International Nuclear Information System (INIS)

    Kelly, John E.; Wright, Steven Alan; Tikare, Veena; MacLean, Heather J.; Parma, Edward J.Jr; Peters, Curtis D.; Vernon, Milton E.; Pickard, Paul S.

    2007-01-01

    The Global Nuclear Energy Partnership fuels development program is currently developing metallic, oxide, and nitride fuel forms as candidate fuels for an Advanced Burner Reactor. The Advance Burner Reactor is being designed to fission actinides efficiently, thereby reducing the long-term storage requirements for spent fuel repositories. Small fuel samples are being fabricated and evaluated with different transuranic loadings and with extensive burnup using the Advanced Test Reactor. During the next several years, numerous fuel samples will be fabricated, evaluated, and tested, with the eventual goal of developing a transmuter fuel database that supports the down selection to the most suitable fuel type. To provide a comparative database of safety margins for the range of potential transmuter fuels, this report describes a plan to conduct a set of early transient tests in the Annular Core Research Reactor at Sandia National Laboratories. The Annular Core Research Reactor is uniquely qualified to perform these types of tests because of its wide range of operating capabilities and large dry central cavity which extents through the center of the core. The goal of the fuels testing program is to demonstrate that the design and fabrication processes are of sufficient quality that the fuel will not fail at its design limit--up to a specified burnup, power density, and operating temperature. Transient testing is required to determine the fuel pin failure thresholds and to demonstrate that adequate fuel failure margins exist during the postulated design basis accidents

  19. Global nuclear energy partnership fuels transient testing at the Sandia National Laboratories nuclear facilities : planning and facility infrastructure options.

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, John E.; Wright, Steven Alan; Tikare, Veena; MacLean, Heather J. (Idaho National Laboratory, Idaho Falls, ID); Parma, Edward J., Jr.; Peters, Curtis D.; Vernon, Milton E.; Pickard, Paul S.

    2007-10-01

    The Global Nuclear Energy Partnership fuels development program is currently developing metallic, oxide, and nitride fuel forms as candidate fuels for an Advanced Burner Reactor. The Advance Burner Reactor is being designed to fission actinides efficiently, thereby reducing the long-term storage requirements for spent fuel repositories. Small fuel samples are being fabricated and evaluated with different transuranic loadings and with extensive burnup using the Advanced Test Reactor. During the next several years, numerous fuel samples will be fabricated, evaluated, and tested, with the eventual goal of developing a transmuter fuel database that supports the down selection to the most suitable fuel type. To provide a comparative database of safety margins for the range of potential transmuter fuels, this report describes a plan to conduct a set of early transient tests in the Annular Core Research Reactor at Sandia National Laboratories. The Annular Core Research Reactor is uniquely qualified to perform these types of tests because of its wide range of operating capabilities and large dry central cavity which extents through the center of the core. The goal of the fuels testing program is to demonstrate that the design and fabrication processes are of sufficient quality that the fuel will not fail at its design limit--up to a specified burnup, power density, and operating temperature. Transient testing is required to determine the fuel pin failure thresholds and to demonstrate that adequate fuel failure margins exist during the postulated design basis accidents.

  20. Survey test of canister, geology, and fuel cladding material interactions

    International Nuclear Information System (INIS)

    Krogness, J.C.; Almassy, M.Y.; Cantley, D.A.; Davis, R.B.

    1979-08-01

    A series of Material Interaction Test (MIT) is being conducted. The first test is being conducted in the Dry Surface Storage Demonstration at the EMAD facility on NTS. This paper discusses details of this first test and gives a status report on the MIT series. 17 figures

  1. IRRADIATION DEVICE FOR IRRADIATION TESTING OF COATED PARTICLE FUEL AT HANARO

    Directory of Open Access Journals (Sweden)

    BONG GOO KIM

    2013-12-01

    Full Text Available The Korean Nuclear-Hydrogen Technology Development (NHTD Plan will be performing irradiation testing of coated particle fuel at HANARO to support the development of VHTR in Korea. This testing will be carried out to demonstrate and qualify TRISO-coated particle fuel for use in VHTR. The testing will be irradiated in an inert gas atmosphere 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 has nine fuel compacts and the other five compacts and eight graphite specimens. Each compact contains about 260 TRISO-coated particles. The irradiation device is being loaded and irradiated into the OR5 hole of the in HANARO core from August 2013. The device will be operated for about 150 effective full-power days at a peak temperature of about 1030°C in BOC (Beginning of Cycle during irradiation testing. After a peak burn-up of about 4 atomic percentage and a peak fast neutron fluence of about 1.7×1021 n/cm2, PIE (Post-Irradiation Examination of the irradiated coated particle fuel will be performed at IMEF (Irradiated Material Examination Facility. This paper reviews the design of test rod and irradiation device for coated particle fuel, and discusses the technical results for irradiation testing at HANARO.

  2. Corrosion Testing of Stainless Steel Fuel Cell Hardware

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, M.S.; Zawodzinski, C.; Gottesfeld, S.

    1998-11-01

    Metal hardware is gaining increasing interest in polymer electrolyte fuel cell (PEFC) development as a possible alternative to machined graphite hardware because of its potential for low-cost manufacturing combined with its intrinsic high conductivity, minimal permeability and advantageous mechanical properties. A major barrier to more widespread use of metal hardware has been the susceptibility of various metals to corrosion. Few pure metals can withstand the relatively aggressive environment of a fuel cell and thus the choices for hardware are quite limited. Precious metals such as platinum or gold are prohibitively expensive and so tend to be utilized as coatings on inexpensive substrates such as aluminum or stainless steel. The main challenge with coatings has been to achieve pin-hole free surfaces that will remain so after years of use. Titanium has been used to some extent and though it is very corrosion-resistant, it is also relatively expensive and often still requires some manner of surface coating to prevent the formation of a poorly conducting oxide layer. In contrast, metal alloys may hold promise as potentially low-cost, corrosion-resistant materials for bipolar plates. The dozens of commercially available stainless steel and nickel based alloys have been specifically formulated to offer a particular advantage depending upon their application. In the case of austenitic stainless steels, for example, 316 SS contains molybdenum and a higher chromium content than its more common counterpart, 304 SS, that makes it more noble and increases its corrosion resistance. Likewise, 316L SS contains less carbon than 316 SS to make it easier to weld. A number of promising corrosion-resistant, highly noble alloys such as Hastelloy{trademark} or Duplex{trademark} (a stainless steel developed for seawater service) are available commercially, but are expensive and difficult to obtain in various forms (i.e. wire screen, foil, etc.) or in small amounts for R and D

  3. Re-irradiation tests of spent fuel at JMTR by means of re-instrumentation technique

    International Nuclear Information System (INIS)

    Nakamura, Jinichi; Shimizu, Michio; Endo, Yasuichi; Nabeya, Hideaki; Ichise, Kenichi; Saito, Junichi; Oshima, Kunio; Uetsuka, Hiroshi

    1999-01-01

    JAERI has developed re-irradiation test procedures of spent fuel irradiated at commercial reactor by means of re-instrumentation technique. Full length rods irradiated at commercial LWRs were re-fabricated to short length rods, and rod inner pressure gauges and fuel center thermocouples were re-instrumented to the rods. Re-irradiation tests to study the fuel behavior during power change were carried out by means of BOCA/OSF-1 facility at the JMTR. In the tests to study the fission gas release during power change, the rod inner pressure increase was observed during power change, especially during power reduction. The fission gas release during power reduction is estimated to be the release from fission gas bubbles on the grain boundary caused by the thermal stress in the pellet during power reduction. Re-irradiation test of gadolinia added fuel was performed by means of dual re-instrumentation technique (fuel center thermocouples and rod inner pressure gauge). A stepwise fission gas release during power change, and the following fuel center temperature change due to gap conductance change were observed. (author)

  4. Design, fabrication and testing of a liquid hydrogen fuel tank for a long duration aircraft

    Science.gov (United States)

    Mills, Gary L.; Buchholtz, Brian; Olsen, Al

    2012-06-01

    Liquid hydrogen has distinct advantages as an aircraft fuel. These include a specific heat of combustion 2.8 times greater than gasoline or jet fuel and zero carbon emissions. It can be utilized by fuel cells, turbine engines and internal combustion engines. The high heat of combustion is particularly important in the design of long endurance aircraft with liquid hydrogen enabling cruise endurance of several days. However, the mass advantage of the liquid hydrogen fuel will result in a mass advantage for the fuel system only if the liquid hydrogen tank and insulation mass is a small fraction of the hydrogen mass. The challenge is producing a tank that meets the mass requirement while insulating the cryogenic liquid hydrogen well enough to prevent excessive heat leak and boil off. In this paper, we report on the design, fabrication and testing of a liquid hydrogen fuel tank for a prototype high altitude long endurance (HALE) demonstration aircraft. Design options on tank geometry, tank wall material and insulation systems are discussed. The final design is an aluminum sphere insulated with spray on foam insulation (SOFI). Several steps and organizations were involved in the tank fabrication and test. The tank was cold shocked, helium leak checked and proof pressure tested. The overall thermal performance was verified with a boil off test using liquid hydrogen.

  5. Design, Fabrication, and Testing of an External-Fuel [UO2] Full-Length Thermionic Converter

    Energy Technology Data Exchange (ETDEWEB)

    Schock, Alfred; Raab, B; Giorgio, F.

    1971-09-01

    The development of a double-ended full-core-length external-fuel converter, a prototypical fuel module for a 200- to 300-ekw thermionic reactor, is described. The converter design is based on a revolver-shaped tungsten emitter body, with six peripheral fuel chambers loaded with enriched UO2 pellets. The columbium collector is water-cooled through a sub-atmospheric adjustable-pressure helium gap. The converter employs graded metal-ceramic seals, and its double-ended construction is made possible by bellows to compensate for differential axial expansion. Fission gases are vented from the fuel chambers and collected in an accumulator designed for continuous monitoring of the pressure buildup. Component fabrication, assembly sequence, and joining methods are described; also the test procedures, and the converter load control. All tests are performed in vacuum. During inpile testing, the fuel is triply contained, with thermal insulation between the secondary and tertiary containers. Before insertion inpile, the fully fueled converter is qualification-tested by rf-induction heating using a specially developed high vacuum rf-feedthrough.

  6. Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

    Science.gov (United States)

    Keiser, Dennis; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Madden, James

    2017-12-01

    A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-01

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

  8. The effect of test configuration on the true operating conditions of PEM fuel cells. Paper no. IGEC-1-124

    International Nuclear Information System (INIS)

    Simpson, T.; Li, X.

    2005-01-01

    The operating conditions of a single PEM fuel cell can be significantly affected by the configuration in which the fuel cell test is setup. This study investigates the effect on the gas dewpoint temperature of not insulating the inlet fittings to a PEM fuel cell and the effect of non-optimal stack control thermocouple placement on fuel cell stack operating temperature. Both of these setup configurations can significantly affect fuel cell membrane humidification conditions, especially in a single fuel cell as demonstrated through the sample test conditions presented in this paper. (author)

  9. Evaluation of bonding in Kamini plate type fuel elements using ultrasonic immersion testing and metallography

    International Nuclear Information System (INIS)

    Muralidhar, S.; Pandey, V.D.; Mahule, K.N.; Prasad, G.J.; Ghosh, J.K.; Ganguly, C.

    1989-01-01

    Plate type fuel elements have been fabricated for the neutron source reactor 'KAMINI' in the Radiometallurgy Division. These plate fuel elements have been fabricated using the picture frame technique. An important step in the evaluation of the quality of the plate fuel elements is the non-destructive inspection of bonding between the clad plates and the meat and the picture frame. From the available techniques for lack of bond inspection, a choice was made to use the ultrasonic immersion testing technique. This report records the experience gained in the development of the ultrasonic immersion testing technique for the evaluation of bonding in these fuel plates, with a back-up support of metallography. (author). 7 refs., 9 figs

  10. Automated post irradiation handling of spent nuclear fuel in PNC's FMF-2 facility test hot cell

    International Nuclear Information System (INIS)

    Price, M.H.; Frantz, T.R.

    1994-01-01

    The post irradiation examination test cell of the Fuel Monitoring Facility (FMF-2) presents a challenging operational environment of high radiation and operation temperatures in an inert gas atmosphere. Extensive computer integration of analytical instruments and fuel handling equipment is incorporated into the facility design. Two fully programmable overhead type robot systems will be used in the test hot cell for transfer of spent fuel pin magazines. FMF-2 represents the application of remote handling and robotic technology to a hazardous operational environment. Manned entry into the hot cells for equipment maintenance is impossible after that start of operations. In-cell conditions require that the robotic systems be hardened and remotely maintainable. FMF-2 also demonstrates the integration of diverse remote handling system technologies including robots, electromechanical manipulators and in-cell cranes. Similar design techniques may have applicability in design of future US spent fuel handling facilities

  11. Non-Flow-Through Fuel Cell System Test Results and Demonstration on the SCARAB Rover

    Science.gov (United States)

    Scheidegger, Brianne, T.; Burke, Kenneth A.; Jakupca, Ian J.

    2012-01-01

    This paper describes the results of the demonstration of a non-flow-through PEM fuel cell as part of a power system on the SCARAB rover. A 16-cell non-flow-through fuel cell stack from Infinity Fuel Cell and Hydrogen, Inc. was incorporated into a power system designed to act as a range extender by providing power to the rover s hotel loads. This work represents the first attempt at a ground demonstration of this new technology aboard a mobile test platform. Development and demonstration were supported by the Office of the Chief Technologist s Space Power Systems Project and the Advanced Exploration System Modular Power Systems Project.

  12. Full scale tests on remote handled FFTF fuel assembly waste handling and packaging

    International Nuclear Information System (INIS)

    Allen, C.R.; Cash, R.J.; Dawson, S.A.; Strode, J.N.

    1986-01-01

    Handling and packaging of remote handled, high activity solid waste fuel assembly hardware components from spent FFTF reactor fuel assemblies have been evaluated using full scale components. The demonstration was performed using FFTF fuel assembly components and simulated components which were handled remotely using electromechanical manipulators, shielding walls, master slave manipulators, specially designed grapples, and remote TV viewing. The testing and evaluation included handling, packaging for current and conceptual shipping containers, and the effects of volume reduction on packing efficiency and shielding requirements. Effects of waste segregation into transuranic (TRU) and non-transuranic fractions also are discussed

  13. Packaging and transport case of test fuel assembly irradiated in the Creys-Malville reactor

    International Nuclear Information System (INIS)

    Geffroy, J.; Vivien, J.; Pouard, M.; Dujardin, G.N.; Veron, B.; Michoux, H.

    1986-06-01

    Some irradiated fuel assemblies from the fast neutron Creys Malville reactor will be sent to hot laboratories to follow fuel behavior. These test assemblies will be examined after a limited cooling time and transport is realized at high residual power (about 10kW) and cladding temperature should not rise over 500deg C. The fuel assemblies are not dismantled and transported into sodium. The assembly is placed into a case containing sodium plugged and put into a packaging. Dimensioning, thermal behavior, radiation protection and containment are examined [fr

  14. 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.

  15. Experience using individually supplied heater rods in critical power testing of advanced BWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Majed, M.; Morback, G.; Wiman, P. [ABB Atom AB, Vasteras (Sweden)] [and others

    1995-09-01

    The ABB Atom FRIGG loop located in Vasteras Sweden has during the last six years given a large experience of critical power measurements for BWR fuel designs using indirectly heated rods with individual power supply. The loop was built in the sixties and designed for maximum 100 bar pressure. Testing up to the mid eighties was performed with directly heated rods using a 9 MW, 80 kA power supply. Providing test data to develop critical power correlations for BWR fuel assemblies requires testing with many radial power distributions over the full range of hydraulic conditions. Indirectly heated rods give large advantages for the testing procedure, particularly convenient for variation of individual rod power. A test method being used at Stern Laboratories (formerly Westinghouse Canada) since the early sixties, allows one fuel assembly to simulate all required radial power distributions. This technique requires reliable indirectly heated rods with independently controlled power supplies and uses insulated electric fuel rod simulators with built-in instrumentation. The FRIGG loop was adapted to this system in 1987. A 4MW power supply with 10 individual units was then installed, and has since been used for testing 24 and 25 rod bundles simulating one subbundle of SVEA-96/100 type fuel assemblies. The experience with the system is very good, as being presented, and it is selected also for a planned upgrading of the facility to 15 MW.

  16. Thermal simulations and tests in the development of a helmet transport spent fuel elements Research Reactor

    International Nuclear Information System (INIS)

    Saliba, R.; Quintana, F.; Márquez Turiello, R.; Furnari, J.C.; Pimenta Mourão, R.

    2013-01-01

    A packaging for the transport of irradiated fuel from research reactors was designed by a group of researchers to improve the capability in the management of spent fuel elements from the reactors operated in the region. Two half-scale models for MTR fuel were constructed and tested so far and a third one for both MTR and TRIGA fuels will be constructed and tested next. Four test campaigns have been carried out, covering both normal and hypothetical accident conditions of transportation. The thermal test is part of the requirements for the qualification of transportation packages for nuclear reactors spent fuel elements. In this paper both the numerical modelling and experimental thermal tests performed are presented and discussed. The cask is briefly described as well as the finite element model developed and the main adopted hypotheses for the thermal phenomena. The results of both numerical runs and experimental tests are discussed as a tool to validate the thermal modelling. The impact limiters, attached to the cask for protection, were not modelled. (author) [es

  17. 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...... controller is auto-tuned based on the obtained FOPDT model and the SIMC method in Step-Three. The proposed method is implemented on a real-sized physical system and the experimental results showed a promising potential to apply the proposed method into commercial development....

  18. Thermal calculations for the design, construction, operation, and evaluation of the Spent Fuel Test - Climax, Nevada Test Site

    International Nuclear Information System (INIS)

    Montan, D.N.; Patrick, W.C.

    1981-01-01

    The Spent Fuel Test-Climax (SFT-C) is a test of retrievable deep geologic storage of commercially generated spent nuclear reactor fuel in granitic rock. Eleven spent fuel assemblies, together with six 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 LLNL secured funding for the SFT-C, and completed spent fuel emplacement May 28, 1980. This report documents a series of thermal calculations that were performed in support of the SFT-C. Early calculations employed analytical solutions to address such design and construction issues as drift layout and emplacement hole spacings. Operational aspects of the test required more detailed numerical solutions dealing with ventilation and guard-heater power levels. The final set of calculations presented here provides temperature histories throughout the test facility for evaluation of the response of the SFT-C and for comparison of calculations with acquired data. This final set of calculations employs the as-built test geometry and best-available material properties

  19. Eddy current testing of PWR fuel pencils in the pool of the Osiris reactor

    International Nuclear Information System (INIS)

    Faure, M.; Marchand, L.

    1983-12-01

    A nondestructive testing bench is described. It is devoted to examination of high residual power fuel pencils without stress on the cladding nor interference with cooling. Guiding by fluid bearings decrease the background noise. Scanning speed is limited only by safety criteria and data acquisition configuration. Simultaneous control of various parameters is possible. Associated to an irradiation loop, loaded and unloaded in a reactor swinning pool, this bench can follow fuel pencil degradation after each irradiation cycle [fr

  20. Postirradiation examination and evaluation of Peach Bottom fuel test element FTE-6

    International Nuclear Information System (INIS)

    Wallroth, C.F.; Holzgraf, J.F.; Jensen, D.D.

    1977-09-01

    Fuel test element FTE-6 was irradiated in the Peach Bottom high-temperature gas-cooled reactor (HTGR) for 645 equivalent full power days. Four fuel varieties, contained in H-327 graphite bodies, were tested. A primary result of this test has been to demonstrate acceptable performance even with calculated high stresses in the graphite bodies. Heterogeneous fuel loadings in the element caused local power peaking and azimuthal power variations, deforming the graphite fuel bodies and thereby causing bowing nearly five times as large as the diametral clearance within the sleeve. The axial stresses resulting from interference between the fuel bodies and sleeve were estimated to have reached 45% of the ultimate material strength at the end of the irradiation. Residual stresses from differential contraction within the fuel body resulted in probable in-plane stress levels of 130% of the material strength at the end-of-life shutdown and of up to 150% of the strength at shutdown during the irradiation cycle. The high in-plane stresses are local peaks at the corners of a sharp notch in the element, which may account for the stresses failing to cause damage. The lack of observable damage, however, indicates that the methods and data used for stress analysis give results that are either fairly accurate or conservative

  1. Fuel rod-grid interaction wear: in-reactor tests (LWBR development program)

    International Nuclear Information System (INIS)

    Stackhouse, R.M.

    1979-11-01

    Wear of the Zircaloy cladding of LWBR irradiation test fuel rods, resulting from relative motion between rod and rod support contacts, is reported. Measured wear depths were small, 0.0 to 2.7 mils, but are important in fuel element behavior assessment because of the local loss of cladding thickness, as well as the effect on grid spring forces that laterally restrain the rods. An empirical wear analysis model, based on out-of-pile tests, is presented. The model was used to calculate the wear on the irradiation test fuel rods attributed to a combination of up-and-down motions resulting from power and pressure/temperature cycling of the test reactor, flow-induced vibrations, and assembly handling scratches. The calculated depths are generally deeper than the measured depths

  2. High-silicon 238PuO2 fuel characterization study: Half module impact tests

    International Nuclear Information System (INIS)

    Reimus, M.A.H.

    1997-01-01

    The General-Purpose Heat Source (GPHS) provides power for space missions by transmitting the heat of [sup 238]Pu decay to an array of thermoelectric elements. The modular GPHS design was developed to address both survivability during launch abort and return from orbit. Previous testing conducted in support of the Galileo and Ulysses missions documented the response of GPHSs to a variety of fragment- impact, aging, atmospheric reentry, and Earth-impact conditions. The evaluations documented in this report are part of an ongoing program to determine the effect of fuel impurities on the response of the heat source to conditions baselined during the Galileo/Ulysses test program. In the first two tests in this series, encapsulated GPHS fuel pellets containing high levels of silicon were aged, loaded into GPHS module halves, and impacted against steel plates. The results show no significant differences between the response of these capsules and the behavior of relatively low-silicon fuel pellets tested previously

  3. 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)

  4. Status of the NGNP Fuel Experiment AGR-2 Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and support systems will be briefly discussed, followed by the progress and status of the experiment to date.

  5. Combustion test of refuse derived fuel in fluidized bed

    Energy Technology Data Exchange (ETDEWEB)

    Piao, Guilin; Aono, Shegeru; Yamazaki, Ryohei; Fujima, Yukihisa; Mori, Shigekatsu [Nagoya University, Aichi (Japan); Kondo, Motohiro; Yamaguchi, Masataka [Toyota Motor Corp., Aichi (Japan)

    1999-08-01

    Power generation from Refuse Derived Fuel (RFD) is a promising utilization technology for municipal municipal solid waste. To explain the combustion behaviors of RDF was fed continuously in to a 0.3 x 0.3 m x 2.73 m in height bubbling-type fluidized bed combustor and CO, NO{sub x}, SO{sub x} and HCl concentrations in the flue gas from the combustor were detected by a continuos measurement system. It is found that CO concentration in the flue gas is greatly increased by increasing the RFD feed rate, and CO concentration in the flue gas is greatly decreased by increasing the air ratio because the volatile matters rate is extremely great in combustion of RFD. It is also significantly affected by the secondary air injection and the air distribution ratio from the distributor. HCl concentration in the flue gas is effectively controlled by the calcium compound contained in RDF.HCl concentration is maintained to be less than 60 ppm when the bed temperature is 800 degree C, and the HCl removal ratio by the calcium component is higher than 70% even though under the higher bed temperature than 900 degree C. NO{sub x} concentration was among 50-150 ppm and SO{sub x} concentration is less than 0.5 ppm. (author)

  6. Japanese programme on the development of high duty fuel and related power ramping tests

    International Nuclear Information System (INIS)

    Mishima, Y.

    1983-01-01

    Power ramping tests hitherto planned or carried out in Japan can be classified into two categories: 1) the tests programme by private organizations on fuel behaviour under various conditions of power ramping, in participating international programmes; and 2) a partially government sponsored programme, which was officially inaugurated in 1981 under the title of High-Duty Fuel Development Programme. The latter has been carried out by the Nuclear Power Engineering Test Centre, based on the schedule decided by the MITI Committee (chaired by the author), for a period of 10 years. These programmes will be described with emphasis on the latter (National Programme). (author)

  7. Recovery of Information from the Fast Flux Test Facility for the Advanced Fuel Cycle Initiative

    Energy Technology Data Exchange (ETDEWEB)

    Nielsen, Deborah L.; Makenas, Bruce J.; Wootan, David W.; Butner, R. Scott; Omberg, Ronald P.

    2009-09-30

    The Fast Flux Test Facility is the most recent Liquid Metal Reactor to operate in the United States. Information from the design, construction, and operation of this reactor was at risk as the facilities associated with the reactor are being shut down. The Advanced Fuel Cycle Initiative is a program managed by the Office of Nuclear Energy of the U.S. Department of Energy with a mission to develop new fuel cycle technologies to support both current and advanced reactors. Securing and preserving the knowledge gained from operation and testing in the Fast Flux Test Facility is an important part of the Knowledge Preservation activity in this program.

  8. Spent fuel storage cask testing and operational experience at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Eslinger, L.E.; Schmitt, R.C.

    1989-01-01

    Spent-fuel storage cask research, development, and demonstration activities are being performed for the U.S. Department of Energy's (DOE's) Office of Civilian Radioactive Waste Management (OCRWM) as a part of the storage cask testing program. The cask testing program at federal sites and other locations supports the Nuclear Waste Policy Act (NWPA) and DOE objectives for cooperative demonstrations with the cask vendors and utilities for development of at-reactor dry cask storage capabilities for spent nuclear fuel assemblies. One research and development program for the storage cask performance testing of metal storage cask was initiated through a cooperative agreement between Virginia Power and DOE in 1984. The performance testing was conducted for the DOE and the Electric Power Research Institute by the Pacific Northwest laboratory, operated for DOE by Battelle Memorial Institute, and the Idaho National Engineering Laboratory (INEL), operated for DOE by EG ampersand G Idaho, Inc. In 1988 a cooperative agreement was entered into by DOE with Pacific Sierra Nuclear Associates (PSN) for performance testing of the PSN concrete Ventilated Storage Cask. Another closely related activity involving INEL is a transportable storage cask project identified as the Nuclear Fuel Services Spent-Fuel Shipping/Storage Cask Demonstration Project. The purpose of this project is to demonstrate the feasibility of packing, transporting, and storing commercial spent fuel in dual-purpose transport/storage casks

  9. Health assessment of gasoline and fuel oxygenate vapors: generation and characterization of test materials.

    Science.gov (United States)

    Henley, Michael; Letinski, Daniel J; Carr, John; Caro, Mario L; Daughtrey, Wayne; White, Russell

    2014-11-01

    In compliance with the Clean Air Act regulations for fuel and fuel additive registration, the petroleum industry, additive manufacturers, and oxygenate manufacturers have conducted comparative toxicology testing on evaporative emissions of gasoline alone and gasoline containing fuel oxygenates. To mimic real world exposures, a generation method was developed that produced test material similar in composition to the re-fueling vapor from an automotive fuel tank at near maximum in-use temperatures. Gasoline vapor was generated by a single-step distillation from a 1000-gallon glass-lined kettle wherein approximately 15-23% of the starting material was slowly vaporized, separated, condensed and recovered as test article. This fraction was termed vapor condensate (VC) and was prepared for each of the seven test materials, namely: baseline gasoline alone (BGVC), or gasoline plus an ether (G/MTBE, G/ETBE, G/TAME, or G/DIPE), or gasoline plus an alcohol (G/EtOH or G/TBA). The VC test articles were used for the inhalation toxicology studies described in the accompanying series of papers in this journal. These studies included evaluations of subchronic toxicity, neurotoxicity, immunotoxicity, genotoxicity, reproductive and developmental toxicity. Results of these studies will be used for comparative risk assessments of gasoline and gasoline/oxygenate blends by the US Environmental Protection Agency. Copyright © 2014 Elsevier Inc. All rights reserved.

  10. Performance Degradation Tests of Phosphoric Acid Doped Polybenzimidazole Membrane Based High Temperature Polymer Electrolyte Membrane Fuel Cells

    DEFF Research Database (Denmark)

    Zhou, Fan; Araya, Samuel Simon; Grigoras, Ionela

    2015-01-01

    Degradation tests of two phosphoric acid (PA) doped PBI membrane based HT-PEM fuel cells were reported in this paper to investigate the effects of start/stop and the presence of methanol in the fuel to the performance degradation of the HT-PEM fuel cell. Continuous tests with pure dry H2 and meth......Degradation tests of two phosphoric acid (PA) doped PBI membrane based HT-PEM fuel cells were reported in this paper to investigate the effects of start/stop and the presence of methanol in the fuel to the performance degradation of the HT-PEM fuel cell. Continuous tests with pure dry H2...... to the corrosion of carbon support in the catalyst layer and degradation of the PBI membrane. During the continuous test with methanol containing H2 as the fuel the reaction kinetic resistance and mass transfer resistance of both single cells increased, which may be caused by the adsorption of methanol...

  11. 40 CFR 80.582 - What are the sampling and testing methods for the fuel marker?

    Science.gov (United States)

    2010-07-01

    .... The sampling, sample preparation, and testing methods qualified for use in accordance with the... methods for the fuel marker? 80.582 Section 80.582 Protection of Environment ENVIRONMENTAL PROTECTION... the results of a minimum of 20 repeat tests made over 20 days on samples taken from a homogeneous...

  12. Seismic hazard analysis for the NTS spent reactor fuel test site

    International Nuclear Information System (INIS)

    Campbell, K.W.

    1980-01-01

    An experiment is being directed at the Nevada Test Site to test the feasibility for storage of spent fuel from nuclear reactors in geologic media. As part of this project, an analysis of the earthquake hazard was prepared. This report presents the results of this seismic hazard assessment. Two distinct components of the seismic hazard were addressed: vibratory ground motion and surface displacement

  13. Parametric Sensitivity Tests—European Polymer Electrolyte Membrane 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......, pressure, and stoichiometry at varying current density. Furthermore, procedures for polarization curve recording were also tested both in ascending and descending current directions....

  14. Safety demonstration test on solvent fire in fuel reprocessing plant

    International Nuclear Information System (INIS)

    Nishio, Gunji; Hashimoto, Kazuichiro

    1989-03-01

    This report summarizes a fundamental of results obtained in the Reprocessing Plant Safety Demonstration Test Program which was performed under the contract between the Science and Technology Agency of Japan and the Japan Atomic Energy Research Institute. In this test program, a solvent fire was hypothesized, and such data were obtained as fire behavior, smoke behavior and integrity of exhaust filters in the ventilation system. Through the test results, it was confirmed that under the fire condition in hypothetical accident, the integrity of the cell and the cell ventilation system were maintained, and the safety function of the exhaust filters was maintained against the smoke loading. Analytical results by EVENT code agreed well with the present test data on the thermofluid flow in a cell ventilation system. (author)

  15. OECD/NEA Sandia Fuel Project phase I: Benchmark of the ignition testing

    Energy Technology Data Exchange (ETDEWEB)

    Adorni, Martina, E-mail: martina_adorni@hotmail.it [UNIPI (Italy); Herranz, Luis E. [CIEMAT (Spain); Hollands, Thorsten [GRS (Germany); Ahn, Kwang-II [KAERI (Korea, Republic of); Bals, Christine [GRS (Germany); D' Auria, Francesco [UNIPI (Italy); Horvath, Gabor L. [NUBIKI (Hungary); Jaeckel, Bernd S. [PSI (Switzerland); Kim, Han-Chul; Lee, Jung-Jae [KINS (Korea, Republic of); Ogino, Masao [JNES (Japan); Techy, Zsolt [NUBIKI (Hungary); Velazquez-Lozad, Alexander; Zigh, Abdelghani [USNRC (United States); Rehacek, Radomir [OECD/NEA (France)

    2016-10-15

    Highlights: • A unique PWR spent fuel pool experimental project is analytically investigated. • Predictability of fuel clad ignition in case of a complete loss of coolant in SFPs is assessed. • Computer codes reasonably estimate peak cladding temperature and time of ignition. - Abstract: The OECD/NEA Sandia Fuel Project provided unique thermal-hydraulic experimental data associated with Spent Fuel Pool (SFP) complete drain down. The study conducted at Sandia National Laboratories (SNL) was successfully completed (July 2009 to February 2013). The accident conditions of interest for the SFP were simulated in a full scale prototypic fashion (electrically heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate severe accident code validation and to reduce modeling uncertainties within the codes. Phase I focused on axial heating and burn propagation in a single PWR 17 × 17 assembly (i.e. “hot neighbors” configuration). Phase II addressed axial and radial heating and zirconium fire propagation including effects of fuel rod ballooning in a 1 × 4 assembly configuration (i.e. single, hot center assembly and four, “cooler neighbors”). This paper summarizes the comparative analysis regarding the final destructive ignition test of the phase I of the project. The objective of the benchmark is to evaluate and compare the predictive capabilities of computer codes concerning the ignition testing of PWR fuel assemblies. Nine institutions from eight different countries were involved in the benchmark calculations. The time to ignition and the maximum temperature are adequately captured by the calculations. It is believed that the benchmark constitutes an enlargement of the validation range for the codes to the conditions tested, thus enhancing the code applicability to other fuel assembly designs and configurations. The comparison of

  16. Enhanced Low-Enriched Uranium Fuel Element for the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pope, M. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, M. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Morrell, S. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Jamison, R. K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nef, E. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nigg, D. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    Under the current US Department of Energy (DOE) policy and planning scenario, the Advanced Test Reactor (ATR) and its associated critical facility (ATRC) will be reconfigured to operate on low-enriched uranium (LEU) fuel. This effort has produced a conceptual design for an Enhanced LEU Fuel (ELF) element. This fuel features monolithic U-10Mo fuel foils and aluminum cladding separated by a thin zirconium barrier. As with previous iterations of the ELF design, radial power peaking is managed using different U-10Mo foil thicknesses in different plates of the element. The lead fuel element design, ELF Mk1A, features only three fuel meat thicknesses, a reduction from the previous iterations meant to simplify manufacturing. Evaluation of the ELF Mk1A fuel design against reactor performance requirements is ongoing, as are investigations of the impact of manufacturing uncertainty on safety margins. The element design has been evaluated in what are expected to be the most demanding design basis accident scenarios and has met all initial thermal-hydraulic criteria.

  17. Dissolution of Material and Test reactor Fuel in an H-Canyon Dissolver

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-01-26

    In an amended record of decision for the management of spent nuclear fuel (SNF) at the Savannah River Site, the US Department of Energy has authorized the dissolution and recovery of U from 1000 bundles of Al-clad SNF. The SNF is fuel from domestic and foreign research reactors and is typically referred to as Material Test Reactor (MTR) fuel. Bundles of MTR fuel containing assemblies fabricated from U-Al alloys (or other U compounds) are currently dissolved using a Hg-catalyzed HNO3 flowsheet. Since the development of the existing flowsheet, improved experimental methods have been developed to more accurately characterize the offgas composition and generation rate during laboratory dissolutions. Recently, these new techniques were successfully used to develop a flowsheet for the dissolution of High Flux Isotope Reactor (HFIR) fuel. Using the data from the HFIR dissolution flowsheet development and necessary laboratory experiments, the Savannah River National Laboratory (SRNL) was requested to define flowsheet conditions for the dissolution of MTR fuels. With improved offgas characterization techniques, SRNL will be able define the number of bundles of fuel which can be charged to an H-Canyon dissolver with much less conservatism.

  18. Fluid-structure interaction analysis of the drop impact test for helicopter fuel tank.

    Science.gov (United States)

    Yang, Xianfeng; Zhang, Zhiqiang; Yang, Jialing; Sun, Yuxin

    2016-01-01

    The crashworthiness of helicopter fuel tank is vital to the survivability of the passengers and structures. In order to understand and improve the crashworthiness of the soft fuel tank of helicopter during the crash, this paper investigated the dynamic behavior of the nylon woven fabric composite fuel tank striking on the ground. A fluid-structure interaction finite element model of the fuel tank based on the arbitrary Lagrangian-Eulerian method was constructed to elucidate the dynamic failure behavior. The drop impact tests were conducted to validate the accuracy of the numerical simulation. Good agreement was achieved between the experimental and numerical results of the impact force with the ground. The influences of the impact velocity, the impact angle, the thickness of the fuel tank wall and the volume fraction of water on the dynamic responses of the dropped fuel tank were studied. The results indicated that the corner of the fuel tank is the most vulnerable location during the impact with ground.

  19. Identification and Quantification of Carbon Phases in Conversion Fuel for the Transient Reactor Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Steele, Robert; Mata, Angelica; Dunzik-Gougar, Mary Lou; van Rooyen, Isabella

    2016-06-01

    As part of an overall effort to convert US research reactors to low-enriched uranium (LEU) fuel use, a LEU conversion fuel is being designed for the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory. TREAT fuel compacts are comprised of UO2 fuel particles in a graphitic matrix material. In order to refine heat transfer modeling, as well as determine other physical and nuclear characteristics of the fuel, the amount and type of graphite and non-graphite phases within the fuel matrix must be known. In this study, we performed a series of complementary analyses, designed to allow detailed characterization of the graphite and phenolic resin based fuel matrix. Methods included Scanning Electron and Transmission Electron Microscopies, Raman spectroscopy, X-ray Diffraction, and Dual-Beam Focused Ion Beam Tomography. Our results indicate that no single characterization technique will yield all of the desired information; however, through the use of statistical and empirical data analysis, such as curve fitting, partial least squares regression, volume extrapolation and spectra peak ratios, a degree of certainty for the quantity of each phase can be obtained.

  20. Status of research reactor fuel test in the High Flux Reactor (Petten)

    Energy Technology Data Exchange (ETDEWEB)

    Guidez, J.; Casalta, S. [European Commission, JRC, NL-1755 ZG Petten (Netherlands); Wijtsma, F.J.; Thijssen, P.J.M.; Hendriks, J.A.; Dassei, G. [NRG Petten (Netherlands); Vacelet, H. [Cerca Framatome, F-26104 Romans (France); Languille, A. [CEA Cadarache, F-13108 Saint Paul Lez Durance (France)

    2001-07-01

    Even if the research reactors are using very well known MTR-fuel, a need exists for research in this field mainly for the reasons of industrial qualification of fuel assemblies (built with qualified fuel), improvement or modifications on a qualified fuel ( e.g. increase of density), and qualification of a new fuels such as UMo. For these types of tests, the High Flux Reactor located in Petten (the Netherlands) has a lot of specific advantages: 1) a large core with various interesting positions ranging from high to low fluence rate; 2) a high number of operating days (>280 days/year) that gives - with the high flux available - a possibility to reach quickly high burnup; 3) a downward coolant flow that simplifies the device engineering; 4) all possibilities of non-destructive and destructive examinations in the hot-cells (visual inspection, swelling, {gamma}-scanning, macro- and light microscopy, SEM and EPMA examinations, tomography). Two types of tests can be performed at the plant: either a full-scale test or a test of plates in dedicated devices. A presentation is made of the irradiation test on four UMo plates, begun in March 2000 in the device UMUS. A status report is provided of the full-scale test to be done in the near future, especially the UMo tests to begin the next year. In conclusion it appears that the HFR, that had already given an excellent contribution to silicide fuel qualification in the 1980s, will also give a significant contribution to the current UMo qualification programs. (author)

  1. Research reactor fuel bundle design review by means of hydrodynamic testing

    International Nuclear Information System (INIS)

    Pastorini, A.; Belinco, C.

    1997-01-01

    During the design steps of a fuel bundle for a nuclear reactor, some vibration tests are usually necessary to verify the prototype dynamical response characteristics and the structural integrity. To perform these tests, the known hydrodynamic loop facilities are used to evaluate the vibrational response of the bundle under the different flow conditions that may appear in the reactor. This paper describes the tests performed on a 19 plate fuel bundle prototype designed for a low power research reactor. The tests were done in order to know the dynamical characteristics of the plates and also of the whole bundle under different flow rate conditions. The paper includes a description of the test facilities and the results obtained during the dynamical characterization tests and some preliminary comments about the tests under flowing water are also presented. (author) [es

  2. Spent fuel test - Climax: technical measurements. Interim report, Fiscal Year 1983

    Energy Technology Data Exchange (ETDEWEB)

    Patrick, W.C.; Butkovich, T.R.; Carlson, R.C.; Durham, W.B.; Ganow, H.C.; Hage, G.L.; Majer, E.L.; Montan, D.N.; Nyholm, R.A.; Rector, N.L.

    1984-02-01

    The Spent Fuel Test - Climax (SFT-C) is located 420 m below surface in the Climax stock granite on the Nevada Test Site. The test is being conducted as part of the Nevada Nuclear Waste Storage Investigations. Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized April-May 1980. The spent-fuel canisters were retrieved and the thermal sources were de-energized in March-April 1983 when test data indicated that test objectives were met during the 3-year storage phase. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. In addition to emplacement and retrieval operations, three exchanges of spent-fuel between the SFT-C and a surface storage facility, conducted during the storage phase, furthered this demonstration. Technical objectives of the test led to development of a technical measurements program, which is the subject of this and three previous interim reports. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the 3-1/2 year duration of the test on more than 900 channels. Data acquisition from the test is now limited to instrumentation calibration and evaluation activities. Data now available for analysis are presented here. Highlights of activities this year include a campaign of in situ stress measurements, mineralogical and petrological studies of pretest core samples, microfracture analyses of laboratory irradiated cores, improved calculations of near-field heat transfer and thermomechanical response during the final months of heating as well as during a six-month cool-down period, metallurgical analyses of selected test components, and further development of the data acquisition and data management systems. 27 references, 68 figures, 10 tables.

  3. First steps towards fuel cells testing harmonisation: Procedures and parameters for single cell performance evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Lunghi, P. [Department of Industrial Engineering, University of Perugia, Via Duranti 93, Perugia (Italy); Ubertini, S. [Department of Mechanical Engineering, University of Rome ' ' Tor Vergata' ' , Via di Torvergata, 110, Rome (Italy)

    2004-01-01

    The great interest in Fuel Cell Systems, combined with the innovation of the device itself, has led to a huge developmental effort to make the steps necessary for future FC plant commissioning. The clearest and most effective way to evaluate the performance of a fuel cell is to measure it directly and, since few fuel cell test rigs are available at the moment, standard experimental procedures have not been realized so far. Our research group is currently performing single cell testing at the University of Perugia fuel cell laboratory and particular emphasis has been put on the definition of procedures and the testing of parameterisation. The work team strongly believes that this is the key to effective system testing and reliable performance evaluation. In this work, the test parameterisation developed by the team, and the resulting advanced control procedure used for a single MCFC experimental characterization are presented. Efforts have been dedicated to obtain some relevant non-dimensional parameters to allow an easy understanding of the results and quick comparisons with other tests under different operating conditions, or with results obtained on different cells eventually tested in different laboratories. The authors strongly emphasise this topic to avoid the data that developers and research institutions collect being of no practical use due to a lack of shared rules. (Abstract Copyright [2003], Wiley Periodicals, Inc.)

  4. Testing a CANDU-fueling machine at the Institute for Nuclear Research Pitesti

    International Nuclear Information System (INIS)

    Cojocaru, Virgil

    2006-01-01

    In 2003, as a national and European premiere, the Fueling Machine Head no. 4 (F/M) for the Nuclear Power Plant Cernavoda Unit 2 (NPP) was successfully tested at the Institute for Nuclear Research Pitesti (INR). In 2005, the second Fueling Machine (no. 5) has tested for the Nuclear Power Plant Cernavoda Unit 2. The Institute's main objective is to develop scientific and technological support for the Romanian Nuclear Power Program. Testing the Fueling Machines at INR Pitesti is part of the overall program to assimilate the CANDU technology in Romania. To perform the tests of these machines at INR Pitesti, a special testing rig has built being available for this goal. Both the testing rig and staff had successfully assessed by the AECL representatives during two missions. There was a delivery contract between GEC Canada and Nuclear Power Plant Cernavoda - Unit 2 to provide the Fueling Machines no. 4 and no. 5 in Romania before testing activity. As a first conclusion, the Institute for Nuclear Research Pitesti has the facilities, the staff and the experience to perform possible co-operations with any CANDU Reactor owner

  5. Response of unirradiated and irradiated PWR fuel rods tested under power-cooling-mismatch conditions

    International Nuclear Information System (INIS)

    MacDonald, P.E.; Quapp, W.J.; Martinson, Z.R.; McCardell, R.K.; Mehner, A.S.

    1978-01-01

    This report summarizes the results from the single-rod power-cooling-mismatch (PCM) and irradiation effects (IE) tests conducted to date in the Power Burst Facility (PBF) at the U.S. DOE Idaho National Engineering Laboratory. This work was performed for the U.S. NRC under contact to the Department of Energy. These tests are part of the NRC Fuel Behavior Program, which is designed to provide data for the development and verification of analytical fuel behavior models that are used to predict fuel response to abnormal or postulated accident conditions in commercial LWRs. The mechanical, chemical and thermal response of both previously unirradiated and previously irradiated LWR-type fuel rods tested under power-cooling-mismatch condition is discussed. A brief description of the test designs is presented. The results of the PCM thermal-hydraulic studies are summarized. Primary emphasis is placed on the behavior of the fuel and cladding during and after stable film boiling. (orig.) [de

  6. Application of PLUTO Test Facility for U. S. NRC Licensing of a Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dongseok; Shin, Changhwan; Lee, Kanghee; Kang, Heungseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The fuel assembly of the PLUS-7 loaded in the APR-1400 follows the same schedule. Meanwhile, In July 1998, the U.S. NRC adopted a research plan to address the effects of high burnup from a Loss of Coolant Accident (LOCA). From these programs, several important technical findings for rule revision were obtained. Based on the technical findings, the U. S. NRC has amended the 10 CFR 50.46 which will be proclaimed sooner or later. Through the amendment, a LOCA analysis on the fuel assembly has to show the safety at both a fresh and End of Life (EOL) state. The U. S. NRC has already required EOL effects on seismic/LOCA performance for a fuel assembly since 1998. To obtain U.S NRC licensing of a fuel assembly, based on the amendment of 10CFR50.46, a LOCA analysis of the fuel assembly has to show safety both fresh and EOL states. The proper damping factor of the fuel assembly measured at the hydraulic test loop for a dynamic model in a LOCA and a seismic analysis code are at least required. In this paper, we have examined the damping technologies and compared the test facility of PLUTO with others in terms of performance. PLUTO has a better performance on the operating conditions than any others.

  7. Advanced coal-fueled gas turbine systems: Subscale combustion testing. Topical report, Task 3.1

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-01

    This is the final report on the Subscale Combustor Testing performed at Textron Defense Systems` (TDS) Haverhill Combustion Laboratories for the Advanced Coal-Fueled Gas Turbine System Program of the Westinghouse Electric Corp. This program was initiated by the Department of Energy in 1986 as an R&D effort to establish the technology base for the commercial application of direct coal-fired gas turbines. The combustion system under consideration incorporates a modular staged, rich-lean-quench, Toroidal Vortex Slogging Combustor (TVC) concept. Fuel-rich conditions in the first stage inhibit NO{sub x} formation from fuel-bound nitrogen; molten coal ash and sulfated sorbent are removed, tapped and quenched from the combustion gases by inertial separation in the second stage. Final oxidation of the fuel-rich gases, and dilution to achieve the desired turbine inlet conditions are accomplished in the third stage, which is maintained sufficiently lean so that here, too, NO{sub x} formation is inhibited. The primary objective of this work was to verify the feasibility of a direct coal-fueled combustion system for combustion turbine applications. This has been accomplished by the design, fabrication, testing and operation of a subscale development-type coal-fired combustor. Because this was a complete departure from present-day turbine combustors and fuels, it was considered necessary to make a thorough evaluation of this design, and its operation in subscale, before applying it in commercial combustion turbine power systems.

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

    DEFF Research Database (Denmark)

    Montgomery, Melanie; Carlsen, B; Biede, O

    2002-01-01

    In Denmark, biomass such as straw or woodchip is utilized as a fuel for generating energy. Straw is a "carbon dioxide neutral fuel" and therefore does not contribute to the greenhouse effect. When straw is combusted, potassium chloride and potassium sulfate are present in ash products, which...

  9. Mechanical Fatigue Testing of High-Burnup Fuel for Transportation Applications

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Wang, Hong [ORNL

    2015-05-01

    This report describes testing designed to determine the ability of high burnup (HBU) (>45 GWd/MTU) spent fuel to maintain its integrity under normal conditions of transportation. An innovative system, Cyclic Integrated Reversible-bending Fatigue Tester (CIRFT), has been developed at Oak Ridge National Laboratory (ORNL) to test and evaluate the mechanical behavior of spent nuclear fuel (SNF) under conditions relevant to storage and transportation. The CIRFT system is composed of a U-frame equipped with load cells for imposing the pure bending loads on the SNF rod test specimen and measuring the in-situ curvature of the fuel rod during bending using a set up with three linear variable differential transformers (LVDTs).

  10. A follow-up test of failed fuel element of a nuclear reactor

    International Nuclear Information System (INIS)

    Peerasathien, W.

    1974-01-01

    This thesis is a result of test of a number of nuclear fuel rods which have not been used for a long time due to leakage of radioactivity. Water was circulated through each fuel rod in a test cylinder and radioactivity in water was measured. It was found that the detection of Cesium-137 which has a long half-life, does not indicate the extent of leakage of short-lived radioisotopes, some of which are gaseous. These gases are harmful to the reactor operators and users. A better result was obtained by placing the failed fuel rod in the test cylinder close to the reactor to induce fission. Short half-life gases or other nuclides of the same series were then directly measured

  11. The development and testing of reduced enrichment fuels for Canadian research reactors

    International Nuclear Information System (INIS)

    Wood, J.C.; Foo, M.T.; Berthiaume, L.C.

    1983-01-01

    Fuel rods of uranium silicide dispersed in aluminum and clad in aluminum have been developed and tested in the laboratory and in-reactor. The properties of the dispersion fuel materials proved satisfactory with regard to thermal conductivity, aqueous corrosion resistance, strength and ductility, and thermal stability below 473 K. A vacancy condensation model is proposed to account for the thermally-induced swelling that occurs above 473 K by virtue of the chemical reactions that occur between the dispersed silicide fuel particles and the aluminum matrix. The in-reactor fuel core swelling was less than % after irradiation at high powers 76-131 kW/m) to a high terminal burnup (79.2 at% of U-235 atoms). (author)

  12. 400 W High Temperature PEM Fuel Cell Stack Test

    DEFF Research Database (Denmark)

    Andreasen, Søren Juhl; Kær, Søren Knudsen

    2006-01-01

    -1000 series MEAs by Pemeas, with an active area of 45cm2. The low pressure gas channels enable the use of low power blowers instead of a compressor which increases the overall system efficiency. This initial system was made to test the bipolar plate design, and there is no need for humidification...

  13. Post-Irradiation Examination Test of the Parts of X-Gen Nuclear Fuel Assembly

    International Nuclear Information System (INIS)

    Ahn, S. B.; Ryu, W. S.; Choo, Y. S.

    2008-08-01

    The mechanical properties of the parts of a nuclear fuel assembly are degraded during the operation of the reactor, through the mechanism of irradiation damage. The properties changes of the parts of the fuel assembly should be quantitatively estimated to ensure the safety of the fuel assembly and rod during the operation. The test techniques developed in this report are used to produce the irradiation data of the grid 1x1 cell spring, the grid 1x1 cell, the spring on one face of the 1x1 cell, the inner/outer strip of the grid and the welded part. The specimens were irradiated in the CT test hole of HANARO of a 30 MW thermal output at 300 deg. C during about 100 days From the spring test of mid grid 1x1 cell and grid plate, the irradiation effects can be examined. The irradiation effects on the irradiation growth also were occurred. The buckling load of mid grid 1x1 cell does not change with a neutron irradiation. From the tensile tests, the strengths increased but the elongations decreased due to an irradiation. The tensile test and microstructure examination of the spot and fillet welded parts are performed for the evaluation of an irradiation effects. Through these tests of components, the essential data on the fuel assembly design could be obtained. These results will be used to update the irradiation behavior databases, to improve the performance of fuel assembly, and to predict the service life of the fuel assembly in a reactor

  14. Teledyne Energy Systems, Inc., Proton Exchange Member (PEM) Fuel Cell Engineering Model Powerplant. Test Report: Initial Benchmark Tests in the Original Orientation

    Science.gov (United States)

    Loyselle, Patricia; Prokopius, Kevin

    2011-01-01

    Proton Exchange Membrane (PEM) fuel cell technology is the leading candidate to replace the alkaline fuel cell technology, currently used on the Shuttle, for future space missions. During a 5-yr development program, a PEM fuel cell powerplant was developed. This report details the initial performance evaluation test results of the powerplant.

  15. 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)

  16. Physics-based modeling of live wildland fuel ignition experiments in the Forced Ignition and Flame Spread Test apparatus

    Science.gov (United States)

    C. Anand; B. Shotorban; S. Mahalingam; S. McAllister; D. R. Weise

    2017-01-01

    A computational study was performed to improve our understanding of the ignition of live fuel in the forced ignition and flame spread test apparatus, a setup where the impact of the heating mode is investigated by subjecting the fuel to forced convection and radiation. An improvement was first made in the physics-based model WFDS where the fuel is treated as fixed...

  17. 78 FR 33132 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

    Science.gov (United States)

    2013-06-03

    ... Fuel Elements for Use in Research and Test Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Regulatory Guide (RG) 2.3, ``Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in... assurance program for verifying the quality of plate-type uranium-aluminum fuel elements used in research...

  18. The in-pile proving test for fuel assembly of Qinshan nuclear power plant

    International Nuclear Information System (INIS)

    Chen Dianshan; Zhang Shucheng; Kang Rixin; Wang Huarong; Chen Guanghan

    1989-10-01

    The in-pile proving test for fuel assembly of Qinshan nuclear power plant had been conducted in the experimental loop of HWRR at IAE (Institute of Atomic Energy) in Beijing, China, from January 1985 to December 1986. Average burnup of 27000 MWd/tU and peak burnup of 34000 MWd/tU of fuel rod had already been reached. The basic status of the experiment are described, emphasis is placed on the discussion of proving test parameters and analysis of experiment results

  19. Examination of an instrumented fuel element after dryout tests in Winfrith SGHWR

    International Nuclear Information System (INIS)

    Garlick, A.

    1977-01-01

    Examination of a fuel element that had been subjected to dryout tests in the Winfrith steam generating heavy water reactor proved it to be in good condition. There was clear evidence of cladding softening attributable to dryout overheating on several pins, and the boundaries of the softened zones were determined by hardness measurement. There was, however, no suggestion that softening of the cladding had adversely affected integrity of the fuel element. Annealing tests on cladding from selected regions of the pins led to estimated values for dryout temperatures that were consistent with those measured in-reactor. (author)

  20. Surrogate/spent fuel sabotage aerosol ratio testing:phase 1 summary and results.

    Energy Technology Data Exchange (ETDEWEB)

    Vigil, Manuel Gilbert; Sorenson, Ken Bryce; Lange, F. (Gesellschaft fur Anlagen- und reaktorsicherheit (GRS), Germany); Nolte, O. (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Koch, W. (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Dickey, Roy R.; Yoshimura, Richard Hiroyuki; Molecke, Martin Alan; Autrusson, Bruno (Institut de Radioprotection et de Surete Nucleaire (IRSN), France); Young, F. I. (U.S. Nuclear Regulatory Commission); Pretzsch, Gunter Guido (Gesellschaft fur Anlagen- und reaktorsicherheit (GRS), Germany)

    2005-10-01

    This multinational test program is quantifying the aerosol particulates produced when a high energy density device (HEDD) impacts surrogate material and actual spent fuel test rodlets. The experimental work, performed in four consecutive test phases, has been in progress for several years. The overall program provides needed data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. This program also provides significant political benefits in international cooperation for nuclear security related evaluations. The spent fuel sabotage--aerosol test program is coordinated with the international Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC), and supported by both the U.S. Department of Energy and Nuclear Regulatory Commission. This report summarizes the preliminary, Phase 1 work performed in 2001 and 2002 at Sandia National Laboratories and the Fraunhofer Institute, Germany, and documents the experimental results obtained, observations, and preliminary interpretations. Phase 1 testing included: performance quantifications of the HEDD devices; characterization of the HEDD or conical shaped charge (CSC) jet properties with multiple tests; refinement of the aerosol particle collection apparatus being used; and, CSC jet-aerosol tests using leaded glass plates and glass pellets, serving as representative brittle materials. Phase 1 testing was quite important for the design and performance of the following Phase 2 test program and test apparatus.

  1. GRIST-2 preliminary test plan and requirements for fuel fabrication and preirradiation

    International Nuclear Information System (INIS)

    Tang, I.M.; Harmon, D.P.; Torri, A.

    1978-12-01

    The preliminary version of the GRIST-2 test plan has been developed for the planned initial 5 years (1984 to 1989) of TREAT-Upgrade in-pile tests. These tests will be employed to study the phenomenology and integral behavior of GCFR core disruptive accidents (CDAs) and to support the Final Safety Analysis Report (FSAR) CDA analyses for the demonstration plant licensing. The preliminary test plan is outlined. Test Phases I and II are for the fresh fuel (preconditioned or not) CDA behavior at the beginning-of-life (BOL) reactor state. Phase III is for the reactor state that contains irradiated fuel with a saturated content of helium and fission gas. Phase IV is for larger bundle tests and scaling effects

  2. 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)

  3. Performance evaluation of two CANDU fuel elements tested in the TRIGA reactor

    International Nuclear Information System (INIS)

    Nita, Raluca Florentina; Uta, Octavian; Parvan, Marcel; Mincu, Marin

    2010-01-01

    Nuclear Research Institute at Pitesti has a set of facilities, which allow the testing, manipulation and examination of nuclear fuel and structure materials irradiated in CANDU reactors from Cernavoda NPP. These facilities consist of TRIGA materials testing reactor and Post-Irradiation Examination Laboratory (LEPI). The purpose of this work is to describe the post-irradiation examination, of two experimental CANDU fuel elements (EC1 and EC2). The fuel elements were mounted into a pattern port, one in extension of the other in a measuring test for the central temperature evolution. The results of post-irradiation examination are obtained from: Visual inspection and photography of the outer appearance of sheath; Profilometry (diameter, bending, ovalization) and length measuring; Determination of axial and radial distribution of the fission products activity by gamma scanning; Measurement of pressure, volume and isotopic composition of fission gas; Microstructural characterization by metallographic and ceramographic analyzes; Isotopic composition and burn-up determination. The post-irradiation examination results are used, on one hand, to confirm the security, reliability and performance of the irradiated fuel, and on the other hand, for further development of CANDU fuel. (authors)

  4. Development of an Optimal Controller and Validation Test Stand for Fuel Efficient Engine Operation

    Science.gov (United States)

    Rehn, Jack G., III

    There are numerous motivations for improvements in automotive fuel efficiency. As concerns over the environment grow at a rate unmatched by hybrid and electric automotive technologies, the need for reductions in fuel consumed by current road vehicles has never been more present. Studies have shown that a major cause of poor fuel consumption in automobiles is improper driving behavior, which cannot be mitigated by purely technological means. The emergence of autonomous driving technologies has provided an opportunity to alleviate this inefficiency by removing the necessity of a driver. Before autonomous technology can be relied upon to reduce gasoline consumption on a large scale, robust programming strategies must be designed and tested. The goal of this thesis work was to design and deploy an autonomous control algorithm to navigate a four cylinder, gasoline combustion engine through a series of changing load profiles in a manner that prioritizes fuel efficiency. The experimental setup is analogous to a passenger vehicle driving over hilly terrain at highway speeds. The proposed approach accomplishes this using a model-predictive, real-time optimization algorithm that was calibrated to the engine. Performance of the optimal control algorithm was tested on the engine against contemporary cruise control. Results indicate that the "efficient'' strategy achieved one to two percent reductions in total fuel consumed for all load profiles tested. The consumption data gathered also suggests that further improvements could be realized on a different subject engine and using extended models and a slightly modified optimal control approach.

  5. Fuel rod failure during film boiling (PCM-1 test in the PBF)

    International Nuclear Information System (INIS)

    Domenico, W.F.; Stanley, C.J.; Mehner, A.S.

    1978-01-01

    The Power-Cooling-Mismatch (PCM) Test, PCM-1 was conducted in the Power Burst Facility (PFB) in March of 1978. The PCM Test Series is being conducted at the Idaho National Engineering Laboratory by EG and G Idaho, Inc., under contract to the USNRC and is designed to characterize the behavior of nuclear fuel rods operating under conditions of high power or low coolant flow or both leading to departure from nucleate boiling. The PCM-1 test was performed to provide in-pile data for a ''worst case'' PCM incident. The objective of this experiment was to study the behavior of a single pressurized water reactor (PWR) fuel rod subjected to a high-power and low flow environment which would result in cladding failure at full power. The ''worst case'' conditions established for the experiment consisted of a rod peak power of 78.7 kW/m and a coolant mass flux of 1356 kg/s.m 2 . Fuel temperatures at the stipulated operating conditions were such that a significant volume of molten fuel was present when failure occurred which produced a high probability of molten fuel-coolant interaction (MFCI) with the possibility of a vapor explosion

  6. Test run of 50 MW steam generator for prototype fast breeder reactor of PNC

    International Nuclear Information System (INIS)

    Tanaka, Kenji

    1974-01-01

    A sodium heated steam generator is one of the most important machinery in a fast reactor plant, influencing the aspects of economy and reliability. In the Power Reactor and Nuclear Fuel Development Corporation (PNC), the 50 MW steam generator and associated testing facility have been developed, which lead to the demonstration of the steam generator and the establishment of plant control technology for the fast breeder reactor rackled by PNC. The steam generator is separate helical-coil type with an evaporator, a superheater and a reheater. Taking two to three years in construction, in June 1974, the steam generator with the testing facility was successfully operated at rated power of 50 MW continuously for 72 hours. The steam generator planned for the prototype FBR ''Monju'' is first outlined, together with its devlopment program. The 50 MW steam generator and the testing facility are then described, including construction and operation; finally, the results of test and future test plans are explained. (Mori, K.)

  7. Fabrication and Testing of Full-Length Single-Cell Externally Fueled Converters for Thermionic Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Schock, Alfred

    1994-06-01

    The preceding paper described designs and analyses of thermionic reactors employing full-core-length single-cell converters with their heated emitters located on the outside of their internally cooled collectors, and it presented results of detailed parametric analyses which illustrate the benefits of this unconventional design. The present paper describes the fabrication and testing of full-length prototypical converters, both unfueled and fueled, and presents parametric results of electrically heated tests. The unfueled converter tests demonstrated the practicality of operating such long converters without shorting across a 0.3-mm interelectrode gap. They produced a measured peak output of 751 watts(e) from a single diode and a peak efficiency of 15.4%. The fueled converter tests measured the parametric performance of prototypic UO(subscript 2)-fueled converters designed for subsequent in-pile testing. They employed revolver-shaped tungsten elements with a central emitter hole surrounded by six fuel chambers. The full-length converters were heated by a water-cooled RF-induction coil inside an ion-pumped vacuum chamber. This required development of high-vacuum coaxial RF feedthroughs. In-pile test rules required multiple containment of the UO (subscript 2)-fuel, which complicated the fabrication of the test article and required successful development of techniques for welding tungsten and other refractory components. The test measured a peak power output of 530 watts(e) or 7.1 watts/cm (superscript 2) at an efficiency of 11.5%. There are three copies in the file. Cross-Reference a copy FSC-ESD-217-94-529 in the ESD files with a CID #8574.

  8. Testing device for PWR fuel irradiation Isabelle loop in the Osiris experimental reactor

    International Nuclear Information System (INIS)

    Lucot, Michel; Vidal, Henri.

    1979-01-01

    Description of an irradiation device to test PWR fuel pins in fast and thermal neutron flux. 1 to 4 pins can be irradiated in water under a pressure of 150 bar in a temperature range of 250-300 0 C, a flow of 200g/s and a power of 400W/cm for each pin. Tests include normal, exceptional, transient or accidental conditions. First tests are briefly described [fr

  9. Proof test on thermal and hydraulic design reliability of Japanese PWR fuel assemblies

    International Nuclear Information System (INIS)

    Akiyama, Mamoru; Inoue, Akira; Miyazaki, Keiji; Abeta, Sadaaki; Hori, Keiichi; Mukasa, Tomio; Oishi, Masao; Aoki, Toshimasa; Makihara, Yoshiaki.

    1990-01-01

    A series of departure from nucleate boiling (DNB) tests for pressurized water reactors (PWRs) was performed at the Nuclear Power Engineering Test Center. The objective was to prove the reliability of fuel assembly design by confirming the thermal margin of heat transfer. The present method for evaluating the DNB ratio in a Japanese 17 x 17 PWR core is adequate according to the newly obtained DNB test data

  10. The Test for Flow Characteristics of Tubular Fuel Assembly(II) - Experimental results and CFD analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-12-15

    A test facility had been established for the experiment of velocity distribution and pressure drop in a tubular fuel. A basic test had been conducted to examine the performance of the test loop and to verify the accuracy of measurement by pitot-tube. In this report, test results and CFD analysis for the hydraulic characteristics of a tubular fuel, following the previous tests, are described. Coolant velocities in all channels were measured using pitot-tube and the effect of flow rate change on the velocity distribution was also examined. The pressure drop through the tubular fuel was measured for various flow rates in range of 1 kg/s to 21 kg/s to obtain a correlation of pressure drop with variation of flow rate. In addition, a CFD(Computational Fluid Dynamics) analysis was also done to find out the hydraulic characteristics of tubular fuel such as velocity distribution and pressure drop. As the results of CFD analysis can give us a detail insight on coolant flow in the tubular fuel, the CFD method is a very useful tool to understand the flow structure and phenomena induced by fluid flow. The CFX-10, a commercial CFD code, was used in this study. The two results by the experiment and the CFD analysis were investigated and compared with each other. Overall trend of velocity distribution by CFD analysis was somewhat different from that of experiment, but it would be reasonable considering measurement uncertainties. The CFD prediction for pressure drop of a tubular fuel shows a tolerably good agreement with experiment within 8% difference.

  11. NSRR experiment with un-irradiated uranium-zirconium hydride fuel. Design, fabrication process and inspection data of test fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Sasajima, Hideo; Fuketa, Toyoshi; Ishijima, Kiyomi; Kuroha, Hiroshi; Ikeda, Yoshikazu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Aizawa, Keiichi

    1998-08-01

    An experiment plan is progressing in the Nuclear Safety Research Reactor (NSRR) to perform pulse-irradiation with uranium-zirconium hydride (U-ZrH{sub x}) fuel. This fuel is widely used in the training research and isotope production reactor of GA (TRIGA). The objectives of the experiment are to determine the fuel rod failure threshold and to investigate fuel behavior under simulated reactivity initiated accident (RIA) conditions. This report summarizes design, fabrication process and inspection data of the test fuel rods before pulse-irradiation. The experiment with U-ZrH{sub x} fuel will realize precise safety evaluation, and improve the TRIGA reactor performance. The data to be obtained in this program will also contribute development of next-generation TRIGA reactor and its safety evaluation. (author)

  12. Behavior of pre-irradiated fuel under a simulated RIA condition. Results of NSRR Test JM-5

    International Nuclear Information System (INIS)

    Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide; Tanzawa, Sadamitsu; Ishijima, Kiyomi; Kobayashi, Shinsho; Kamata, Hiroshi; Homma, Kozo; Sakai, Haruyuki.

    1995-11-01

    This report presents results from the power burst experiment with pre-irradiated fuel rod, Test JM-5, conducted in the Nuclear Safety Research Reactor (NSRR). The data concerning test method, pre-irradiation, pre-pulse fuel examination, pulse irradiation, transient records and post-pulse fuel examination are described, and interpretations and discussions of the results are presented. Preceding to the pulse irradiation in the NSRR, test fuel rod was irradiated in the Japan Materials Testing Reactor (JMTR) up to a fuel burnup of 25.7 MWd/kgU with average linear heat rate of 33.4 kW/m. The fuel rod was subjected to the pulse irradiation resulting in a desposited energy of 223 ± 7 cal/g·fuel (0.93 ± 0.03 kJ/g·fuel) and a peak fuel enthalpy of 167 ± 5 cal/g·fuel (0.70 ± 0.02 kJ/g·fuel) under stagnant water cooling condition at atmospheric pressure and ambient temperature. Test fuel rod behavior was assessed from pre- and post-pulse fuel examinations and transient records during the pulse. The Test JM-5 resulted in cladding failure. More than twenty small cracks were found in the post-test cladding, and most of the defects located in pre-existing locally hydrided region. The result indicates an occurrence of fuel failure by PCMI (pellet/cladding mechanical interaction) in combination with decreased integrity of hydrided cladding. (author)

  13. Evaluation of Various Solid Biomass Fuels Using Thermal Analysis and Gas Emission Tests

    Directory of Open Access Journals (Sweden)

    Hiroshi Koseki

    2011-04-01

    Full Text Available Various recently proposed biomass fuels are reviewed from the point of view of their safety. Many biomass materials are proposed for use as fuels, such as refuse derived fuel (RDF, wood chips, coal-wood mixtures, etc. However, these fuels have high energy potentials and can cause fires and explosions. We have experienced many such incidents. It is very difficult to extinguish fires in huge piles of biomass fuel or storage facilities. Here current studies on heat generation for these materials and proposed evaluation methods for these new developing materials in Japan are introduced, which are consistent with measurements using highly sensitive calorimeters such as C80, or TAM, and gas emission tests. The highly sensitive calorimeters can detect small heat generation between room temperature and 80 °C, due to fermentation or other causes. This heat generation sometimes initiates real fires, and also produces combustible gases which can explode if fuel is stored in silos or indoor storage facilities.

  14. Can handling E85 motor fuel cause positive breath alcohol test results?

    Science.gov (United States)

    Ran, Ran; Mullins, Michael E

    2013-09-01

    Hand-held breath alcohol analyzers are widely used by police in traffic stops of drivers suspected of driving while intoxicated (DWI). E85 is a motor fuel consisting of 85% ethanol and 15% gasoline or other hydrocarbons, and is available at nearly 2,600 stations in the USA. We sought to determine whether handling E85 fuel could produce measurable breath alcohol results using a hand-held analyzer and to see if this would be a plausible explanation for a positive breath alcohol test. Five healthy adult subjects dispensed or transferred 8 US gallons of E85 fuel in each of four scenarios. We measured breath alcohol concentration in g/210 L of exhaled breath using the BACTrack S50 at 0, 2, 4, 6, 8, 10, 15 and 20 min after each fuel-handling scenario. Most of the subjects had no detectable breath alcohol after handling E85 motor fuel. Transient elevations (0.02-0.04 g/210 L) in breath alcohol measurement occurred up to 6 min after handling E85 in a minority of subjects. We conclude that it is unlikely that handling E85 motor fuel would result in erroneous prosecution for DWI.

  15. Nondestructive testing of fresh and irradiated fuel of research reactor RA

    International Nuclear Information System (INIS)

    Martinc, R.; Bulovic, V.; Sotic, O.; Ekarv, R.; Petrunin, D.; Delegard, C.; Stevovic, J.; Jacimovic, Lj.; Maksimovic, Z.

    1978-01-01

    The NDA procedures, applied on fresh and irradiated fuel of the research reactor RA in Vinca, are described and the obtained results are presented. To measure the relative quantity of U-235 in the fresh uranium fuel elements the passive gamma ray emission method, as well as the zero power reactor RB criticality measurements procedure, have been employed. The gamma-radiography method is also used for inspection of the fresh fuel of the RA reactor. The semiempirical method, based on spatial power distribution determination, and gamma-spectrometric procedure, developed at the Boris Kidric Institute in Vinca, were used to measure the burn-up of irradiated fuel of reactor RA. The procedures presented are of interest for the accounting for and control of nuclear material, the fuel quality control and the economy and safety analysis of the RA reactor. The principles of the NDA procedures reported, and the gained experience, are also of interest for the power reactor fuel testing, at the nuclear power stations that will be constructed in Yugoslavia in the next future [sr

  16. Interim dry cask storage of irradiated Fast Flux Test Facility fuel

    International Nuclear Information System (INIS)

    Scott, P.L.

    1994-09-01

    The Fast Flux Test Facility (FFTF), located at the US Department of Energy's (DOE'S) Hanford Site, is the largest, most modern, liquid metal-cooled test reactor in the world. This paper will give an overview of the FFTF Spent Fuel Off load project. Major discussion areas will address the status of the fuel off load project, including an overview of the fuel off load system and detailed discussion on the individual components that make up the dry cask storage portion of this system. These components consist of the Interim Storage Cask (ISC) and Core Component Container (CCC). This paper will also discuss the challenges that have been addressed in the evolution of this project

  17. Instrumentation Technologies for Improving an Irradiation Testing of Nuclear Fuels and Materials at the HANARO

    International Nuclear Information System (INIS)

    Kim, Bong Goo; Park, Sung Jae; Choo, Ki Nam

    2011-01-01

    Over 50 years of nuclear fuels and materials irradiation testing has led to many countries developing significant improvements in instrumentation to monitor physical parameters and to control the test conditions in Materials Test Reactors (MTRs) or research reactors. Recent effort to deploy new fuels and materials in existing and advanced reactors has increased the demand for well-instrumented irradiation tests. Specifically, demand has increased for tests with sensors capable of providing real-time measurement of key parameters, such as temperature, geometry changes, thermal conductivity, fission gas release, cracking, coating buildup, thermal and fast flux, etc. This review paper documents the current state of instrumentation technologies in MTRs in the world and summarizes on-going research efforts to deploy new sensors. There is increased interest to irradiate new materials and reactor fuels for advanced PWRs and the Gen-IV reactor systems, such as SFRs (Sodium-cooled Fast Reactors), VHTRs (Very-High-Temperature Reactors), SCWRs (Supercritical-Water-cooled Reactors) and GFRs (Gas-cooled Fast Reactor). This review documents the current state of instrumentation technologies in MTRs in the world, identifies challenges faced by previous testing methods and how these challenges were overcome. A wide range of sensors are available to measure key parameters of interest during fuels and materials irradiations in MTRs. Such sensors must be reliable, small size, highly accurate, and able to withstand harsh conditions. On-going development efforts are focusing on providing MTR users a wider range of parameter measurements with increased accuracy. In addition, development efforts are focusing on reducing the impact of sensor on measurements by reducing sensor size. This report includes not only status of instrumentation using research reactors in the world to irradiate nuclear fuels and materials but also future directions relating to instrumentation technologies for

  18. Analysis and evaluation of the in-pile irradiation test of the fuel element for Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Song Wenhui

    1989-10-01

    In order to verify in-pile performances of the fuel element for Qinshan Nuclear Power Plant, the comprehensive in-pile irradiation test of a 3 x 3 - 2 mock fuel element was performed in the Heavy Water Experimental Reactor. Some technical problems and test parameters which have influence on the test results were analyzed in more detail to fully prove that the test is of actual significance. The element was tested under the conditions corresponding to the normal and abnormal conditions of Qinshan NPP. The maximum burnup of the test fuel rods reached 34 GWd/tU and the maximum heat flux of the rods was 1.39 MW/m 2 . The fuel rods didn't fail in the course of test. Finally the test results were evaluated in verifying the performances of the fuel element and its safe operation in Qinshan NPP

  19. Fabrication and Comparison of Fuels for Advanced Gas Reactor Irradiation Tests

    International Nuclear Information System (INIS)

    Phillips, Jeffrey; Barnes, Charles; Hunn, John

    2010-01-01

    As part of the program to demonstrate TRISO-coated fuel for the Next Generation Nuclear Plant, a series of irradiation tests of Advanced Gas Reactor (AGR) fuel are being performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. In the first test, called 'AGR-1,' graphite compacts containing approximately 300,000 coated particles were irradiated from December 2006 until November 2009. Development of AGR-1 fuel sought to replicate the properties of German TRISO-coated particles. No particle failures were seen in the nearly 3-year irradiation to a burn up of 19%. The AGR-1 particles were coated in a two-inch diameter coater. Following fabrication of AGR-1 fuel, process improvements and changes were made in each of the fabrication processes. Changes in the kernel fabrication process included replacing the carbon black powder feed with a surface-modified carbon slurry and shortening the sintering schedule. AGR-2 TRISO particles were produced in a six-inch diameter coater using a change size about twenty-one times that of the two-inch diameter coater used to coat AGR-1 particles. Changes were also made in the compacting process, including increasing the temperature and pressure of pressing and using a different type of press. Irradiation of AGR-2 fuel began in late spring 2010. Properties of AGR-2 fuel compare favorably with AGR-1 and historic German fuel. Kernels are more homogeneous in shape, chemistry and density. TRISO-particle sphericity, layer thickness standard deviations, and defect fractions are also comparable. In a sample of 317,000 particles from deconsolidated AGR-2 compacts, 3 exposed kernels were found in a leach test. No SiC defects were found in a sample of 250,000 deconsolidated particles, and no IPyC defects in a sample of 64,000 particles. The primary difference in properties between AGR-1 and AGR-2 compacts is that AGR-2 compacts have a higher matrix density, 1.6 g/cm3 compared to about 1.3 g/cm3 for AGR-1 compacts. Based on fuel

  20. Partial Defect Verification of Spent Fuel Assemblies by PDET: Principle and Field Testing in Interim Spent Fuel Storage Facility (CLAB) in Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Y.S.; Kerr, P.; Sitaraman, S.; Swan, R. [Global Security Directorate, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Rossa, R. [SCK-CEN, Mol (Belgium); Liljenfeldt, H. [SKB in Oskarshamn (Sweden)

    2015-07-01

    The need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called 'difficult-to-access' areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into 'difficult-to-access' areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reported the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17x17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly burnup levels. (authors)

  1. Fuel Economy and Exhaust Emissions Characteristics of a Diesel Vehicle : Results of the Prototype Volkswagen 1.5 Liter Turbocharged Rabbit Tests

    Science.gov (United States)

    1981-01-01

    Tests were performed on a prototyple Vokswagen (VW) Turbocharged (TC) Rabbit diesel vehicle on a chassis dynamometer. The vheicle was tested for fuel economy and emissions on the urban Federal test Procedure (FTP), Highway Fuel Economy Test (HFET), C...

  2. Spent fuel sabotage test program, characterization of aerosol dispersal : interim final report.

    Energy Technology Data Exchange (ETDEWEB)

    Gregson, Michael Warren; Brockmann, John E.; Loiseau, Olivier (Institut de Radioprotection et de Surete Nucleaire, France); Klennert, Lindsay A.; Nolte, Oliver (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Molecke, Martin Alan; Autrusson, Bruno A. (Institut de Radioprotection et de Surete Nucleaire, France); Koch, Wolfgang (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Pretzsch, Gunter Guido (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Brucher, Wenzel (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Steyskal, Michele D.

    2008-03-01

    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 provides source-term data that are relevant to plausible sabotage scenarios in relation to spent fuel transport and storage casks and associated risk assessments. We present details and significant results obtained from this program from 2001 through 2007. Measured aerosol results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; measurements of volatile fission product species enhanced sorption--enrichment factors onto respirable particles; and, status on determination of the spent fuel ratio, SFR, needed for scaling studies. Emphasis is provided on recent Phase 3 tests using depleted uranium oxide pellets plus non-radioactive fission product dopants in surrogate spent fuel test rodlets, plus the latest surrogate cerium oxide results and aerosol laboratory supporting calibration work. The DUO{sub 2}, CeO{sub 2}, plus fission product dopant aerosol particle results are compared with available historical data. We also provide a status review on continuing preparations for the final Phase 4 in this program, tests using individual short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. The source-term data, aerosol results, and program design have been tailored to support and guide follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments. This spent fuel sabotage, aerosol test program was performed primarily at Sandia National Laboratories, with support provided by both the U.S. Department of Energy and the Nuclear Regulatory Commission. This program has significant input from, and is cooperatively

  3. Gas Generation from K East Basin Sludges and Irradiated Metallic Uranium Fuel Particles Series III Testing

    International Nuclear Information System (INIS)

    Schmidt, Andrew J.; Delegard, Calvin H.; Bryan, Samuel A.; Elmore, Monte R.; Sell, Rachel L.; Silvers, Kurt L.; Gano, Susan R.; Thornton, Brenda M.

    2003-01-01

    The path forward for managing of Hanford K Basin sludge calls for it to be packaged, shipped, and stored at T Plant until final processing at a future date. An important consideration for the design and cost of retrieval, transportation, and storage systems is the potential for heat and gas generation through oxidation reactions between uranium metal and water. This report, the third in a series (Series III), describes work performed at the Pacific Northwest National Laboratory (PNNL) to assess corrosion and gas generation from irradiated metallic uranium particles (fuel particles) with and without K Basin sludge addition. The testing described in this report consisted of 12 tests. In 10 of the tests, 4.3 to 26.4 g of fuel particles of selected size distribution were placed into 60- or 800-ml reaction vessels with 0 to 100 g settled sludge. In another test, a single 3.72-g fuel fragment (i.e., 7150-mm particle) was placed in a 60 ml reaction vessel with no added sludge. The twelfth test contained only sludge. The fuel particles were prepared by crushing archived coupons (samples) from an irradiated metallic uranium fuel element. After loading the sludge materials (whether fuel particles, mixtures of fuel particles and sludge, or sludge-only) into reaction vessels, the solids were covered with an excess of K Basin water, the vessels closed and connected to a gas measurement manifold, and the vessels back-flushed with inert neon cover gas. The vessels were then heated to a constant temperature. The gas pressures and temperatures were monitored continuously from the times the vessels were purged. Gas samples were collected at various times during the tests, and the samples analyzed by mass spectrometry. Data on the reaction rates of uranium metal fuel particles with water as a function of temperature and particle size were generated. The data were compared with published studies on metallic uranium corrosion kinetics. The effects of an intimate overlying sludge layer

  4. USNRC-OECD Halden Project fuel behavior test program: experiment data report for test assemblies IFA-226 and IFA 239

    International Nuclear Information System (INIS)

    Laats, E.T.; MacDonald, P.E.; Quapp, W.J.

    1975-12-01

    The experimental data which were obtained from the IFA-226 and IFA-239 test assemblies during operation in the Halden Boiling Water Reactor are reported. Included are cladding elongation, fuel centerline temperature, internal gas pressure, and power history data from IFA-226 which were obtained from November 1971 through April 1974, and cladding elongation, diametral profile, and power history data from IFA-239 covering the period from March 1973 through April 1974. The data, presented in the form of composite graphs, have been analyzed only to the extent necessary to assure that they are reasonable and correct. A description of these mixed oxide fuel test assemblies and their instrumentation is presented. Test pin fabrication history, instrument calibration data, assembly power calibration methods, and the neutron detector data reduction technique are included as appendices

  5. Production and Testing of Coconut Oil Biodiesel Fuel and its Blend

    Directory of Open Access Journals (Sweden)

    Oguntola J ALAMU

    2010-12-01

    Full Text Available Many researchers have successfully worked on generating energy from different alternative sources including solar and biological sources such as the conversion of trapped energy from sunlight to electricity and conversion of some renewable agricultural products to fuel. This work considers the use of coconut oil for the production of alternative renewable and environmental friendly biodiesel fuel as an alternative to conventional diesel fuel. Test quantities of coconut oil biodiesel were produced through transesterification reaction using 100g coconut oil, 20.0% ethanol (wt% coconut oil, 0.8% potassium hydroxide catalyst at 65°C reaction temperature and 120 min. reaction time. The experiment was carried out three times and average results evaluated. Low yield of the biodiesel (10.4% was obtained. The coconut oil biodiesel produced was subsequently blended with petroleum diesel and characterized as alternative diesel fuel through some ASTM standard fuel tests. The products were further evaluated by comparing specific gravity and viscosity of the biodiesel blend, the raw coconut oil and conventional petroleum diesel.

  6. Test Specifications and the Design of the Wire Wrapped 37-Pin Fuel Assembly for Hydrodynamic Experiments

    International Nuclear Information System (INIS)

    Chang, S. K.; Euh, D. J.; Bae, H.; Lee, H. Y.; Choi, S. R.

    2013-01-01

    Most influencing parameters on uncertainties and sensitivities of the CFD analyses are the friction coefficient and the mixing coefficient. The friction coefficient is related to the flow distribution in reactor sub-channels. The mixing coefficient is defined with the cross flow between neighboring sub-channels. The eventual purpose of the thermal hydraulic design considering these parameters is to guarantee the fuel cladding integrity as the design limit parameter. At the moment, the experimental program is being undertaken to quantify these friction and mixing parameters which characterize the flow distribution in sub-channels, and the wire wrapped 37-pin rod assembly and its hexagonal test rig have been designed and fabricated. The quantified thermal hydraulic experimental data from this program are utilized primarily to estimate the accuracy of the safety analysis codes and their thermal hydraulic model. A wire wrapped 37 pin fuel assembly has been designed for the measurements of the flow distribution, where the measurements are utilized to quantify the friction coefficient and the mixing coefficient. The test rig of the wire wrapped 37 pin fuel assembly has been fabricated considering the geometric and flow dynamic similarities. It comprises four components i. e., the upper plenum, the fuel housing, the lower plenum, and the wire wrapped 37 pin fuel assembly. At further works, the quantified friction and mixing coefficients through the experiments are going to be utilized for insuring the reliability of the CFD analysis results

  7. Multicylinder Diesel Engine Tests with Unstabilized Water-in-Fuel Emulsions

    Science.gov (United States)

    1981-06-01

    Two diesel engines representative of the four-stroke cycle and two-stroke cycle main propulsion units installed in U.S. Coast Guard WPB class cutters were operated in a test environment in an attempt to demonstrate significant fuel savings associated...

  8. Verification test of advanced LWR fuel components of Westinghouse type nuclear power plants

    International Nuclear Information System (INIS)

    Kim, Hyung Kyu; Yoon, Kyung Ho; Lee, Young Ho

    2004-08-01

    The purpose of this project is to independently conduct the performance test of the spacer grids and the cladding material of the 16x16 and 17x17 advanced fuels for Westinghouse type plants, and to improve the relevant test technology. Major works and results of the present research are as follows. 1. The design and structural features of the spacer grids were investigated, especially the finally determined I-spring was thoroughly analyzed in the point of the mechanical damage and characteristic. 2. As for the mechanical tests of the space grids, the characterization, the impact and the fretting wear tests were carried out. The block as well as the in-grid tests were conducted for the spring/dimple characterization, from which a simple method was developed that simulated the boundary conditions of the assembled grid straps. The impact tester was modified and improved to accommodate a full size grid assembly. The impact result showed that the grid assembly fulfilled the design criteria. As for the fretting wear tests, a sliding test under the room temperature air/water, a sliding/impact test under the room temperature air and a sliding/impact tests under the high temperature and pressure environments were carried out. To this end, a high temperature and pressure fretting wear tester was newly developed. The wear characteristic and the resistibility of the advanced grid spring/dimple were analyzed in detail. The test results were verified through comparing those with the test results by the Westinghouse company. 3. The properties and performance of the newly adopted material for the cladding, Low Sn Zirlo was investigated by a room and high temperature tensile tests and a corrosion tests under the environments of 360 .deg. C water, 400 steam and 360 .deg. C 70ppm LiOH. Through the present project, all the test equipment and technologies for the fuel components were procured, which will be used for future domestic development of a new fuel

  9. Two CANDU fueling machines tested at the Institute For Nuclear Research - Pitesti

    International Nuclear Information System (INIS)

    Doca, Cezar; Cojocaru, Virgil

    2005-01-01

    In 2003, as a national and European premiere, at the Institute for Nuclear Research Pitesti (INR), the Fueling Machine Head no.4 (F/M) for the Nuclear Power Plant Cernavoda - Unit 2 was successfully tested. In 2005, a second Fueling Machine (no.5) was tested for the Nuclear Power Plant Cernavoda - Unit 2. The Institute's main objective is to develop scientific and technological support for the Romanian Nuclear Power Program. Testing the Fueling Machines at INR Pitesti is part of the overall program to assimilate in Romania the CANDU technology. To perform the tests of these machines at INR Pitesti, a special testing rig was built and is available for this goal. Both the testing rig and staff had successfully assessed by the AECL representatives during two missions. There was a delivery contract between GEC Canada and Nuclear Power Plant Cernavoda - Unit 2 to provide the Fueling Machines no. 4 and no. 5 in Romania before testing operation. As a first conclusion, the Institute for Nuclear Research Pitesti has the facilities, the staff and the experience to perform possible co-operations with any other CANDU Reactor owner. This experience will support the next steps concerning F/M commissioning in the NPP Cernavoda - Unit 2 and also give the confidence to the end-users that the Institute's team can provide technical assistance during the operation. Also, the obtained results demonstrate that the overall refurbishment of the F/M control system in Unit 1 and Unit 2 will be possible. The paper presents: - a short description of the F/M head;- a short description of the F/M test rig; - the computer control system; - the F/M testing activities; -results and expectations. (authors)

  10. Two CANDU fueling machines tested at the Institute For Nuclear Research - Pitesti

    International Nuclear Information System (INIS)

    Doca, C.; Cojocaru, V.

    2005-01-01

    Full text: In 2003, as a national and European premiere, at the Institute for Nuclear Research Pitesti (INR), the Fueling Machine Head no.4 (F/M) for the Nuclear Power Plant Cernavoda - Unit 2 was successfully tested. In 2005, a second Fueling Machine (no.5) was tested for the Nuclear Power Plant Cernavoda - Unit 2. The Institute's main objective is to develop scientific and technological support for the Romanian Nuclear Power Program. Testing the Fueling Machines at INR Pitesti is part of the overall program to assimilate in Romania the CANDU technology. To perform the tests of these machines at INR Pitesti, a special testing rig was built and is available for this goal. Both the testing rig and staff had successfully assessed by the AECL representatives during two missions. There was a delivery contract between GEC Canada and Nuclear Power Plant Cernavoda - Unit 2 to provide the Fueling Machines no. 4 and no. 5 in Romania before testing operation. As a first conclusion, the Institute for Nuclear Research Pitesti has the facilities, the staff and the experience to perform possible co-operations with any other CANDU Reactor owner. This experience will support the next steps concerning F/M commissioning in the NPP Cernavoda - Unit 2 and also give the confidence to the end-users that the Institute's team can provide technical assistance during the operation. Also, the obtained results demonstrate that the overall refurbishment of the F/M control system in Unit 1 and Unit 2 will be possible. The paper presents: - a short description of the F/M head;- a short description of the F/M test rig; - the computer control system; - the F/M testing activities; -results and expectations. (authors)

  11. Procedure for vibration test of the fuel rod supported by spacer grids

    International Nuclear Information System (INIS)

    Choi, Myoung Hwan; Kang, Heung Seok; Yoon, Kyung Ho; Kim, Hyung Kyu; Song, Kee Nam

    2002-07-01

    One of the methods that are used to compare and verify the supporting performance of the spacer grids developed is the vibration characteristic test. In this report there are two aims. One is of the understand of the experimental method and procedure performing the modal testing using I-DEAS TDAS module. The other is the investigation of the vibration behaviors of a dummy fuel rod supported by 8 optimized H type spacer grids. This report describes the method and procedure of modal testing to obtain the vibration characteristics such as amplitudes, natural frequencies and mode shapes of the fuel rod using a shaker, a non-contact gap sensor and an accelerometer. This report provides a test procedure in detail so that anyone can be easily understood and use the I-DEAS TDAS program. The I-DEAS TDAS program related to the modal testing has several tasks including the Modal analysis, Signal Processing et al.. This report includes model preparation to prepare the geometrical model, Signal Processing (Sine/Standard measurement) to acquire the signal, Modal analysis to obtain the frequencies and mode shapes, Correlation to analyze the relation between the test and FE analysis and Post Processing tasks. In addition, this report contains the actual test and analysis data of a dummy fuel rod in length 3847mm supported by 8 optimized H type spacer grids

  12. Spent fuel test - Climax: technical measurements. Interim report, fiscal year 1981

    Energy Technology Data Exchange (ETDEWEB)

    Patrick, W.C.; Ballou, L.B.; Butkovich, T.R.

    1982-04-30

    The Spent Fuel Test-Climax (SFT-C) is located 420 m below surface in the Climax granite stock on the Nevada Test Site. Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized from April to May 1980, initiating the 3- to 5-year-duration test. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. Technical objectives of the test led to development of a technical measurements program, which is the subject of this report. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the first 1-1/2 years of the test on more than 900 channels. Much of the acquired data are now available for analysis and are presented here. Highlights of activities this year include completion of site characterization field work, major modifications to the data acquisition and the management systems, and the addition of instrument evaluation as an explicit objective of the test.

  13. Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Maolong; Ryals, Matthew; Ali, Amir; Blandford, Edward; Jensen, Colby; Condie, Keith; Svoboda, John; O' Brien, Robert

    2016-08-01

    A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentally investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.

  14. Behavior of EBR-II Mk-V-type fuel elements in simulated loss-of-flow tests

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Y.Y.; Tsai, H.; Billone, M.C.; Holland, J.W.; Kramer, J.M. (Argonne National Lab., IL (United States))

    1992-11-01

    This report discusses three furnace heating tests which were conducted with irradiated, HT9-clad and U-19wt.%Pu-l0wt.%Zr-alloy fuel, Mk-V-type fuel elements in the Alpha-Gamma Hot Cell Facility at Argonne National Laboratory, Illinois. In general, very significant safety margins for fuel-element cladding breaching have been demonstrated in these tests, under conditions that would envelop a bounding unlikely loss-of-flow event in EBR-II. Highlights of the test results will be given, as well as discussions of the cladding breaching mechanisms, axial fuel motion, and fuel surface liquefaction found in high-temperature testing of irradiated metallic fuel elements.

  15. Improved ASTM G72 Test Method for Ensuring Adequate Fuel-to-Oxidizer Ratios

    Science.gov (United States)

    Juarez, Alfredo; Harper, Susana Tapia

    2016-01-01

    The ASTM G72/G72M-15 Standard Test Method for Autogenous Ignition Temperature of Liquids and Solids in a High-Pressure Oxygen-Enriched Environment is currently used to evaluate materials for the ignition susceptibility driven by exposure to external heat in an enriched oxygen environment. Testing performed on highly volatile liquids such as cleaning solvents has proven problematic due to inconsistent test results (non-ignitions). Non-ignition results can be misinterpreted as favorable oxygen compatibility, although they are more likely associated with inadequate fuel-to-oxidizer ratios. Forced evaporation during purging and inadequate sample size were identified as two potential causes for inadequate available sample material during testing. In an effort to maintain adequate fuel-to-oxidizer ratios within the reaction vessel during test, several parameters were considered, including sample size, pretest sample chilling, pretest purging, and test pressure. Tests on a variety of solvents exhibiting a range of volatilities are presented in this paper. A proposed improvement to the standard test protocol as a result of this evaluation is also presented. Execution of the final proposed improved test protocol outlines an incremental step method of determining optimal conditions using increased sample sizes while considering test system safety limits. The proposed improved test method increases confidence in results obtained by utilizing the ASTM G72 autogenous ignition temperature test method and can aid in the oxygen compatibility assessment of highly volatile liquids and other conditions that may lead to false non-ignition results.

  16. Chemical Reactivity Testing for the National Spent Nuclear Fuel Program. Quality Assurance Project Plan

    International Nuclear Information System (INIS)

    Newsom, H.C.

    1999-01-01

    This quality assurance project plan (QAPjP) summarizes requirements used by Lockheed Martin Energy Systems, Incorporated (LMES) Development Division at Y-12 for conducting chemical reactivity testing of Department of Energy (DOE) owned spent nuclear fuel, sponsored by the National Spent Nuclear Fuel Program (NSNFP). The requirements are based on the NSNFP Statement of Work PRO-007 (Statement of Work for Laboratory Determination of Uranium Hydride Oxidation Reaction Kinetics.) This QAPjP will utilize the quality assurance program at Y-12, QA-101PD, revision 1, and existing implementing procedures for the most part in meeting the NSNFP Statement of Work PRO-007 requirements, exceptions will be noted

  17. Seismic testing of the base-isolated PWR spent-fuel storage rack

    International Nuclear Information System (INIS)

    Fujita, Katsuhisa; Tanaka, Mamoru; Nakamura, Masaaki; Tsujikura, Yonezo.

    1990-01-01

    The present paper aims to verify the seismic safety of the base-isolated spent-fuel storage rack. A series of seismic tests has been conducted using a three-dimensional shaking table. A sliding-type base-isolation system was employed for the prototype rack considering environmental conditions in an actual plant. A non linear seismic response analysis was also performed, and it is verified that the prototype of a base-isolated spent-fuel storage rack has a sufficient seismic safety margin for design seismic conditions from the viewpoint of seismic response. (author)

  18. Hardware in the loop simulation test platform of fuel cell backup system

    Directory of Open Access Journals (Sweden)

    Ma Tiancai

    2015-01-01

    Full Text Available Based on an analysis of voltage mechanistic model, a real-time simulation model of the proton exchange membrane (PEM fuel cell backup system is developed, and verified by the measurable experiment data. The method of online parameters identification for the model is also improved. Based on the software LabVIEW/VeriStand real-time environment and the PXI Express hardware system, the PEM fuel cell system controller hardware in the loop (HIL simulation plat-form is established. Controller simulation test results showed the accuracy of HIL simulation platform.

  19. Relevance of IAEA tests to severe accidents in nuclear fuel cycle transport

    Energy Technology Data Exchange (ETDEWEB)

    Wilkinson, W.L. [World Nuclear Transport Inst., London (United Kingdom)

    2004-07-01

    The design and performance standards for packages used for the transport of nuclear fuel cycle materials, are defined in the IAEA Regulations for the Safe Transport of Radioactive Materials, TS-R-1, in order to ensure safety under both normal and accident conditions of transport. The underlying philosophy is that safety is vested principally in the package and the design and performance criteria are related to the potential hazard. Type B packages are high duty packages which are used for the transport of the more radioactive materials, notably spent fuel and vitrified high-level waste (VHLW). Tests are specified in the IAEA Regulations to ensure the integrity of these packages in potential transport accidents involving impacts, fires or immersion in water. The mechanical tests for Type B packages include drop tests onto an unyielding surface without giving rise to a significant release of radioactivity. The objects which a package could impact in real life transport accidents, such as concrete roads, bridge abutments and piers, will yield to some extent and absorb some of the energy of the moving package. Impact tests onto an unyielding surface are therefore relevant to impacts onto real-life objects at much higher speeds. The thermal test specifies that Type B packages should be able to withstand a fully engulfing fire of 8000 C for 30 minutes. Analytical studies backed up by experimental tests have shown that these packages can withstand such conditions without significant release of radioactivity. The Regulations also specify immersion tests for Type B packages; 15 metres for 8 hours without significant release of radioactivity and, in addition for spent fuel and VHLW packages, 200 metres for 1 hour without rupture of the containment. Studies have shown that spent fuel and VHLW casks would meet these conditions. Therefore, there is a large body of evidence to show that the current IAEA Type B test requirements are severe and cover all the situations which can

  20. Relevance of IAEA tests to severe accidents in nuclear fuel cycle transport

    International Nuclear Information System (INIS)

    Wilkinson, W.L.

    2004-01-01

    The design and performance standards for packages used for the transport of nuclear fuel cycle materials, are defined in the IAEA Regulations for the Safe Transport of Radioactive Materials, TS-R-1, in order to ensure safety under both normal and accident conditions of transport. The underlying philosophy is that safety is vested principally in the package and the design and performance criteria are related to the potential hazard. Type B packages are high duty packages which are used for the transport of the more radioactive materials, notably spent fuel and vitrified high-level waste (VHLW). Tests are specified in the IAEA Regulations to ensure the integrity of these packages in potential transport accidents involving impacts, fires or immersion in water. The mechanical tests for Type B packages include drop tests onto an unyielding surface without giving rise to a significant release of radioactivity. The objects which a package could impact in real life transport accidents, such as concrete roads, bridge abutments and piers, will yield to some extent and absorb some of the energy of the moving package. Impact tests onto an unyielding surface are therefore relevant to impacts onto real-life objects at much higher speeds. The thermal test specifies that Type B packages should be able to withstand a fully engulfing fire of 8000 C for 30 minutes. Analytical studies backed up by experimental tests have shown that these packages can withstand such conditions without significant release of radioactivity. The Regulations also specify immersion tests for Type B packages; 15 metres for 8 hours without significant release of radioactivity and, in addition for spent fuel and VHLW packages, 200 metres for 1 hour without rupture of the containment. Studies have shown that spent fuel and VHLW casks would meet these conditions. Therefore, there is a large body of evidence to show that the current IAEA Type B test requirements are severe and cover all the situations which can

  1. CRBRP design and test results for fuel handling systems, plugs, and seals

    International Nuclear Information System (INIS)

    Berg, G.E.

    1977-01-01

    The fuel handling system and reactor rotating plugs for the Clinch River Breeder Reactor Plant (CRBRP) are based primarily on existing technology and, in many respects, follow the concept developed for the Fast Flux Test Facility (FFTF). The equipment and the development programs initiated to verify its performance are described. Test results obtained from the development program, and the extent to which these results verified original design selections, or suggested potential improvements, are discussed

  2. Handling and disposal of SP-100 ground test nuclear fuel and equipment

    International Nuclear Information System (INIS)

    Wilson, C.E.; Potter, J.D.; Hodgson, R.D.

    1990-05-01

    The post SP-100 reactor testing period will focus on defueling the reactor, packaging the various radioactive waste forms, and shipping this material to the appropriate locations. Remote-handling techniques will be developed to defuel the reactor. Packaging the spent fuel and activated reactor components is a challenge in itself. This paper presents an overview of the strategy, methods, and equipment that will be used during the closeout phase of nuclear testing

  3. Handling and disposal of SP-100 ground test nuclear fuel and equipment

    International Nuclear Information System (INIS)

    Wilson, C.E.; Potter, J.D.; Hodgson, R.D.

    1991-01-01

    The post SP-100 reactor testing period will focus on defueling the reactor, packaging the various radiactive waste forms, and shipping this material to the appropriate locations. Remote-handling techniques will be developed to defuel the reactor. Packaging the spent fuel and activated reactor components is a challenge in itself. This paper presents an overview of the strategy, methods, and equipment that will be used during the closeout phase of nuclear testing

  4. Post irradiation tests on CANDU fuel irradiated in power ramp conditions

    International Nuclear Information System (INIS)

    Ionescu, Silviu; Uta, Octavian; Parvan, Marcel; Gentea, Cristian; Ichim, Ovidiu

    2008-01-01

    Nuclear Research Branch Pitesti disposes of facilities, which allow the testing, manipulation and examination of nuclear fuel and of irradiated structure materials in CANDU reactor from Cernavoda NPP. These facilities imply the materials testing reactor TRIGA and the Post-Irradiation Examination Laboratory (PIEL). The purpose of this work is to determine by post-irradiate examination, the behavior of CANDU indigenous fuel, irradiated in 14 MW(th) TRIGA reactor into a multiple/various power ramp tests. The results of post-irradiate examination consist of: - Visual inspection and photography of the outer appearance of sheath; - Profilometry (diameter, bending, ovalization) and length measuring; - Determination of axial and radial distribution of the fission products activity by gamma scanning and tomography; - measurement of pressure, volume and isotopic composition of fission gas; - Microstructural characterization by metallographic and ceramographic analyzes; - Isotopic composition and burn-up determination; - Mechanical properties determination. The obtained data from the post-irradiate examinations are used, on one hand, in order to confirm the security, reliability and nuclear fuel performance, and on the other hand, for further optimization of the CANDU fuel. (authors)

  5. 10 CFR 830 Major Modification Determination for Advanced Test Reactor LEU Fuel Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Boyd D. Christensen; Michael A. Lehto; Noel R. Duckwitz

    2012-05-01

    The Advanced Test Reactor (ATR), located in the ATR Complex of the Idaho National Laboratory (INL), was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. The ATR is fueled with high-enriched uranium (HEU) matrix (UAlx) in an aluminum sandwich plate cladding. The National Nuclear Security Administration Global Threat Reduction Initiative (GTRI) strategic mission includes efforts to reduce and protect vulnerable nuclear and radiological material at civilian sites around the world. Converting research reactors from using HEU to low-enriched uranium (LEU) was originally started in 1978 as the Reduced Enrichment for Research and Test Reactors (RERTR) Program under the U.S. Department of Energy (DOE) Office of Science. Within this strategic mission, GTRI has three goals that provide a comprehensive approach to achieving this mission: The first goal, the driver for the modification that is the subject of this determination, is to convert research reactors from using HEU to LEU. Thus the mission of the ATR LEU Fuel Conversion Project is to convert the ATR and Advanced Test Reactor Critical facility (ATRC) (two of the six U.S. High-Performance Research Reactors [HPRR]) to LEU fuel by 2017. The major modification criteria evaluation of the project pre-conceptual design identified several issues that lead to the conclusion that the project is a major modification.

  6. Onsite 40-kilowatt fuel cell power plant manufacturing and field test program

    Science.gov (United States)

    1985-01-01

    A joint Gas Research Institute and U.S. Department of Energy Program was initiated in 1982 to evaluate the use of fuel cell power systems for on-site energy service. Forty-six 40 kW fuel cell power plants were manufactured at the United Technologies Corporation facility in South Windsor, Connecticut, and are being delivered to host utilities and other program participants in the United States and Japan for field testing. The construction of the 46 fully-integrated power plants was completed in January 1985 within the constraints of the contract plan. The program has provided significant experience in the manufacture, acceptance testing, deployment, and support of on-site fuel cell systems. Initial field test results also show that these experimental power plants meet the performance and environmental requirements of a commercial specification. This Interim Report encompasses the design and manufacturing phases of the 40 kW Power Plant Manufacturing and Field Test program. The contract between UTC and NASA also provides UTC field engineering support to the host utilities, training programs and associated manuals for utility operating and maintenance personnel, spare parts support for a defined test period, and testing at UTC of a power plant made available from a preceding program phase. These activities are ongoing and will be reported subsequently.

  7. Simulation of Irradiated BWR fuel rod (TS) test in NSRR using FRAP-T6 and NSR-77

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Murofushi, Akira; Hosoyamada, Ryuji.

    1994-03-01

    Series of pulse irradiation tests have been performed in the Nuclear Safety Research Reactor (NSRR) to investigate irradiated fuel rod performance under the Reactivity Initiated Accident (RIA) conditions. Five tests, called Tests TS-1 through TS-5, were conducted in a period from 1989 to 1993 with irradiated 7x7 type BWR fuel rods provided from a commercial power plant. Simulation calculations of the TS tests were carried out with the FRAP-T6 code, which is widely used in the world to estimate fuel performance under various accident conditions, and with the NSR77 code, which describes fresh fuel rod performance well in the NSRR tests. Results of the calculation are compiled in this report and applicability of the codes to the irradiated BWR fuel rod tests is discussed. (author)

  8. Failure problems in superheater spacers of steam generators; Problematica de fallas en espaciadores de sobrecalentadores de generadores de vapor

    Energy Technology Data Exchange (ETDEWEB)

    Chacon Nava, Jose G.; Martinez Villafane, Alberto [Instituto de Investigaciones Electricas, Cuernavaca (Mexico); Fuentes Samaniego, Raul [Universidad Autonoma de Nuevo Leon (Mexico); Mojica Calderon, Cecilio [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

    1988-12-31

    In this article the general aspects of the steam generator superheater fixed spacers failures are analyzed, emphasis is made on the influence several aspects such as the operation of the unit have, the appropriate execution of welds and the selection of binding materials. Likewise several recommendations are made to bring the failures to a minimum. [Espanol] En este articulo se analizan aspectos generales de fallas en espaciadores fijos de sobrecalentadores de generadores de vapor, y se hace hincapie en la influencia que tienen diversos aspectos tales como la operacion de la unidad, la adecuada ejecucion de soldaduras y la seleccion del material de aporte. Asimismo, se proponen algunas recomendaciones para reducir al minimo las fallas.

  9. Evaluation of the behavior of shrouded plasma spray coatings in the platen superheater of coal-fired boilers

    Energy Technology Data Exchange (ETDEWEB)

    Sidhu, B.S.; Prakash, S. [GZS College of Engineering & Technology, Bathinda (India). Dept. of Mechanical Engineering

    2006-06-15

    Nickel- and cobalt-based coatings were formulated by a shrouded plasma spray process on boiler tube steels, namely, ASTM-SA210-grade A1 (GrA1), ASTM-SA213-T-11 (T11), and ASTM-SA213-T-22 (T22). The Ni-22Cr-10A1-1Y alloy powder was sprayed as a bond in each case before the final coating. The degradation behavior of the bared and coated steels was studied in the platen superheater of the coal-fired boiler. The samples were inserted through the soot blower dummy points with the help of stainless steel wires. The coatings were found to be effective in increasing resistance to degradation in the given boiler environment. The maximum protection was observed in the case of Stellite-6 (St-6) coating.

  10. Pre-oxidation and its effect on reducing high-temperature corrosion of superheater tubes during biomass firing

    DEFF Research Database (Denmark)

    Okoro, Sunday Chukwudi; Kvisgaard, M.; Montgomery, Melanie

    2017-01-01

    with a synthetic deposit of KCl and exposed at 560°C for 1 week to a gas mixture typical of biomass firing. Results show that pre-oxidation could hinder the corrosion attack; however, the relative success was different for the two alloys. While corrosion attack was observed on the pre-oxidised Kanthal APM, the pre......-oxidised Nimonic 80A remained unaffected suggesting protection of the alloy from the corrosive environment.......Superheater tubes in biomass-fired power plants experience high corrosion rates due to condensation of corrosive alkali chloride-rich deposits. To explore the possibility of reducing the corrosion attack by the formation of an initial protective oxide layer, the corrosion resistance of pre...

  11. Durability test with fuel starvation using a Pt/CNF catalyst in PEMFC.

    Science.gov (United States)

    Jung, Juhae; Park, Byungil; Kim, Junbom

    2012-01-05

    In this study, a catalyst was synthesized on carbon nanofibers [CNFs] with a herringbone-type morphology. The Pt/CNF catalyst exhibited low hydrophilicity, low surface area, high dispersion, and high graphitic behavior on physical analysis. Electrodes (5 cm2) were prepared by a spray method, and the durability of the Pt/CNF was evaluated by fuel starvation. The performance was compared with a commercial catalyst before and after accelerated tests. The fuel starvation caused carbon corrosion with a reverse voltage drop. The polarization curve, EIS, and cyclic voltammetry were analyzed in order to characterize the electrochemical properties of the Pt/CNF. The performance of a membrane electrode assembly fabricated from the Pt/CNF was maintained, and the electrochemical surface area and cell resistance showed the same trend. Therefore, CNFs are expected to be a good support in polymer electrolyte membrane fuel cells.

  12. Lean mixture engine testing and evaluation program. [for automobile engine pollution and fuel performances

    Science.gov (United States)

    Dowdy, M. W.; Hoehn, F. W.; Griffin, D. C.

    1975-01-01

    Experimental results for fuel consumption and emissions are presented for a 350 CID (5.7 liter) Chevrolet V-8 engine modified for lean operation with gasoline. The lean burn engine achieved peak thermal efficiency at an equivalence ratio of 0.75 and a spark advance of 60 deg BTDC. At this condition the lean burn engine demonstrated a 10% reduction in brake specific fuel consumption compared with the stock engine; however, NOx and hydrocarbon emissions were higher. With the use of spark retard and/or slightly lower equivalence ratios, the NOx emissions performance of the stock engine was matched while showing a 6% reduction in brake specific fuel consumption. Hydrocarbon emissions exceeded the stock values in all cases. Diagnostic data indicate that lean performance in the engine configuration tested is limited by ignition delay, cycle-to-cycle pressure variations, and cylinder-to-cylinder distribution.

  13. The feasibility study of AGR 7-position fuel testing assembly in neft position - HTR2008-58098

    International Nuclear Information System (INIS)

    Chang, G. S.; Grover, B.; Lillo, M. A.; Maki, J. T.

    2008-01-01

    In order to support the Next Generation Nuclear Plant (NGNP) Program 2018 deployment schedule, the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program must reduce the AGR fuel irradiation testing time in the Advanced Test Reactor (ATR) from approximately 2 1/2 calendar years to 1 1/2 calendar years. The AGR fuel irradiation testing requirements are: (a) burn-up of at least 14% FIMA; (b) Fast neutron fluence (E> 0.18 MeV) - maximum 25 n/m 2 ; (c) limit of fission power density is 350 W/cc; and (d) irradiation time 1.0 MeV) to Thermal (E 0.18 MeV) results indicate that the average fuel compact burnup and fast neutron fluence reach 14.79% FIMA and 4.16 x 10 25 n/m 2 , respectively. The fuel compact peak burnup reached 16.68% FIMA with corresponding fast neutron fluence for that fuel compact of 5.06 x 10 25 n/m 2 , which satisfied the fuel testing requirements. It is therefore concluded that accelerating the AGR fuel testing using the proposed AGR 7-position fuel test configuration in the NEFT is very feasible. (authors)

  14. Behaviour of HTGR coated fuel particles at high-temperature tests

    International Nuclear Information System (INIS)

    Chernikov, A.S.; Lyutikov, R.A.; Kurbakov, S.D.; Repnikov, V.M.; Khromonozhkin, V.V.; Soloviyov, G.I.

    1990-01-01

    At the temperature range 1200-2600 deg. C prereactor tests of TRISO fuel particles on the base of UO 2 , UC x O y and UO 2 +2Al 2 O 3 . SiO 2 kernels, and also fuel particle models with ZrC kernels were performed. Isothermal annealings carried out at temperatures of 1400-2600 deg. C, thermogradient ones at 1200-2200 deg. C (Δ T = 200-1200 deg. C/cm). It is shown that at heating to 2200 deg. C integrity of fuel particles is limited by different thermal expansion of PyC and SiC coatings, and also by thermal dissociation of SiC. At higher temperatures the failure is caused by development of high pressures within weakened fuel particles. It is found that uranium migration from alloyed fuel (UC x O y , UO 2 +2Al 2 O 3 .SiO 2 ) in the process of annealing is higher than that from UO 2 . (author)

  15. Eucalyptus Biodiesel as an Alternative to Diesel Fuel: Preparation and Tests on DI Diesel Engine

    Directory of Open Access Journals (Sweden)

    Lyes Tarabet

    2012-01-01

    Full Text Available Nowadays, the increasing oil consumption throughout the world induces crucial economical, security, and environmental problems. As a result, intensive researches are undertaken to find appropriate substitution to fossil fuels. In view of the large amount of eucalyptus trees present in arid areas, we focus in this study on the investigation of using eucalyptus biodiesel as fuel in diesel engine. Eucalyptus oil is converted by transesterification into biodiesel. Eucalyptus biodiesel characterization shows that the physicochemical properties are comparable to those of diesel fuel. In the second phase, a single cylinder air-cooled, DI diesel engine was used to test neat eucalyptus biodiesel and its blends with diesel fuel in various ratios (75, 50, and 25 by v% at several engine loads. The engine combustion parameters such as peak pressure, rate of pressure rise, and heat release rate are determined. Performances and exhaust emissions are also evaluated at all operating conditions. Results show that neat eucalyptus biodiesel and its blends present significant improvements of carbon monoxide, unburned hydrocarbon, and particulates emissions especially at high loads with equivalent performances to those of diesel fuel. However, the NOx emissions are slightly increased when the biodiesel content is increased in the blend.

  16. Eucalyptus biodiesel as an alternative to diesel fuel: preparation and tests on DI diesel engine.

    Science.gov (United States)

    Tarabet, Lyes; Loubar, Khaled; Lounici, Mohand Said; Hanchi, Samir; Tazerout, Mohand

    2012-01-01

    Nowadays, the increasing oil consumption throughout the world induces crucial economical, security, and environmental problems. As a result, intensive researches are undertaken to find appropriate substitution to fossil fuels. In view of the large amount of eucalyptus trees present in arid areas, we focus in this study on the investigation of using eucalyptus biodiesel as fuel in diesel engine. Eucalyptus oil is converted by transesterification into biodiesel. Eucalyptus biodiesel characterization shows that the physicochemical properties are comparable to those of diesel fuel. In the second phase, a single cylinder air-cooled, DI diesel engine was used to test neat eucalyptus biodiesel and its blends with diesel fuel in various ratios (75, 50, and 25 by v%) at several engine loads. The engine combustion parameters such as peak pressure, rate of pressure rise, and heat release rate are determined. Performances and exhaust emissions are also evaluated at all operating conditions. Results show that neat eucalyptus biodiesel and its blends present significant improvements of carbon monoxide, unburned hydrocarbon, and particulates emissions especially at high loads with equivalent performances to those of diesel fuel. However, the NOx emissions are slightly increased when the biodiesel content is increased in the blend.

  17. On possibility of application of the parallel-mixed type coolant flow scheme to NPP steam generators linked with superheaters

    International Nuclear Information System (INIS)

    Malkis, V.A.; Lokshin, V.A.

    1983-01-01

    Optimum distribution of the coolant straight-through flow between the superheater, evaporator and economizer is determined and the parallel-mixed type flow scheme is compared with other schemes. The calculations are performed for the 250 MW(e) steam generator for the WWER-1000 reactor unit the inlet and outlet primary coolant temperature of which is 324 and 290 deg C, respectively, while the feed water and saturation temperatures are 220 and 278.5 deg C, respectively. The rated superheating temperature is 300 deg C. The comparison of different schemes has been performed according to the average temperature head value at the steam-generator under the condition of equality as well as essential difference in the heat transfer coefficients in certain steam-generator sections. The calculations have shown that the use of parallel-mixed type flow permits to essentially increase the temperature head of the steam generator. At a constant heat transfer coefficient in all steam generator sections the highest temperature head is reached. At relative flow rates in the steam generator, economizer and evaporator equal to 6, 8 and 86%, respectively. The superheated steam generator temperature head in this case by 12% exceeds the temperature head of the WWER-1000 reactor unit wet steam generator. In case of heat transfer coefficient reduction in the superheater by a factor of three, the choice of the primary coolant, optimum distribution permits to maintain the steam generator temperature head at the level of the WWER-1000 reactor unit wet-steam steam generator. The use of the parallel-mixed type flow scheme permits to design a steam generator of slightly superheated steam for the parameters of the WWER-1000 unit

  18. HIGH-TEMPERATURE SAFETY TESTING OF IRRADIATED AGR-1 TRISO FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Stempien, John D.; Demkowicz, Paul A.; Reber, Edward L.; Chrisensen, Cad L.

    2016-11-01

    High-Temperature Safety Testing of Irradiated AGR-1 TRISO Fuel John D. Stempien, Paul A. Demkowicz, Edward L. Reber, and Cad L. Christensen Idaho National Laboratory, P.O. Box 1625 Idaho Falls, ID 83415, USA Corresponding Author: john.stempien@inl.gov, +1-208-526-8410 Two new safety tests of irradiated tristructural isotropic (TRISO) coated particle fuel have been completed in the Fuel Accident Condition Simulator (FACS) furnace at the Idaho National Laboratory (INL). In the first test, three fuel compacts from the first Advanced Gas Reactor irradiation experiment (AGR-1) were simultaneously heated in the FACS furnace. Prior to safety testing, each compact was irradiated in the Advanced Test Reactor to a burnup of approximately 15 % fissions per initial metal atom (FIMA), a fast fluence of 3×1025 n/m2 (E > 0.18 MeV), and a time-average volume-average (TAVA) irradiation temperature of about 1020 °C. In order to simulate a core-conduction cool-down event, a temperature-versus-time profile having a peak temperature of 1700 °C was programmed into the FACS furnace controllers. Gaseous fission products (i.e., Kr-85) were carried to the Fission Gas Monitoring System (FGMS) by a helium sweep gas and captured in cold traps featuring online gamma counting. By the end of the test, a total of 3.9% of an average particle’s inventory of Kr-85 was detected in the FGMS traps. Such a low Kr-85 activity indicates that no TRISO failures (failure of all three TRISO layers) occurred during the test. If released from the compacts, condensable fission products (e.g., Ag-110m, Cs-134, Cs-137, Eu-154, Eu-155, and Sr-90) were collected on condensation plates fitted to the end of the cold finger in the FACS furnace. These condensation plates were then analyzed for fission products. In the second test, five loose UCO fuel kernels, obtained from deconsolidated particles from an irradiated AGR-1 compact, were heated in the FACS furnace to a peak temperature of 1600 °C. This test had two

  19. Proceedings of the international meeting on research and test reactor core conversions from HEU to LEU fuels

    International Nuclear Information System (INIS)

    1983-09-01

    Separate abstracts have been prepared for each paper presented in the following areas of interest: (1) fuel development; (2) post-irradiation examinations; (3) reprocessing; (4) thermite reaction; (5) fuel fabrication; (6) element tests; (7) core tests; (8) criticals; (9) shipping; and (10) reactors and methods

  20. Analyses of the Anticipated Operational Occurrences for the HANARO Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-12-15

    The analyses of anticipated operational occurrences of the HANARO fuel test loop have been carried out by using the MARS/FTL{sub A} code, which is a modified version of the MARS code. A critical heat flux correlation on the three rods with triangular array was implemented in the MARS/FTL{sub A} code. The correlation was obtained from the critical heat fluxes measured at a test section, which is the same geometry of the in-pile test section of the HANARO fuel test loop. The anticipated operational occurrences of the HANARO fuel test loop are the inadvertent closure of the isolation valves, the over-power transient of the HANARO, the stuck open of the safety valves, and the loss of HANARO class IV power. A minimum DNBR (Departure from Nucleate Boiling Ratio) was predicted in the inadvertent closure of the isolation valves. It is indicated that the minimum DNBR of 1.85 is greater than the design limit DNBR of 1.39. The maximum coolant pressure calculated in the anticipated operational occurrences is also less than the 110 percents of the design pressure.