WorldWideScience

Sample records for fuel failure observed

  1. Analytical criteria for fuel failure modes observed in reactivity initiated accidents

    Luxat, J.C.

    2005-01-01

    The behaviour of nuclear fuel subjected to a short duration power pulse is of relevance to LWR and CANDU reactor safety. A Reactivity Initiated Accident (RIA) in an LWR would subject fuel to a short duration power pulse of large amplitude, whereas in CANDU a large break Loss of Coolant Accident (LOCA) would subject fuel to a longer duration, lower amplitude power excursion. The energy generated in the fuel during the power pulse is a key parameter governing the fuel response. This paper reviews the various power pulse tests that have been conducted in research reactors over the past three decades and summarizes the fuel failure modes that that have been observed in these tests. A simple analytical model is developed to characterize fuel behaviour under power pulse conditions and the model is applied to assess the experimental data from the power pulse tests. It is shown that the simple model provides a good basis for establishing criteria that demarcate the observed fuel failure modes for the various fuel designs that have been used in these tests. (author)

  2. High-frequency data observations from space shuttle main engine low pressure fuel turbopump discharge duct flex joint tripod failure investigation

    Zoladz, T. F.; Farr, R. A.

    1991-01-01

    Observations made by Marshall Space Flight Center (MSFC) engineers during their participation in the Space Shuttle Main Engine (SSME) low pressure fuel turbopump discharge duct flex joint tripod failure investigation are summarized. New signal processing techniques used by the Component Assessment Branch and the Induced Environments Branch during the failure investigation are described in detail. Moreover, nonlinear correlations between frequently encountered anomalous frequencies found in SSME dynamic data are discussed. A recommendation is made to continue low pressure fuel (LPF) duct testing through laboratory flow simulations and MSFC-managed technology test bed SSME testing.

  3. No significant fuel failures (NSFF)

    Domaratzki, Z.

    1979-01-01

    It has long been recognized that no emergency core cooling system (ECCS) could be absolutely guaranteed to prevent fuel failures. In 1976 the Atomic Energy Control Board decided that the objective for an ECCS should be to prevent fuel failures, but if the objective could not be met it should be shown that the consequences are acceptable for dual failures comprising any LOCA combined with an assumed impairment of containment. Out of the review of the Bruce A plant came the definition of 'no significant fuel failures': for any postulated LOCA combined with any one mode of containment impairment the resultant dose to a person at the edge of the exclusion zone is less than the reference dose limits for dual failures

  4. Regulatory experience with fuel failures in Switzerland

    Adam, L.

    2015-01-01

    In this paper the main ENSI activities like: supervision of reactor and radiation safety and security; supervision of safety of transports of nuclear materials and assess the safety of proposed solutions for the geological disposal are listed. Recent events concerning the reactor core, common causes for fuel failures, findings during inspections and potential root cause for fuel failures are discussed. Management of fuel failures, started from reporting of the event – evaluation of the need of imminent action; identification of the fuel element if possible till evaluation by the plant and fuel vendor and allowance by ENSI for repair of the fuel element and definition of measures (short and long term) are also presented. The following Conclusions by ENSI about status of fuel failures are made: 1) Number of fuel failures was reduced regardless more economic operation in all plants; 2) Old PWR and BWR reactors achieved 15 to 29 years operation without leakers, but two minor fuel damage during fuel handling appeared; 3) Newer plants are not better in achieving operation without leakers than older plants; 4) Technical improvements at fuel elements parallel to changes in operation strategy and improvements in manufacturing quality but single effects difficult to judge. The issues about how to implement “Zero Failure Rates” in regulations and how to achieve “Zero Failure Rates” as well as some future measures by ENSI are discussed

  5. Hydriding failure in water reactor fuel elements

    Sah, D.N.; Ramadasan, E.; Unnikrishnan, K.

    1980-01-01

    Hydriding of the zircaloy cladding has been one of the important causes of failure in water reactor fuel elements. This report reviews the causes, the mechanisms and the methods for prevention of hydriding failure in zircaloy clad water reactor fuel elements. The different types of hydriding of zircaloy cladding have been classified. Various factors influencing zircaloy hydriding from internal and external sources in an operating fuel element have been brought out. The findings of post-irradiation examination of fuel elements from Indian reactors, with respect to clad hydriding and features of hydriding failure are included. (author)

  6. Failure analysis for WWER-fuel elements

    Boehmert, J.; Huettig, W.

    1986-10-01

    If the fuel defect rate proves significantly high, failure analysis has to be performed in order to trace down the defect causes, to implement corrective actions, and to take measures of failure prevention. Such analyses are work-consuming and very skill-demanding technical tasks, which require examination methods and devices excellently developed and a rich stock of experience in evaluation of features of damage. For that this work specifies the procedure of failure analyses in detail. Moreover prerequisites and experimental equipment for the investigation of WWER-type fuel elements are described. (author)

  7. Analysis of fuel operational reliability and fuel failures

    Smiesko, I.

    1999-01-01

    In this lecture the fuel failure (loss of fuel rod (cladding) integrity, corruption of second barrier for fission product release from duel and their consequences (increase of primary coolant activity; increase of fission product releases to environment; increase of rad-waste activities and potential increase of personnel exposure) are discussed

  8. Fuel failure detection in operating reactors

    Seigel, B.; Hagen, H.H.

    1977-12-01

    Activity detectors in commercial BWRs and PWRs are examined to determine their capability to detect a small number of fuel rod failures during reactor operation. The off-gas system radiation monitor in a BWR and the letdown line radiation monitor in a PWR are calculated to have this capability, and events are cited that support this analysis. Other common detectors are found to be insensitive to small numbers of fuel failures. While adequate detectors exist for normal and transient operation, those detectors would not perform rapidly enough to be useful during accidents; in most accidents, however, primary system sensors (pressure, temperature, level) would provide adequate warning. Advanced methods of fuel failure detection are mentioned

  9. Failure mechanisms for compacted uranium oxide fuel cores

    Berghaus, D.G.; Peacock, H.B.

    1980-01-01

    Tension, compression, and shear tests were performed on test specimens of aluminum-clad, compacted powder fuel cores to determine failure mechanisms of the core material. The core, which consists of 70% uranium oxide in an aluminum matrix, frequently fails during post-extrusion drawing. Tests were conducted to various strain levels up to failure of the core. Sections were made of tested specimens to microscopically study initiation of failure. Two failure modes wee observed. Tensile failure mode is initiated by prior tensile failure of uranium oxide particles with the separation path strongly influenced by the arrangement of particles. Delamination mode consists of the separation of laminae formed during extrusion of tubes. Separation proceeds from fine cracks formed parallel to the laminae. Tensile failure mode was experienced in tension and shear tests. Delamination mode was produced in compression tests

  10. Timing analysis of PWR fuel pin failures

    Jones, K.R.; Wade, N.L.; Katsma, K.R.; Siefken, L.J.; Straka, M.

    1992-09-01

    Research has been conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B ampersand W) design (Oconee) and a Westinghouse (W) four-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin bumup, axial peaking factor, break size, emergency core cooling system availability, and main coolant pump trip on these times. The analysis was performed using the following codes: FRAPCON-2, for the calculation of steady-state fuel behavior; SCDAP/RELAP5/MOD3 and TRACPF1/MOD1, for the calculation of the transient thermal-hydraulic conditions in the reactor system; and FRAP-T6, for the calculation of transient fuel behavior. In addition to the calculation of fuel pin failure timing, this analysis provides a comparison of the predicted results of SCDAP/RELAP5/MOD3 and TRAC-PFL/MOD1 for large-break LOCA analysis. Using SCDAP/RELAP5/MOD3 thermal-hydraulic data, the shortest time intervals calculated between initiation of containment isolation and fuel pin failure are 10.4 seconds and 19.1 seconds for the B ampersand W and W plants, respectively. Using data generated by TRAC-PF1/MOD1, the shortest intervals are 10.3 seconds and 29.1 seconds for the B ampersand W and W plants, respectively. These intervals are for a double-ended, offset-shear, cold leg break, using the technical specification maximum peaking factor and applied to fuel with maximum design bumup. Using peaking factors commensurate widi actual bumups would result in longer intervals for both reactor designs. This document also contains appendices A through J of this report

  11. Analysis of transient fuel failure mechanisms: selected ANL programs

    Deitrich, L.W.

    1975-01-01

    Analytical programs at Argonne National Laboratory related to fuel pin failure mechanisms in fast-reactor accident transients are described. The studies include transient fuel pin mechanics, mechanics of unclad fuel, and mechanical effects concerning potential fuel failure propagation. (U.S.).

  12. Method of detecting a fuel element failure

    Cohen, P.

    1975-01-01

    A method is described for detecting a fuel element failure in a liquid-sodium-cooled fast breeder reactor consisting of equilibrating a sample of the coolant with a molten salt consisting of a mixture of barium iodide and strontium iodide (or other iodides) whereby a large fraction of any radioactive iodine present in the liquid sodium coolant exchanges with the iodine present in the salt; separating the molten salt and sodium; if necessary, equilibrating the molten salt with nonradioactive sodium and separating the molten salt and sodium; and monitoring the molten salt for the presence of iodine, the presence of iodine indicating that the cladding of a fuel element has failed. (U.S.)

  13. Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels

    Lu, Hongbing; Bukkapatnam, Satish; Harimkar, Sandip; Singh, Raman; Bardenhagen, Scott

    2014-01-01

    Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear

  14. Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels

    Lu, Hongbing [Univ. of Texas, Austin, TX (United States); Bukkapatnam, Satish; Harimkar, Sandip; Singh, Raman; Bardenhagen, Scott

    2014-01-09

    Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear

  15. Method of detecting fuel failure in FBR type reactor and method of estimating fuel failure position

    Sonoda, Yukio; Tamaoki, Tetsuo

    1989-01-01

    Noise components in a normal state contained in detection signals from delayed neutron monitors disposed to a coolant inlet, etc. of an intermediate heat exchanger are forecast by self-recurring model and eliminated, and resultant detection signals are monitored thereby detecting fuel failure high sensitivity. Subsequently, the reactor is controlled to a low power operation state and a new self-recurring model to the detection signals from the delayed neutron monitors are prepared. Then, noise components in this state are removed and control rods near the delayed neutron monitors are extracted in a short stroke successively to examine the change of response of the delayed neutron monitors. Accordingly, the failed position for each of the fuels can be estimated at a level of one fuel assembly or a level of several assemblies containing the above-mentioned fuel assembly. Since the fuel failure can be detected at a high sensitivity and the position can be estimated, diffusion of abnormality can be prevented and plant shutdown for fuel exchange can be minimized. (I.S.)

  16. Fuel failure detection and location in LMFBRs

    Jacobi, S.

    1982-06-01

    The Specialists' Meeting on 'Fuel Failure Detection and Location in LMFBRs' was held at the Kernforschungszentrum Karlsruhe, Federal Republic of Germany, on 11-14 May 1981. The meeting was sponsored by the International Atomic Energy Agency (IAEA) on the recommendation of the International Working Group on Fast Reactors (IWGFR).The purpose of the meeting was to review and discuss methods and experience in the detection and location of failed fuel elements and to recommend future development. The technical sessions were divided into five topical sessions as follows: 1. Reactor Intrumentation, 2. Experience Gained from LMFBRs, 3. In-pile Experiments, 4. Models and Codes, 5. Future Programs. During the meeting papers were presented by the participants on behalf of their countries or organizations. Each presentation was followed by an open discussion in the subject covered by the presentation. After the formal sessions were completed, a final discussion session was held and general conclusions and recommendationswere reached. Session summaries, general conclusions and recommendations, the agenda of the meeting and the list of participants are given. (orig./RW)

  17. Operation experiences of JOYO fuel failure detection system

    Tamura, Seiji; Hikichi, Takayoshi; Rindo, Hiroshi.

    1982-01-01

    Monitoring of fuel failure in the experimental fast reactor JOYO is provided by two different methods, which are cover gas monitoring (FFDCGM) by means of a precipitator, and delayed neutron monitoring (FFDDNM) by means of neutron detectors. The interpretation of signals which were obtained during the reactor operation for performance testings, was performed. The countrate of the CGM is approximately 120 cps at 75MW operation, whose sources are due to Ne 23 , Ar 41 , and Na 24 . And the countrate of the DNM is approximately 2300 cps at 75MW operation which is mainly due to leakage neutron from the core. With those background of the systems, alarm level for monitoring was set at several times of each background level. The reactor has been operated for 5 years, the burn-up of the fuel is 40,000 MWD/T at the most. No trace of any fuel failure has been observed. The fact is also proven by the results of cover gas and sodium sampling analysis. In order to evaluate sensitivity of the FFD systems, a preliminary simulation study has been performed. According to the results, a signal level against one pin failure of 0.5 mm 2 hole may exceed the alarm level of the FFDCGM system. (author)

  18. Failure position detection device for nuclear fuel rod

    Ishida, Takeshi; Higuchi, Shin-ichi; Ito, Masaru; Matsuda, Yasuhiko

    1987-01-01

    Purpose: To easily detect failure position of a nuclear fuel rod by relatively moving an air-tightly shielded detection portion to a fuel rod. Constitution: For detecting the failure position of a leaked fuel assembly, the fuel assembly is dismantled and a portion of withdrawn fuel rod is air-tightly sealed with an inspection portion. The inside of the inspection portion is maintained at a pressure-reduced state. Then, in a case if failed openings are formed at a portion sealed by the inspection portion in the fuel rod, FP gases in the fuel rod are released based on the reduced pressure and the FP gases are detected in the detection portion. Accordingly, by relatively moving the detection portion to the fuel rod, the failure position can be detected. (Yoshino, Y.)

  19. Failure position detection device for nuclear fuel rod

    Ishida, Takeshi; Higuchi, Shin-ichi; Ito, Masaru; Matsuda, Yasuhiko

    1987-03-24

    Purpose: To easily detect failure position of a nuclear fuel rod by relatively moving an air-tightly shielded detection portion to a fuel rod. Constitution: For detecting the failure position of a leaked fuel assembly, the fuel assembly is dismantled and a portion of withdrawn fuel rod is air-tightly sealed with an inspection portion. The inside of the inspection portion is maintained at a pressure-reduced state. Then, in a case if failed openings are formed at a portion sealed by the inspection portion in the fuel rod, FP gases in the fuel rod are released based on the reduced pressure and the FP gases are detected in the detection portion. Accordingly, by relatively moving the detection portion to the fuel rod, the failure position can be detected. (Yoshino, Y.).

  20. Achieving zero fuel failure rates at Armenian NPP

    Muradyan, T.

    2015-01-01

    In spite of the zero fuel failure rates in Armenian NPP there is a continued high level of interest. The generally accepted goal of achieving a zero failure rate requires detailed knowledge of existing failure mechanisms, their root causes and remedies. In this paper the foreign material management; water-chemistry regime; refuel machine management system and the transition into the use of vibration proof fuel of average enrichment 3,82% are presented

  1. Use of fuel failure correlations in accident analysis

    O'Dell, L.D.; Baars, R.E.; Waltar, A.E.

    1975-05-01

    The MELT-III code for analysis of a Transient Overpower (TOP) accident in an LMFBR is briefly described, including failure criteria currently applied in the code. Preliminary results of calculations exploring failure patterns in time and space in the reactor core are reported and compared for the two empirical fuel failure correlations employed in the code. (U.S.)

  2. Failure analysis of carbide fuels under transient overpower (TOP) conditions

    Nguyen, D.H.

    1980-06-01

    The failure of carbide fuels in the Fast Test Reactor (FTR) under Transient Overpower (TOP) conditions has been examined. The Beginning-of-Cycle Four (BOC-4) all-oxide base case, at $.50/sec ramp rate was selected as the reference case. A coupling between the advanced fuel performance code UNCLE-T and HCDA Code MELT-IIIA was necessary for the analysis. UNCLE-T was used to determine cladding failure and fuel preconditioning which served as initial conditions for MELT-III calculations. MELT-IIIA determined the time of molten fuel ejection from fuel pin

  3. Damage and failure of unirradiated and irradiated fuel rods tested under film boiling conditions

    Mehner, A.S.; Hobbins, R.R.; Seiffert, S.L.; MacDonald, P.E.; McCardell, R.K.

    1979-01-01

    Power-cooling-mismatch experiments are being conducted as part of the Thermal Fuels Behavior Program in the Power Burst Facility at the Idaho National Engineering Laboratory to evaluate the behavior of unirradiated and previously irradiated light water reactor fuel rods tested under stable film boiling conditions. The observed damage that occurs to the fuel rod cladding and the fuel as a result of film boiling operation is reported. Analyses performed as a part of the study on the effects of operating failed fuel rods in film boiling, and rod failure mechanisms due to cladding embrittlement and cladding melting upon being contacted by molten fuel are summarized

  4. Fuel-pin cladding transient failure strain criterion

    Bard, F.E.; Duncan, D.R.; Hunter, C.W.

    1983-01-01

    A criterion for cladding failure based on accumulated strain was developed for mixed uranium-plutonium oxide fuel pins and used to interpret the calculated strain results from failed transient fuel pin experiments conducted in the Transient Reactor Test (TREAT) facility. The new STRAIN criterion replaced a stress-based criterion that depends on the DORN parameter and that incorrectly predicted fuel pin failure for transient tested fuel pins. This paper describes the STRAIN criterion and compares its prediction with those of the stress-based criterion

  5. Fuel and coolant motions following pin failure: EPIC models and the PBE-5S experiment

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

    1979-01-01

    The EPIC computer code has been used to analyze the post-fuel-pin-failure behavior in the PBE-5S experiment performed at Sandia Laboratories. The effects of modeling uncertainties on the calculation are examined. The calculations indicate that the majority of the piston motion observed in the test is due to the initial pressurization of the coolant channel by fuel vapor at cladding failure. A more definitive analysis requires improvements in calculational capabilities and experiment diagnostics

  6. Fuel failure assessments based on radiochemistry. Experience feedback and challenges

    Petit, C.; Ziabletsev, D.; Zeh, P.

    2015-01-01

    Significant improvements have been observed in LWR nuclear fuel reliability over the past years. As a result, the number of fuel failures in PWRs and BWRs has recently dramatically decreased. Nevertheless, a few remaining challenges still exist. One of them is that the industry has recently started seeing a relatively new type of fuel failure, so-called 'weak leak failures', which could be characterized by a very small release of gaseous fission products and essentially almost zero release of iodines or any other soluble fission products in the reactor coolant. Correspondingly, the behavior of these weak leakers does not follow typical behavior of a conventional leaker characterized by a proportionality of the amount of released Xe 133 related to the failed rod power. Instead, for a weak leaker, the activity of Xe 133 is directly correlated to the size of the cladding defects of the leaker. The presence of undetected weak leaker in the core may lead to carryover of a leaker into the subsequent cycle. Even if the presence of weak leaker in the core is suspected, it typically requires more effort to identify the leaker which could result in extended duration of the outage and ultimately to economic losses to the utility operating the reactor. To effectively deal with this issue the industry has been facing, several changes have been recently realized, which are different from the methodology of dealing with conventional leaker. These changes include new assessment methods, the need for improved sipping techniques to better identify low release leakers, and correspondingly better equipment to be able to locate small clad defects associated with weak leaker, such as sensitive localization device of failed rods, sensitive eddy current coil for the failed rod, ultra high definition cameras for the failed rod examination and experienced fuel reliability engineers performing cause of failure and rood cause research and analyses. Ultimately, the destructive

  7. Design of fuel failure detection system for multipurpose reactor GA. Siwabessy

    Sujalmo Saiful; Kuntoro Iman; Sato, Mitsugu; Isshiki, Masahiko.

    1992-01-01

    A fuel failure detection system (FFDS) has been designed for the Reactor GA. Siwabessy. The FFDS is aimed to detect fuel failure by observing delayed neutron released by fission products such as N-17, I-137, Br-87 and Br-88 in the primary cooling system. The delayed neutrons will be detected by using four neutron detectors, type BF-3, which are located inside a Sampling Tank. The detector location has been determined and the location is associated with the transit time from the reactor core outlet to the Sampling Tank, which is approximately 60 seconds. The neutron detection efficiency was calculated by using a computer code named MORSE. The FFDS has the capability to detect as quickly as possible, even a small failure of a fuel element occurring in the reactor core. Therefore the presence of FFDS in a reactor must be considered, in order to prevent further progress if the fuel failure occurs. (author)

  8. Evaluation of the behavior of waterlogged fuel rod failures in LWRs

    Siegel, B.

    1977-11-01

    A summary of the available information on waterlogged fuel rod failures is presented. The information includes experimental results from waterlogging tests in research reactors, observations of waterlogging failures in commercial reactors, and reactor vendor assessments. It is concluded that (a) operating restrictions to reduce pellet/cladding interactions also reduce the potential for waterlogging failures during transients, (b) tests to simulate accident conditions produced the worst waterlogging failures, and (c) there is no apparent threat from waterlogging failures to the overall coolability of the core or to safe reactor shutdown

  9. Nuclear fuel pin controlled failure device

    Schlenker, L.D.

    1975-01-01

    Each fuel pin of a fuel assembly for a water-cooled nuclear reactor is provided with means for rupturing the cladding tube at a predetermined location if an abnormal increase in pressure of the gases present occurs due to a loss-of-coolant accident. Preferably all such rupture means are oriented to minimize the hydraulic resistance to the flow of emergency core coolant such as all rupture means pointing in the same direction. Rupture means may be disposed at different elevations in adjacent fuel pins and, further, fuel pins may be provided with two or more rupture means, one of which is in the upper portion of the fuel pin. Rupture means are mechanical as by providing a locally weakened condition of a controlled nature in the cladding. (U.S.)

  10. Water sampling device for detecting fuel failure

    Masubuchi, Yukio.

    1997-01-01

    A notched portion is formed at the lower end of an outer cap, and an extensible air bag is disposed being in contact with the inner side of the notched portion. A compressed air is sent into the outer gap through an air supply pipe to urge coolants thereby lowering the water level. A portion of the compressed air gets out of the outer gap from the notched portion, and if air bubbles are observed on the surface of coolants in a pressure vessel of a reactor, the outer cap is confirmed to be attached to the upper lattice plate. Compressed air is supplied to the air bag to close the notched portion. Then, coolants are sucked from a water level confirmation pipe. The level of coolants is further lowered, and the compressed air is sucked from the water level confirmation pipe instead of the coolants. Then, the level of the coolants at the inner side of the inner cap is confirmed to be made lower than the upper end of the channel box of a reactor fuel assembly. Then, coolants in the channel box are sampled, as a specimen water, through a water sampling pipe. (I.N.)

  11. A probabilistic approach for RIA fuel failure criteria

    Carlo Vitanza, Dr.

    2008-01-01

    Substantial experimental data have been produced in support of the definition of the RIA safety limits for water reactor fuels at high burn up. Based on these data, fuel failure enthalpy limits can be derived based on methods having a varying degree of complexity. However, regardless of sophistication, it is unlikely that any deterministic approach would result in perfect predictions of all failure and non failure data obtained in RIA tests. Accordingly, a probabilistic approach is proposed in this paper, where in addition to a best estimate evaluation of the failure enthalpy, a RIA fuel failure probability distribution is defined within an enthalpy band surrounding the best estimate failure enthalpy. The band width and the failure probability distribution within this band are determined on the basis of the whole data set, including failure and non failure data and accounting for the actual scatter of the database. The present probabilistic approach can be used in conjunction with any deterministic model or correlation. For deterministic models or correlations having good prediction capability, the probability distribution will be sharply increasing within a narrow band around the best estimate value. For deterministic predictions of lower quality, instead, the resulting probability distribution will be broad and coarser

  12. Mode of failure of LMFBR fuel pins

    Washburn, D.F.

    1975-01-01

    The objectives of the irradiation test described were to evaluate mixed-oxide fuel performance and to confirm the design adequacy of the FFTF fuel pins. After attainment of the initial objectives the irradiation of several of the original fuel pins was continued until a cladding breach occurred. The consequences of a cladding breach were evaluated by reconstituting the original 37-pin subassembly into two 19-pin subassemblies after a burnup at 50,000 MWd/MTM (5.2 a/o). The original pins were supplemented with fresh pins as necessary. Irradiation of the subassemblies was continued until a cladding breach occurred. Results are presented and discussed

  13. Fuel rod failure detection method and system

    Assmann, H.; Janson, W.; Stehle, H.; Wahode, P.

    1975-01-01

    The inventor claims a method for the detection of a defective fuel rod cladding tube or of inleaked water in the cladding tube of a fuel rod in the fuel assembly of a pressurized-water reactor. The fuel assembly is not disassembled but examined as a whole. In the examination, the cladding tube is heated near one of its two end plugs, e.g. with an attached high-frequency inductor. The water contained in the cladding tube evaporates, and steam bubbles or a condensate are detected by the ultrasonic impulse-echo method. It is also possible to measure the delay of the temperature rise at the end plug or to determine the cooling energy required to keep the end plug temperature stable and thus to detect water ingression. (DG/AK) [de

  14. Fuel failure detection and location methods in CAGRs

    Harris, A.M.

    1982-06-01

    The release of fission products from AGR fuel failures and the way in which the signals from such failures must be detected against the background signal from uranium contamination of the fuel is considered. Theoretical assessments of failure detection are used to show the limitations of the existing Electrostatic Wire Precipitator Burst Can Detection system (BCD) and how its operating parameters can be optimised. Two promising alternative methods, the 'split count' technique and the use of iodine measurements, are described. The results of a detailed study of the mechanical and electronic performance of the present BCD trolleys are given. The limited experience of detection and location of two fuel failures in CAGR using conventional and alternative methods is reviewed. The larger failure was detected and located using the conventional BCD equipment with a high confidence level. It is shown that smaller failures may not be easy to detect and locate using the current BCD equipment, and the second smaller failure probably remained in the reactor for about a year before it was discharged. The split count technique used with modified BCD equipment was able to detect the smaller failure after careful inspection of the data. (author)

  15. Protecting AREVA ATRIUM™ BWR fuel from debris fretting failure

    Cole, Steven E.; Garner, Norman L.; Lippert, Hans-Joachim; Graebert, Rüdiger; Mollard, Pierre; Hahn, Gregory C.

    2014-01-01

    Historically, debris fretting has been the leading cause of fuel rod failure in BWR fuel assemblies, costing the industry millions of dollars in lost generation and negatively impacting the working area of plant site personnel. In this paper the focus will be on recent BWR fuel product innovation designed to eliminate debris related failures. Experience feedback from more than three decades of operation history with non-line-of-sight FUELGUARD™ lower tie plate debris filters will be presented. The development and relative effectiveness of successive generations of filtration technology will be discussed. It will be shown that modern, state of the art debris filters are an effective defense against debris fretting failure. Protective measures extend beyond inlet nozzle debris filters. The comprehensive debris resistance features built into AREVA’s newest fuel design, the ATRIUM™ 11, reduce the overall risk of debris entrapment as well as providing a degree of protection from debris that may fall down on the fuel assembly from above, e.g., during refueling operations. The positive recent experience in a debris sensitive plant will be discussed showing that the combination of advanced fuel technology and a robust foreign material exclusion program at the reactor site can eliminate the debris fretting failure mechanism. (author)

  16. Fuel element failures caused by iodine stress corrosion

    Videm, K.; Lunde, L.

    1976-01-01

    Sections of unirradiated cladding tubes were plugged in both ends by mechanical seals and internally pressurized with argon containing iodine. The time to failure and the strain at failure as a function of stress was determined for tubing with different heat treatments. Fully annealed tubes suffer cracking at the lowest stress but exhibit the largest strains at failure. Elementary iodine is not necessary for stress corrosion: small amounts of iodides of zirconium, iron and aluminium can also give cracking. Moisture, however, was found to act as an inhibitor. A deformation threshold exists below which stress corrosion failure does not occur regardless of the exposure time. This deformation limit is lower the harder the tube. The deformation at failure is dependent on the deformation rate and has a minimum at 0.1%/hr. At higher deformation rates the failure deformation increases, but only slightly for hard tubes. Fuel was over-power tested at ramp rates varying between 0.26 to 30 W/cm min. For one series of fuel pins the failure deformations of 0.8% at high ramp rates were in good agreement with predictions based on stress corrosion experiments. For another series of experiments the failure deformation was surprisingly low, about 0.2%. (author)

  17. Fuel pin failure in the PFR/TREAT experiments

    Herbert, R.; Hunter, C.W.; Kramer, J.M.; Wood, M.H.; Wright, A.E.

    1986-01-01

    The PFR/TREAT safety testing programme involves the transient testing of fresh and pre-irradiated UK and US fuel pins. This paper summarizes the experimental and calculational results obtained to date on fuel pin failure during transient overpower (resulting from an accidental addition of resolivity) and transient undercooling followed by overpower (arising from an accidental stoppage of the primary sodium circulating pumps) accidents. Companion papers at this conference address: (I) the progress and future plans of the programme, and (II) post-failure material movements

  18. The prediction problems of VVER fuel element cladding failure theory

    Pelykh, S.N.; Maksimov, M.V.; Ryabchikov, S.D.

    2016-01-01

    Highlights: • Fuel cladding failure forecasting is based on the fuel load history and the damage distribution. • The limit damage parameter is exceeded, though limit stresses are not reached. • The damage parameter plays a significant role in predicting the cladding failure. • The proposed failure probability criterion can be used to control the cladding tightness. - Abstract: A method for forecasting of VVER fuel element (FE) cladding failure due to accumulation of deformation damage parameter, taking into account the fuel assembly (FA) loading history and the damage parameter distribution among FEs included in the FA, has been developed. Using the concept of conservative FE groups, it is shown that the safety limit for damage parameter is exceeded for some FA rearrangement, though the limits for circumferential and equivalent stresses are not reached. This new result contradicts the wide-spread idea that the damage parameter value plays a minor role when estimating the limiting state of cladding. The necessary condition of rearrangement algorithm admissibility and the criterion for minimization of the probability of cladding failure due to damage parameter accumulation have been derived, for using in automated systems controlling the cladding tightness.

  19. Mechanical modelling of transient- to- failure SFR fuel cladding

    Feria, F.; Herranz, L. E.

    2014-07-01

    The response of Sodium Fast Reactor (SFR) fuel rods to transient accident conditions is an important safety concern. During transients the cladding strain caused by the stress due to pellet cladding mechanical interaction (PCMI) can lead to failure. Due to the fact that SFR fuel rods are commonly clad with strengthened material made of stainless steel (SS), cladding is usually treated as an elastic-perfectly-plastic material. However, viscoplastic behaviour can contribute to mechanical strain at high temperature (> 1000 K). (Author)

  20. Impact of Fuel Failure on Criticality Safety of Used Nuclear Fuel

    Marshall, William J.; Wagner, John C.

    2012-01-01

    Commercial used nuclear fuel (UNF) in the United States is expected to remain in storage for considerably longer periods than originally intended (e.g., 45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, can result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF is not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on criticality safety of UNF in storage and transportation casks. This effort is primarily motivated by concerns related to the potential for fuel degradation during ES periods and transportation following ES. The criticality analyses consider representative UNF designs and cask systems and a range of fuel enrichments, burnups, and cooling times. The various failed-fuel configurations considered are designed to bound the anticipated effects of individual rod and general cladding failure, fuel rod deformation, loss of neutron absorber materials, degradation of canister internals, and gross assembly failure. The results quantify the potential impact on criticality safety associated with fuel reconfiguration and may be used to guide future research, design, and regulatory activities. Although it can be concluded that the criticality safety impacts of fuel reconfiguration during transportation subsequent to ES are manageable, the results indicate that certain configurations can result in a large increase in the effective neutron multiplication factor, k eff . Future work to inform decision making relative to which configurations are credible, and therefore need to be considered in a safety evaluation, is recommended.

  1. Failure analysis of burst tested fuel tube samples

    Padmaprabu, C.; Ramana Rao, S.V.; Srivatsava, R.K.

    2005-01-01

    The Total Circumferential Elongation (TCE) is an important parameter for evaluation of ductility of the Zircaloy-4 fuel tubes for the PHWR reactors. The TCE values of the fuel tubes were obtained using the burst testing technique. In some lots there is a variation in the values of the TCE. To investigate the reasons for such a large variation in the TCE, samples were selected at appropriate intervals and sectioned at the fractured portion. The surface morphology of the fractured surfaces was examined under Scanning Electron Microscope (SEM) equipped with Energy Dispersive Spectrometer (EDS). The morphologies show segregation of elements at specific locations. Energy dispersive spectra was obtained from those segregated particles. According to the magnitude of TCE value the samples were classified into low, intermediate and high ductility. Low ductility samples were found to contain large amount of segregations along the thickness direction of the tube. This forms a brittle region and a path for the easy crack growth along thickness direction. In the case of intermediate samples the segregation occurred in fewer locations compared to low ductile samples and also confined to the circumferential direction of the outside surface of the tube. Due to this, probability of crack formation at the surface of the tube could be high. But crack growth would be slower in the ductile matrix along the thickness direction resulting in the enhancement of TCE value compared to the low ductile sample. In the high ductile samples, the segregations were very scarce and found to be isolated and embedded in the ductile matrix. The mode of failure in these types of samples was found to be purely ductile. Cracks were found to originate solely from the micro voids in the material. As the probability of crack formation and its propagation is low, very high TCE values were observed in these samples. Microstructural observations of fractured surfaces and EDAX analysis was able to identify the

  2. Research program on conditions to failure of high burnup fuel

    NONE

    2013-08-15

    Regarding the power ramp test to verify the out-of-pile test results on hydrogen-induced cladding failure, situation of the shipping port restoration after the earthquake disaster was investigated for the overseas transportation of test fuel rods which had been interrupted. Its reopening schedule was still currently uncertain and the power ramp test plan also remained suspended. The information about the fuel irradiation performance obtained from JNES projects and international projects, etc. is prepared as database, and based on the recent findings, the fuel irradiation performance models and analysis codes are developed and/or improved. (author)

  3. Study of behavior on bonding and failure mode of pressurized and doped BWR fuel rod

    Yanagisawa, Kazuaki

    1992-03-01

    The study of transient behavior on the bonding and the failure mode was made using the pressurized/doped 8 x 8 BWR type fuel rod. The dopant was mullite minerals consisted mainly of silicon and aluminum up to 1.5 w/o. Pressurization of the fuel rod with pure helium was made to the magnitude about 0.6 MPa. As a reference, the non-pressurized/non-doped 8 x 8 BWR fuel rod and the pressurized/7 x 7 BWR fuel rod up to 0.6 MPa were prepared. Magnitude of energy deposition given to the tested fuel rods was 248, 253, and 269 cal/g·fuel, respectively. Obtained results from the pulse irradiation in NSRR are as follows. (1) It was found from the experiment that alternation of the fuel design by the adoption of pressurization up to 0.6 MPa and the use of wider gap up to 0.38 mm could avoid the dopant BWR fuel from the overall bonding. The failure mode of the present dopant fuel was revealed to be the melt combined with rupture. (2) The time of fuel failure of the pressurized/doped 8 x 8 BWR fuel defected by the melt/rupture mode is of order of two times shorter than that of the pressurized/ 7 x 7 BWR defected by the rupture mode. Failure threshold of the pressurized/doped 8 x 8 BWR BWR tended to be lower than that of non-pressurized/non-doped 8 x 8 BWR one. Cracked area of the pressurized/doped 8 x 8 BWR was more wider and magnitude of oxidation at the place is relatively larger than the other tested fuels. (3) Failure mode of the non-pressurized/ 8 x 8 BWR fuel rod was the melt/brittle accompanied with a significant bonding at failed location. While, failure mode of the pressurized/ 7 x 7 BWR fuel rod was the cladding rupture accompanied with a large ballooning. No bonding at failed location of the latter was observed. (author)

  4. Calculation of fuel pin failure timing under LOCA conditions

    Jones, K.R.; Wade, N.L.; Siefken, L.J.; Straka, M.; Katsma, K.R.

    1991-10-01

    The objective of this research was to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B ampersand W) design (Oconee) and a Westinghouse (W) 4-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin burnup, axial peaking factor, break size, emergency core cooling system (ECCS) availability, and main coolant pump trip on these items. The analysis was performed using a four-code approach, comprised of FRAPCON-2, SCDAP/RELAP5/MOD3, TRAC-PF1/MOD1, and FRAP-T6. In addition to the calculation of timing results, this analysis provided a comparison of the capabilities of SCDAP/RELAP5/MOD3 with TRAC-PF1/MOD1 for large-break LOCA analysis. This paper discusses the methodology employed and the code development efforts required to implement the methodology. The shortest time intervals calculated between initiation of containment isolation and fuel pin failure were 11.4 s and 19.1 for the B ampersand W and W plants, respectively. The FRAP-T6 fuel pin failure times calculated using thermal-hydraulic data generated by SCDAP/RELAP5/MOD3 were more conservative than those calculated using data generated by TRAC-PF1/MOD1. 18 refs., 7 figs., 4 tabs

  5. Prediction of failure enthalpy and reliability of irradiated fuel rod under reactivity-initiated accidents by means of statistical approach

    Nam, Cheol; Choi, Byeong Kwon; Jeong, Yong Hwan; Jung, Youn Ho

    2001-01-01

    During the last decade, the failure behavior of high-burnup fuel rods under RIA has been an extensive concern since observations of fuel rod failures at low enthalpy. Of great importance is placed on failure prediction of fuel rod in the point of licensing criteria and safety in extending burnup achievement. To address the issue, a statistics-based methodology is introduced to predict failure probability of irradiated fuel rods. Based on RIA simulation results in literature, a failure enthalpy correlation for irradiated fuel rod is constructed as a function of oxide thickness, fuel burnup, and pulse width. From the failure enthalpy correlation, a single damage parameter, equivalent enthalpy, is defined to reflect the effects of the three primary factors as well as peak fuel enthalpy. Moreover, the failure distribution function with equivalent enthalpy is derived, applying a two-parameter Weibull statistical model. Using these equations, the sensitivity analysis is carried out to estimate the effects of burnup, corrosion, peak fuel enthalpy, pulse width and cladding materials used

  6. BWR fuel performance under advanced water chemistry conditions – a delicate journey towards zero fuel failures – a review

    Hettiarachchi, S.

    2015-01-01

    Boiling Water Reactors (BWRs) have undergone a variety of chemistry evolutions over the past few decades as a result of the need to control stress corrosion cracking of reactor internals, radiation fields and personnel exposure. Some of the advanced chemistry changes include hydrogen addition, zinc addition, iron reduction using better filtration technologies, and more recently noble metal chemical addition to many of the modern day operating BWRs. These water chemistry evolutions have resulted in changes in the crud distribution on fuel cladding material, Co-60 levels and the Rod oxide thickness (ROXI) measurements using the conventional eddy current techniques. A limited number of Post-Irradiation Examinations (PIE) of fuel rods that exhibited elevated oxide thickness using eddy current techniques showed that the actual oxide thickness by metallography is much lower. The difference in these observations is attributed to the changing magnetic properties of the crud affecting the rod oxide thickness measurement by the eddy current technique. This paper will review and summarize the BWR fuel cladding performance under these advanced and improved water chemistry conditions and how these changes have affected the goal to reach zero fuel failures. The paper will also provide a brief summary of some of the results of hot cell PIE, results of crud composition evaluation, crud spallation, oxide thickness measurements, hydrogen content in the cladding and some fuel failure observations. (author) Key Words: Boiling Water Reactor, Fuel Performance, Hydrogen Addition, Zinc Addition, Noble Metal Chemical Addition, Zero Leakers

  7. Fuel and control rod failure behavior during degraded core accidents

    Chung, K.S.

    1984-01-01

    As a part of the pretest and posttest analyses of Light Water Reactor Source Term Experiments (STEP) which are conducted in the Transient Reactor Test (TREAT) facility, this paper investigates the thermodynamic and material behaviors of nuclear fuel pins and control rods during severe core degradation accidents. A series of four STEP tests are being performed to simulate the characteristics of the power reactor accidents and investigate the behavior of fission product release during these accidents. To determine the release rate of the fission products from the fuel pins and the control rod materials, information concerning the timing of the clad failure and the thermodynamic conditions of the fuel pins and control rods are needed to be evaluated. Because the phase change involves a large latent heat and volume expansion, and the phase change is a direct cause of the clad failure, the understanding of the phase change phenomena, particularly information regarding how much of the fuel pin and control rod materials are melted are very important. A simple energy balance model is developed to calculate the temperature profile and melt front in various heat transfer media considering the effects of natural convection phenomena on the melting and freezing front behavior

  8. Consequences of Fuel Failure on Criticality Safety of Used Nuclear Fuel

    Marshall, William J.; Wagner, John C.

    2012-09-01

    This report documents work performed for the Department of Energy's Office of Nuclear Energy (DOENE) Fuel Cycle Technologies Used Fuel Disposition Campaign to assess the impact of fuel reconfiguration due to fuel failure on the criticality safety of used nuclear fuel (UNF) in storage and transportation casks. This work was motivated by concerns related to the potential for fuel degradation during extended storage (ES) periods and transportation following ES, but has relevance to other potential causes of fuel reconfiguration. Commercial UNF in the United States is expected to remain in storage for longer periods than originally intended. Extended storage time and irradiation of nuclear fuel to high-burnup values (>45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, can result in changes to the geometric configuration of the fuel, which has safety and regulatory implications for virtually all aspects of a UNF storage and transport system's performance. The potential impact of fuel reconfiguration on the safety of UNF in storage and transportation is dependent on the likelihood and extent of the fuel reconfiguration, which is not well understood and is currently an active area of research. The objective of this work is to assess and quantify the impact of postulated failed fuel configurations on the criticality safety of UNF in storage and transportation casks. Although this work is motivated by the potential for fuel degradation during ES periods and transportation following ES, it has relevance to fuel reconfiguration due to the effects of high burnup. Regardless of the ultimate disposition path, UNF will need to be transported at some point in the future. To investigate and quantify the impact of fuel reconfiguration on criticality safety limits, which are given in terms of the effective neutron multiplication factor, a set of failed fuel

  9. [Therapeutic failure in scabies: An observational study].

    De Sainte Marie, B; Mallet, S; Gaudy-Marqueste, C; Baumstarck, K; Bentaleb, N; Loundou, A; Hesse, S; Monestier, S; Grob, J-J; Richard, M-A

    2016-01-01

    Several sources suggest an escalation of scabies in France. To describe a population of patients continuing to present with scabies despite multiple treatments in order to identify factors associated with persistence of infection. A descriptive cross-sectional study in adults and children consulting for persistent scabies despite at least one previous treatment. A standardized questionnaire explored potential sources of treatment failure. Thirty-one patients were analyzed. Initial symptoms were noted to have started between two and 52 weeks earlier (mean: 19 weeks). The mean number of prior consultations with a general practitioner was 3.1 (0-10) and 1.7 with a dermatologist (0-7). The mean number of patients per household was 3.5 (1-9). At least one dose of oral ivermectin (maximum of 6 doses per household) was prescribed for 84 % of patients (29 % of whom were not fasted at the time). Further, 74 % of patients received at least one local application of esdepallethrin and piperonyl butoxide (maximum: 5 courses), four received benzyl benzoate and two received permethrin; however, 58 % did not reapply the substance after hand washing. All households bought the prescribed treatments despite the costs. Close contacts of patients were treated in 58 % of households. Decontamination of bedding and clothing was carried out properly in 90 % of households. Persistence of infection appears to be linked to: (1) insufficient treatment of close contacts; (2) absence of a second treatment between days 7 and 14; (3) insufficient efficacy of the available treatments, doubtless due to multiple factors (intrinsic resistance of Sarcoptes, failure to repeat treatment, poor explanation of methods for dosing and application, and oral intake of treatments). Access to non-reimbursed treatments was not identified as a problem and decontamination of bedding and clothing was correctly performed in most cases. Though certain fundamental aspects of scabies treatment must be better

  10. Nonlinear observer designs for fuel cell power systems

    Gorgun, Haluk

    dynamics, and estimate not only hydrogen but also all other species in its reactors. We design nonlinear observers for the Catalytic Partial Oxidation (CPO), Water Gas Shift (WGS), and Preferential Oxidation (PROX), reactors in the FPS. The observers make use of temperature measurements (and possibly one more variable, such as pressure) to estimate the mole fractions of each species in the reactors. An advantage of these designs is that they are based on reaction invariants and do not rely on knowledge of reaction rate expressions. Finally, in part III, we illustrate how the designs of parts I and II can be incorporated in fault detection and estimation algorithms for common failures encountered in fuel cells, such as the cathode blower failure and the anode valve failure. For this task, we combine geometric tools with our observers.

  11. A collapse mode of failure in powder-filled fuel elements

    Feraday, M.A.; Chalder, G.H.

    1964-01-01

    Two swaged fuel elements containing crushed, fused UO 2 powder were irradiated in a pressurized water loop at high heat ratings (∫Kdθ = 48 w/cm). The fuel elements were 2.0 cm in diameter and were sheathed in nickel-free Zircaloy--2 of 0.038 cm thickness. One element failed when the sheath ruptured at the top of a longitudinal ridge in the sheath after a burn-up of approximately 2550 MWd/TeU. No evidence was found that outgassing of the UO 2 contributed to the failure. Dimensional and structural changes observed in the fuel elements led to the conclusion that ridging of the sheath resulted from the action of coolant pressure on the diametral clearance formed by sintering and shrinkage of the UO 2 . Failure resulted due to severe local deformation accompanying one or more power cycles following ridge formation. (author)

  12. WWER problems and perspectives of moving toward zero fuel failures and mitigation of fuel failure consequences at NPPs with WWER reactors in Russia

    Shestakov, Yu.; Semenovykh, A.

    2015-01-01

    The paper contained information on mitigation of fuel failures and perspectives of moving to ‘Zero Failure Level’. It was stated that driving to ‘zero’ failure rate should include two kinds of efforts, 1) focus on identification of failure mechanisms, and 2) implement corrective actions. It is also important to mitigate the consequences of fuel failures if that inevitably occurs. Fuel failures entail the risks of severe secondary degradation and contamination of primary circuit due to fuel washout. Significant changes of fuel operating conditions (longer fuel cycles, higher fuel burnup, power uprate) and innovations in fuel design bear the risk of higher failure rates for some period of time. Simultaneous implementation of several innovations in one nuclear utility is not advisable since it might be difficult to identify which of the innovations affected the fuel performance positively or negatively. The disadvantage of gradual implementation of corrective actions and any significant changes in operating conditions is the long time needed to evaluate the operational experience. In some cases, it may take up to 4-6 years for each significant change to reflect on the operational performance

  13. Failure analysis of electrolyte-supported solid oxide fuel cells

    Fleischhauer, Felix; Tiefenauer, Andreas; Graule, Thomas; Danzer, Robert; Mai, Andreas; Kuebler, Jakob

    2014-07-01

    For solid oxide fuel cells (SOFCs) one key aspect is the structural integrity of the cell and hence its thermo mechanical long term behaviour. The present study investigates the failure mechanisms and the actual causes for fracture of electrolyte supported SOFCs which were run using the current μ-CHP system of Hexis AG, Winterthur - Switzerland under lab conditions or at customer sites for up to 40,000 h. In a first step several operated stacks were demounted for post-mortem inspection, followed by a fractographic evaluation of the failed cells. The respective findings are then set into a larger picture including an analysis of the present stresses acting on the cell like thermal and residual stresses and the measurements regarding the temperature dependent electrolyte strength. For all investigated stacks, the mechanical failure of individual cells can be attributed to locally acting bending loads, which rise due to an inhomogeneous and uneven contact between the metallic interconnect and the cell.

  14. Some calculations of the failure statistics of coated fuel particles

    Martin, D.G.; Hobbs, J.E.

    1977-03-01

    Statistical variations of coated fuel particle parameters were considered in stress model calculations and the resulting particle failure fraction versus burn-up evaluated. Variations in the following parameters were considered simultaneously: kernel diameter and porosity, thickness of the buffer, seal, silicon carbide and inner and outer pyrocarbon layers, which were all assumed to be normally distributed, and the silicon carbide fracture stress which was assumed to follow a Weibull distribution. Two methods, based respectively on random sampling and convolution of the variations were employed and applied to particles manufactured by Dragon Project and RFL Springfields. Convolution calculations proved the more satisfactory. In the present calculations variations in the silicon carbide fracture stress caused the greatest spread in burn-up for a given change in failure fraction; kernel porosity is the next most important parameter. (author)

  15. Fuel failure monitoring system design approach for KALIMER

    Song, Soon Ja; Hwang, I. K.; Kwon, Kee Choon

    1998-01-01

    Fuel Failure Monitoring System (FFMS) detects fission gas and locates failed fuels in Liquid Metal Reactor. This system comprises three subsystems; delayed neutron monitoring, cover gas monitoring, and gas tagging. The purpose of this system is to improve the integrity and availability of the liquid metal plant. In this paper, FFMS was analyzed on detection method and compared with various existing liquid metal plants. Sampling and detecting methods were classified with specific plant types. Several technologies of them was recognized and used in most liquid metal reactors. Detection technology and analysis performance, however, must be improved because of new technology when liquid metal plant is built, but the FFMS design scheme will not be changed. Thereby this paper suggests the design to implement KALIMER(Korea Advanced LIquid MEtal Reactor) FFMS

  16. Fuel failure in water reactors: Causes and mitigation. Proceedings of a technical meeting

    2003-03-01

    The objective of this technical meeting (TM) was to review the present knowledge of the causes and mechanisms of fuel failure in water reactors during normal operational conditions. Emphasis has been given to analysis of failure causes and their mitigation by means of design as well as plant and core operation including strategies for operation with failed fuel. Some information on detection techniques (on-line monitoring and diagnostics, flux tilting, sipping techniques, etc) has also been presented. This TM presented also the progress on the above-mentioned subjects since the last meeting held in 1992 (Dimitrovgrad, Russian Federation). The topics covered in the papers were as follows: Experience feedback on fuel reliability (8 papers); Strategies to avoid or mitigate fuel failures (4 papers); Experimental studies on fuel failures and degradation mechanisms (4 papers); Modelling of fuel failure mechanisms (3 papers); Detection and monitoring during operation or outage (4 papers); Modelling and assessment of fuel failures (3 papers)

  17. Effects of pellet shape on the fuel failure behavior under a RIA condition

    Hosokawa, Takanori; Hoshi, Tsutao; Yanagihara, Satoshi; Iwamura, Takamichi; Orita, Yoshihiko.

    1980-10-01

    The two types of fuel rods with different pellet shaped, i.e. flat pellets and dished pellets, were tested in the NSRR to investigate the effects of pellet shapes on the fuel failure behavior under an RIA condition and the results were compared with those of the chamfered pellet fuel rods which are used as the reference rod in the NSRR experiments. In addition, the deformation of pellets due to thermal expansion is calculated by using an FEM computer code. Through the above results, following conclusions are obtained. (1) In the experiments, insignificant differences on the cladding surface temperature responses and the appearance of post-irradiated rods are observed in each type of rods. (2) Evident differences on the deformation of fuel pellets have not appeared in the calculation. (3) In the RIA conditions, it is concluded that the fuel failure behavior and threshold energy might not be affected by pellet shape of which size is in the range of the current LWR's fuel rods. (author)

  18. Establishing and sustaining a technical program to achieve zero fuel failures

    Deshon, J.; Whiteside, K.; Burnham, R.

    2015-01-01

    In 2006, Chief Nuclear Officers (CNOs) of electric utilities operating 103 commercial reactors in the United States (U.S.) formally endorsed an initiative to achieve failure-free fuel performance by the end of 2010. This became known as the Zero-by- Ten Fuel Failure Initiative. The endorsement manifested during a meeting at the Institute of Nuclear Power Operations (INPO) while nuclear fuel performance trends were being reviewed. Declining fuel performance generated backing for the Initiative. This paper provides a brief review of some of the drivers that caused those trends, the Initiative elements designed to address them, and a review of the technical program that helped achieve, and thus far sustain, improved fuel performance. (author) Key Words: Nuclear Fuel, Fuel, Fuel Failures, Fuel Failure Initiative, Zero by Ten Initiative

  19. Observation unit management of acute decompensated heart failure.

    Schrock, Jon W; Emerman, Charles L

    2009-01-01

    Acute decompensated heart failure (ADHF) is a common illness presenting to the emergency department (ED) that is amenable to observation unit (OU) treatment. As the number of baby boomers continues to grow and the incidence of heart failure increases, the financial implications of ADHF treatment will become more prominent. Obtaining institutional support and developing a good working relationship with cardiology colleagues is vital to creating workable ADHF protocols for whichever type of OU an institution decides to use.

  20. A model to predict failure of irradiated U–Mo dispersion fuel

    Burkes, Douglas E., E-mail: Douglas.Burkes@pnnl.gov; Senor, David J.; Casella, Andrew M.

    2016-12-15

    Highlights: • Simple model to predict failure of dispersion fuel meat designs. • Evaluated as a function of fabrication parameters and irradiation conditions. • Predictions compare well with experimental measurements of miniature fuel plates. • Interaction layer formation reduces matrix strength and increases temperature. • Si additions to the matrix appear effective only at moderate heat flux and burnup. - Abstract: Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Some of these programs are focused on development and qualification of a fuel design that consists of a uranium–molybdenum (U–Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. The current paper extends a failure model originally developed for UO{sub 2}-stainless steel dispersion fuels and uses currently available thermal–mechanical property information for the materials of interest in the currently proposed design. A number of fabrication and irradiation parameters were investigated to understand the conditions at which failure of the matrix, classified as onset of pore formation in the matrix, might occur. The results compared well with experimental observations published as part of the Reduced Enrichment for Research and Test Reactors (RERTR)-6 and -7 mini-plate experiments. Fission rate, a function of the {sup 235}U enrichment, appeared to be the most influential parameter in premature failure, mainly as a result of increased interaction layer formation and operational temperature, which coincidentally decreased the strength of the matrix and caused more rapid fission gas production and recoil into the surrounding matrix material. Addition of silicon to the matrix appeared effective at reducing the rate of

  1. Effects of cold worked and fully annealed claddings on fuel failure behaviour

    Saito, Shinzo; Hoshino, Hiroaki; Shiozawa, Shusaku; Yanagihara, Satoshi

    1979-12-01

    Described are the results of six differently heat-treated Zircaloy clad fuel rod tests in NSRR experiments. The purpose of the test is to examine the extent of simulating irradiated claddings in mechanical properties by as-cold worked ones and also the effect of fully annealing on the fuel failure bahaviour in a reactivity initiated accident (RIA) condition. As-cold worked cladding does not properly simulated the embrittlement of the irradiated one in a RIA condition, because the cladding is fully annealed before the fuel failure even in the short transient. Therefore, the fuel behaviour such as fuel failure threshold energy, failure mechanism, cladding deformation and cladding oxidation of the fully annealed cladding fuel, as well as that of the as-cold worked cladding fuel, are not much different from that of the standard stress-relieved cladding fuel. (author)

  2. Studies on fuel failure detection in Rikkyo Research Reactor

    Matsuura, T.; Hayashi, S.H.; Harasawa, S.; Tomura, K.

    1992-01-01

    Studies on fuel failure detection have been made since 1986 in Rikkyo Research Reactor. One of the methods is the monitoring of the trace concentration of fission products appearing in the air on the surface of the water tank of the reactor. The interested radionuclides here are 89 Rb and 138 Cs, which are the daughter nuclides of the FP rare gas nuclides, 89 Kr and 138 Xe, respectively and have the half lives of 15.2 min and 32.2 min respectively. They are detected on a filter paper attached on a conventional dust sampler, by sucking the air of the surface of the water for 15 ∼ 30 min during reactor operation (100 kW). In this presentation are reported the results of an attempt to increase the sensitivity of detecting these nuclides by introducing nitrogen gas bubbles into the water. The bubbling of the gas increased the sensitivity as much as several times compared with the case without bubbling. These measurements are giving us the 'background' concentration, the order of which is almost unchanged for these several years, --in 10 -6 Bq/cm 3 . The origin of these nuclides is considered to be not from the fuel but from the uranium contained as an impurity in the reactor material in the core. (author)

  3. Transient Fuel Behavior and Failure Condition in the CABRI-2 Experiments

    Sato, Ikken; Lemoine, Francette; Struwe, Dankward

    2004-01-01

    In the CABRI-2 program, 12 tests were performed under various transient conditions covering a wide range of accident scenarios using two types of preirradiated fast breeder reactor (FBR) fuel pins with different smear densities and burnups. For each fuel, a nonfailure-transient test was performed, and it provided basic information such as fuel thermal condition, fuel swelling, and gas release. From the failure tests, information on failure mode, failure time, and axial location was obtained. Based on this information, failure conditions such as fuel enthalpy and cladding temperature were evaluated. These failure conditions were compared with the CABRI-1 tests in which different fuels as well as different transient conditions were used. This comparison, together with supporting information available from existing in-pile and out-of-pile experiments, allowed an effective understanding on failure mechanisms depending on fuel and transient conditions. It is concluded that pellet-cladding mechanical interaction (PCMI) due to fuel thermal expansion and fission-gas-induced swelling is playing an important role on mechanical clad loading especially with high smear density and low fuel-heating-rate conditions. At very high heating-rate conditions, there is no sufficient time to allow significant fuel swelling, so that cavity pressurization with fuel melting becomes the likely failure mechanism. Fuel smear density and fission-gas retention have a strong impact both on PCMI and cavity pressurization. Furthermore, pin failure is strongly dependent on cladding temperature, which plays an important role in the axial failure location. With the low smear-density fuel, considerable PCMI mitigation is possible leading to a high failure threshold as well as in-pin molten-fuel relocation along the central hole. However, even with the low smear density fuel, PCMI failure could take place with an elevated cladding-temperature condition. On the other hand, in case of a sufficiently long

  4. Temperature Analysis and Failure Probability of the Fuel Element in HTR-PM

    Yang Lin; Liu Bing; Tang Chunhe

    2014-01-01

    Spherical fuel element is applied in the 200-MW High Temperature Reactor-Pebble-bed Modular (HTR-PM). Each spherical fuel element contains approximately 12,000 coated fuel particles in the inner graphite matrix with a diameter of 50mm to form the fuel zone, while the outer shell with a thickness of 5mm is a fuel-free zone made up of the same graphite material. Under high burnup irradiation, the temperature of fuel element rises and the stress will result in the damage of fuel element. The purpose of this study is to analyze the temperature of fuel element and to discuss the stress and failure probability. (author)

  5. Survey of potential light water reactor fuel rod failure mechanisms and damage limits

    Courtright, E.L.

    1979-07-01

    The findings and conclusions are presented of a survey to evaluate current information applicable to the development of fuel rod damage and failure limits for light water reactor fuel elements. The survey includes a review of past fuel failures, and identifies potential damage and failure mechanisms for both steady state operating conditions and postulated accident events. Possible relationships between the various damage and failure mechanisms are also proposed. The report identifies limiting criteria where possible, but concludes that sufficient data are not currently available in many important areas

  6. Failure analysis of edge discoloration of galvanized fuel tank

    Jitendra Mathur

    2015-10-01

    Full Text Available A peculiar type of edge discoloration defect on the surface of some galvanized fuel tank was observed, causing significant appearance problems. In the present study, the surface defect was characterized by visual inspection, optical microscopy, scanning electron microscopy and energy dispersive spectroscopic analysis to understand the source and mechanism of the defect. In the visual inspection, these peculiar surface appearances were observed in fuel tank at three distinct locations. The SEM examination exhibited two distinct regions on the surface apart from the normal galvanized surface: (1 galvannealed, (2 mixture of galvanized and galvannealed texture. The energy dispersive spectroscopic analysis of galvannealed region indicated enrichment of Zn and Al whereas in the region of galvanized majorly Zn was observed. Surface texture of galvannealed region showed majorly zeta crystals along with skin pass marks; whereas no such zeta crystals were observed in case of galvanized regions. Based on the preliminary results, the following hypothesis was made: Coil processed during galvanizing to galvannealing transition. Thickness and width changed to wider and thicker section, which resulted into lower line speed. Due to the lower Al content, lower speed and thicker section combination resulted in formation of partial GA in the coil owing to the internal heat content of the coil. This paper presents the results of the investigation.

  7. PCI/SCC failure behavior of KWU/CE fuel rods

    Kikuchi, Akira

    1983-10-01

    The Over Ramp (Studsvik Over Ramp-STOR) project is an international power ramping irradiation program for studying PCI/SCC failure behavior of PWR-fuel rods. The project had its activities for about three years (Apr., 1977 - Dec., 1980) as the cooperation works of twelve participants composing nine countries. The present report introduces the irradiation data on the KWU/CE fuel rods in the project and discusses the failure behavior of PWR-fuel rods. (author)

  8. Detection of gaseous fission products in water - a method of monitoring fuel sheathing failures

    Tunnicliffe, P. R.; Whittier, A. C.

    1959-05-15

    The gaseous activities stripped from samples of effluent coolant from the NRU fuel elements tested in the central thimble of the NRX reactor (NRU loop) and from the NRX main effluent have been investigated. The activities obtained from the NRU loop can be attributed to gaseous fission products only. Design data have been obtained for a 'Gaseous Fission Product Monitor' to be installed for use with the NRU reactor. It is expected that this monitor will have high sensitivity to activity indicative of an incipient fuel element sheath failure. No qualitative determination of the various gaseous activities obtained from the NRX effluent has been made. A strong component of 25 {+-}1 seconds half-life is not consistent with O-19. Limited information concerning sheath failures in NRX was obtained. Of six failures observed in parallel with the installed delayed neutron monitors, three of these gave pre-warnings and in each case the gaseous fission product monitor showed a substantially greater sensitivity. An experiment in which small samples of uranium, inserted into the NRX reactor, could be exposed at will to a stream of water showed the behaviour of the two types of monitors to be similar. However, a number of signals were detected only by the gaseous fission product monitor. These can be attributed to its sensitivity to relatively long lived fission products. (author)

  9. FFTF fuel failure detection and characterization by cover gas monitoring. Final report

    Miller, W.C.; Holt, F.E.

    1977-01-01

    The Fast Flux Test Facility (FFTF) will include a Fuel Failure Monitoring (FFM) System designed to detect, characterize, and locate fuel and absorber pin failures (i.e., cladding breaches) using a combination of delayed neutron detection, cover gas radioisotope monitoring, and gas tagging. During the past several years the Hanford Engineering Development Laboratory has been involved in the development, design, procurement, and installation of this integrated system. The paper describes one portion of the FFM System, the Cover Gas Monitoring System (CGMS), which has the primary function of fuel failure detection and characterization in the FFTF. By monitoring the various radioisotopes in the cover gas, the CGMS will both detect fuel and absorber pin failures and characterize those failures as to magnitude and severity

  10. Cladding failure margins for metallic fuel in the integral fast reactor

    Bauer, T.H.; Fenske, G.R.; Kramer, J.M.

    1987-01-01

    The reference fuel for Integral Fast Reactor (IFR) is a ternary U-Pu-Zr alloy with a low swelling austenitic or ferritic stainless steel cladding. It is known that low melting point eutectics may form in such metallic fuel-cladding systems which could contribute to cladding failure under accident conditions. This paper will present recent measurements of cladding eutectic penetration rates for the ternary IFR alloy and will compare these results with earlier eutectic penetration data for other fuel and cladding materials. A method for calculating failure of metallic fuel pins is developed by combining cladding deformation equations with a large strain analysis where the hoop stress is calculated using the instantaneous wall thickness as determined from correlations of the eutectic penetration-rate data. This method is applied to analyze the results of in-reactor and out-of-reactor fuel pin failure tests on uranium-fissium alloy EBR-II Mark-II driver fuel

  11. Creep test observation of viscoelastic failure of edible fats

    Vithanage, C R; Grimson, M J; Wills, P R [Department of Physics, University of Auckland, Private Bag 92019 (New Zealand); Smith, B G, E-mail: cvit002@aucklanduni.ac.nz [Food Science Programmes, Department of Chemistry, University of Auckland, Private Bag 92019 (New Zealand)

    2011-03-01

    A rheological creep test was used to investigate the viscoelastic failure of five edible fats. Butter, spreadable blend and spread were selected as edible fats because they belong to three different groups according to the Codex Alimentarius. Creep curves were analysed according to the Burger model. Results were fitted to a Weibull distribution representing the strain-dependent lifetime of putative fibres in the material. The Weibull shape and scale (lifetime) parameters were estimated for each substance. A comparison of the rheometric measurements of edible fats demonstrated a clear difference between the three different groups. Taken together the results indicate that butter has a lower threshold for mechanical failure than spreadable blend and spread. The observed behaviour of edible fats can be interpreted using a model in which there are two types of bonds between fat crystals; primary bonds that are strong and break irreversibly, and secondary bonds, which are weaker but break and reform reversibly.

  12. Failure mechanisms in high temperature gas cooled reactor fuel particles

    Soo, P.; Uneberg, G.; Sabatini, R.L.; Schweitzer, D.G.

    1979-01-01

    BISO coated UO 2 and ThO 2 particles were heated to high temperatures to determine failure mechanisms during hypothetical loss of coolant scenarios. Rapid failure begins when the oxides are reduced to liquid carbides. Several failure mechanisms are applicable, ranging from hole and crack formation in the coatings to catastrophic particle disintegration

  13. New phenomena observed during fuel assemblies testing

    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

  14. Development of evaluation method of fuel failure fraction during the High Temperature Engineering Test Reactor operation

    Sawa, Kazuhiro; Yoshimuta, Shigeharu; Tobita, Tsutomu; Sato, Masashi [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1997-05-01

    The High Temperature Engineering Test Reactor (HTTR) uses coated particles as fuel. During normal operation, short-lived noble gases are mainly released by diffusion from fuel particles with defects in their coating layers (i.e., failed particle). Since noble gases do not plate out on the inner surfaces of primary cooling system, their activities in primary coolant reflect fuel failure fraction in the core. An evaluation method was developed to predict failure fraction of coated fuel particles during normal operation of the HTTR. The method predicts core-average and hot plenum regionwise failure fractions based on the fractional releases, (R/B)s, of noble gases. The (R/B)s are calculated by fission gas concentration measurements in the primary cooling system of the HTTR. Recent fabrication data show that through-coatings failure fraction is extremely low. Then, fractional release from matrix contamination uranium, which is background for accurate evaluation of the fuel failure fraction, should be precisely predicted. This report describes an evaluation method of fuel failure fraction from measurements in the HTTR together with a fission gas release model from fuel compact containing failed particles and matrix contamination uranium. (author)

  15. Performance of water cooled nuclear power reactor fuels in India – Defects, failures and their mitigation

    Ganguly, Chaitanyamoy

    2015-01-01

    Water cooled and moderated nuclear power reactors account for more than 95% of the operating reactors in the world today. Light water reactors (LWRs) consisting of pressurized water reactor (PWR), their Russian counterpart namely VVER and boiling water reactor (BWR) will continue to dominate the nuclear power market. Pressurized heavy water reactor (PHWR), also known as CANDU, is the backbone of the nuclear power program in India. Updates on LWR and PHWR fuel performance are being periodically published by IAEA, OECD-NEA and the World Nuclear Association (WNA), highlighting fuel failure rate and the mitigation of fuel defects and failures. These reports clearly indicate that there has been significant improvement in in – pile fuel performance over the years and the present focus is to achieve zero fuel failure in high burn up and high performance fuels. The present paper summarizes the status of PHWR and LWR fuel performance in India, highlighting the manufacturing and the related quality control and inspection steps that are being followed at the PHWR fuel fabrication plant in order to achieve zero manufacturing defect which could contribute to achieving zero in – pile failure rate in operating and upcoming PHWR units in India. (author)

  16. Ambition to reach zero level failure in VVER 1000 with russian fuel

    Mečíř, V.

    2015-01-01

    The purpose of “The Zero Failure Level Project” is to bring to real operation of VVER 1000 units the dream of all utilities such as problem free and cost effective operation. This essentially turns into requirement on failure free fuel operation. At the same time the general requirements such as safety, cost effectiveness, operational flexibility, fuel cycle and fuel flexibility need to be satisfied. Several specific tasks were performed and many of them are still in process. Specific failure tree was developed in a format, which allows step by step failure tree improvement. Fuel types and its modifications, taking into account manufacturing conditions, were specified. In parallel with fuel types classification, real operational conditions were evaluated based on approximately 280 parameters by fuel assembly design features, operational procedures and practices and about 250 reactor unit parameters. As a result of this stage, groups of units with similar fuel operational conditions should be revealed and experience sharing database created. It is also recognized a need for consistent methods of operational data and data from pool side fuel assembly inspection. In the area of Foreign Material Exclusion activities closer cooperation between utility and supplier should be established including foreign material classification and improvement in root cause investigation

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

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

    2008-01-01

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

  18. HEDL empirical correlation of fuel pin top failure thresholds, status 1976

    Baars, R.E.

    1976-01-01

    The Damage Parameter (DP) empirical correlation of fuel pin cladding failure thresholds for TOP events has been revised and recorrelated to the results of twelve TREAT tests. The revised correlation, called the Failure Potential (FP) correlation, predicts failure times for the tests in the data base with an average error of 35 ms for $3/s tests and of 150 ms for 50 cents/s tests

  19. Calculating failure probabilities for TRISO-coated fuel particles using an integral formulation

    Miller, Gregory K.; Maki, John T.; Knudson, Darrell L.; Petti, David A.

    2010-01-01

    The fundamental design for a gas-cooled reactor relies on the safe behavior of the coated particle fuel. The coating layers surrounding the fuel kernels in these spherical particles, termed the TRISO coating, act as a pressure vessel that retains fission products. The quality of the fuel is reflected in the number of particle failures that occur during reactor operation, where failed particles become a source for fission products that can then diffuse through the fuel element. The failure probability for any batch of particles, which has traditionally been calculated using the Monte Carlo method, depends on statistical variations in design parameters and on variations in the strengths of coating layers among particles in the batch. An alternative approach to calculating failure probabilities is developed herein that uses direct numerical integration of a failure probability integral. Because this is a multiple integral where the statistically varying parameters become integration variables, a fast numerical integration approach is also developed. In sample cases analyzed involving multiple failure mechanisms, results from the integration methods agree closely with Monte Carlo results. Additionally, the fast integration approach, particularly, is shown to significantly improve efficiency of failure probability calculations. These integration methods have been implemented in the PARFUME fuel performance code along with the Monte Carlo method, where each serves to verify accuracy of the others.

  20. Application of different failure criteria in fuel pin modelling and consequences for overpower transients in LMFBRs

    Kuczera, B.; Royl, P.

    1975-01-01

    The CAPRI-2 code system for analysis of hypothetical core disruptive accidents in LMFBRs has recently been coupled with the transient deformation model BREDA-2. The new code system determines thermal and mechanical loads under transient conditions for both, fresh and irradiated fuel and cladding, taking into account fuel restructuring as well as effects from fission gas and fuel and clad swelling. The system has been used for analysis of mild uncontrolled overpower transients in the SNR-300 to predict failure, and to initialize and calculate subsequent fuel coolant interaction (FCI). Thirteen channels have been coupled by point kinetics for the whole core analysis. Three different failure mechanisms and their influence on accident sequence have been investigated: clad melt-through; clad burst caused by internal pressure build-up; clad straining due to differential thermal expansion between fuel and clad cylinders. The results of these analyses show that each failure mechanism will lead to rather different failure and accident sequences. There is still a lack of experimental data from which failure thresholds can be derived. To get better predictions from the applied models an improved understanding of fission release and its relation to fuel porosity also some better experimental data on fluence and temperature dependent rupture strains of the cladding material should be available

  1. Fission product concentration evolution in sodium pool following a fuel subassembly failure in an LMFBR

    Natesan, K.; Velusamy, K.; Selvaraj, P.; Kasinathan, N.; Chellapandi, P.; Chetal, S.; Bhoje, S.

    2003-01-01

    During a fuel element failure in a liquid metal cooled fast breeder reactor, the fission products originating from the failed pins mix into the sodium pool. Delayed Neutron Detectors (DND) are provided in the sodium pool to detect such failures by way of detection of delayed neutrons emitted by the fission products. The transient evolution of fission product concentration is governed by the sodium flow distribution in the pool. Transient hydraulic analysis has been carried out using the CFD code PHOENICS to estimate fission product concentration evolution in hot pool. k- ε turbulence model and zero laminar diffusivity for the fission product concentration have been considered in the analysis. Times at which the failures of various fuel subassemblies (SA) are detected by the DND are obtained. It has been found that in order to effectively detect the failure of every fuel SA, a minimum of 8 DND in hot pool are essential

  2. Fission product release modelling for application of fuel-failure monitoring and detection - An overview

    Lewis, B.J., E-mail: lewibre@gmail.com [Department of Chemistry and Chemical Engineering, Royal Military College of Canada, Kingston, Ontario, K7K 7B4 (Canada); Chan, P.K.; El-Jaby, A. [Department of Chemistry and Chemical Engineering, Royal Military College of Canada, Kingston, Ontario, K7K 7B4 (Canada); Iglesias, F.C.; Fitchett, A. [Candesco Division of Kinectrics Inc., 26 Wellington Street East, 3rd Floor, Toronto, Ontario M5E 1S2 (Canada)

    2017-06-15

    A review of fission product release theory is presented in support of fuel-failure monitoring analysis for the characterization and location of defective fuel. This work is used to describe: (i) the development of the steady-state Visual-DETECT code for coolant activity analysis to characterize failures in the core and the amount of tramp uranium; (ii) a generalization of this model in the STAR code for prediction of the time-dependent release of iodine and noble gas fission products to the coolant during reactor start-up, steady-state, shutdown, and bundle-shifting manoeuvres; (iii) an extension of the model to account for the release of fission products that are delayed-neutron precursors for assessment of fuel-failure location; and (iv) a simplification of the steady-state model to assess the methodology proposed by WANO for a fuel reliability indicator for water-cooled reactors.

  3. Bio-fuels barometer - EurObserv'ER - July 2016

    2016-07-01

    The European bio-fuel market is now regulated by the directive, known as ILUC, whose wording focuses on the environmental impact of first generation bio-fuel development. This long-awaited clarification has arrived against the backdrop of falling oil prices and shrinking European Union bio-fuel consumption, which should drop by 1.7% between 2014 and 2015, according to EurObserv'ER

  4. Assessment of the impact of fueling machine failure on the safety of the CANDU-PHWR

    Al-Kusayer, T.A.

    1982-01-01

    A survey of possible LOCA (Loss-of-Coolant Accident) initiating events that might take place for CANDU-PHWRs (Canadian Deuterium Uranium-Pressurized Heavy Water Reactors) has been conducted covering both direct and indirect initiators. Among the 22 initiating events that were surveyed in this study, four direct initiators have been selected and analyzed briefly. Those selected were a pump suction piping break, an isolation valve piping break, a bleed valve failure, and a fueling machine interface failure. These were selected as examples of failures that could take place in the inlet side, outlet side, or PHTS (Primary Heat Transport System) interfaces. The Pickering NGS (Unit-A) was used for this case study. Double failure (failure of the protective devices to operate when the process equipment fault occurs) and a triple failure (failure of the protective devices and the ECCS as well as the process equipment) were found to be highly improbable

  5. Investigation of a Shock Absorber for Safeguard of Fuel Assemblies Failure

    Karalevicius, Renatas; Dundulis, Gintautas; Rimkevicius, Sigitas; Uspuras, Eugenijus

    2006-01-01

    The Ignalina NPP has two reactors. The Unit 1 was shut down, therefore the special equipment was designed for transportation of the fuel from Unit 1 to Unit 2. The fuel-loaded basket can drop during transportation. The special shock absorber was designed in order to avoid failure of fuel assemblies during transportation. In case of drop of fuel loaded basket, the failure of fuel assemblies can occur. This shock absorber was studied by scaled experiments at Lithuanian Energy Institute. Static and dynamic investigations of shock absorber are presented in this paper, including dependency of axial force versus axial compression. The finite element codes BRIGADE/Plus and ABAQUS/Explicit were used for analysis. Static simulation was used to optimize the dimensions of shock absorber. Dynamic analysis shows that shock absorber is capable to withstand the dynamic load for successful force suppression function in case of an accident. (authors)

  6. A study of fuel failure behavior in high burnup HTGR fuel. Analysis by STRESS3 and STAPLE codes

    Martin, David G.; Sawa, Kazuhiro; Ueta, Shouhei; Sumita, Junya

    2001-05-01

    In current high temperature gas-cooled reactors (HTGRs), Tri-isotropic coated fuel particles are employed as fuel. In safety design of the HTGR fuels, it is important to retain fission products within particles so that their release to primary coolant does not exceed an acceptable level. From this point of view, the basic design criteria for the fuel are to minimize the failure fraction of as-fabricated fuel coating layers and to prevent significant additional fuel failures during operation. This report attempts to model fuel behavior in irradiation tests using the U.K. codes STRESS3 and STAPLE. Test results in 91F-1A and HRB-22 capsules irradiation tests, which were carried out at the Japan Materials Testing Reactor of JAERI and at the High Flux Isotope Reactor of Oak Ridge National Laboratory, respectively, were employed in the calculation. The maximum burnup and fast neutron fluence were about 10%FIMA and 3 x 10 25 m -2 , respectively. The fuel for the irradiation tests was called high burnup fuel, whose target burnup and fast neutron fluence were higher than those of the first-loading fuel of the High Temperature Engineering Test Reactor. The calculation results demonstrated that if only mean fracture stress values of PyC and SiC are used in the calculation it is not possible to predict any particle failures, by which is meant when all three load bearing layers have failed. By contrast, when statistical variations in the fracture stresses and particle specifications are taken into account, as is done in the STAPLE code, failures can be predicted. In the HRB-22 irradiation test, it was concluded that the first two particles which had failed were defective in some way, but that the third and fourth failures can be accounted for by the pressure vessel model. In the 91F-1A irradiation test, the result showed that 1 or 2 particles had failed towards the end of irradiation in the upper capsule and no particles failed in the lower capsule. (author)

  7. Acceleration Test Method for Failure Prediction of the End Cap Contact Region of Sodium Cooled Fast Reactor Fuel Rod

    Kim, Hyung-Kyu; Lee, Young-Ho; Lee, Hyun-Seung; Lee, Kang-Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-05-15

    This paper reports the results of an acceleration test to predict the contact-induced failure that could occur at the cylinder-to-hole joint for the fuel rod of a sodium-cooled fast reactor (SFR). To incorporate the fuel life of the SFR currently under development at KAERI (around 35,000 h), the acceleration test method of reliability engineering was adopted in this work. A finite element method was used to evaluate the flow-induced vibration frequency and amplitude for the test parameter values. Five specimens were tested. The failure criterion during the life of the SFR fuel was applied. The S-N curve of the HT-9, the material of concern, was used to obtain the acceleration factor. As a result, a test time of 16.5 h was obtained for each specimen. It was concluded that the B{sub 0.004} life would be guaranteed for the SFR fuel rods with 99% confidence if no failure was observed at any of the contact surfaces of the five specimens.

  8. The failure mechanisms of HTR coated particle fuel and computer code

    Yang Lin; Liu Bing; Shao Youlin; Liang Tongxiang; Tang Chunhe

    2010-01-01

    The basic constituent unit of fuel element in HTR is ceramic coated particle fuel. And the performance of coated particle fuel determines the safety of HTR. In addition to the traditional detection of radiation experiments, establishing computer code is of great significance to the research. This paper mainly introduces the structure and the failure mechanism of TRISO-coated particle fuel, as well as a few basic assumptions,principles and characteristics of some existed main overseas codes. Meanwhile, this paper has proposed direction of future research by comparing the advantages and disadvantages of several computer codes. (authors)

  9. Consequences assessment for fuel channel failure with consequential moderator drain

    Wahba, N.N.; Bayoumi, M.H.

    2002-01-01

    This paper documents the consequences of spontaneous pressure tube/consequential calandria tube rupture followed by the ejection of end fittings (as a result of guillotine failure of pressure tube) leading to the drain of the moderator. The event is postulated to occur in conjunction with an independent failure of Emergency Coolant Injection System (ECIS). The results of the detailed consequence assessments are used to propose a course of action to mitigate the consequences of such an event. A methodology based on a lumped-parameter model was developed to assess the consequences of the postulated event. (author)

  10. Observations on analysis, testing and failure of prestressed concrete containments

    Murray, D.W.

    1984-01-01

    The paper reviews the mechanics which indicate that a bursting failure with large energy release is the failure mechanism to be expected from ductile lined containment structures pressurized to failure. It reviews a study which shows that, because of leakage, this is not the case for unlined prestressed containments. It argues that current practice, since it does not specifically address the bursting failure problem for lined prestressed containments, is inadequate to ensure that this type of failure could not occur. It concludes that, in view of the inadequacy of the current state-of-the-art to predict leakage from lined structures, the logical remedy is to eliminate all possibility of bursting failure by making provision for venting of containments. (orig.)

  11. Fuel-element failures in Hanford single-pass reactors 1944--1971

    Gydesen, S.P.

    1993-07-01

    The primary objective of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate the radiation dose that individuals could have received as a result of emissions since 1944 from the US Department of Energy`s (DOE) Hanford Site near Richland, Washington. To estimate the doses, the staff of the Source Terms Task use operating information from historical documents to approximate the radioactive emissions. One source of radioactive emissions to the Columbia River came from leaks in the aluminum cladding of the uranium metal fuel elements in single-pass reactors. The purpose of this letter report is to provide photocopies of the documents that recorded these failures. The data from these documents will be used by the Source Terms Task to determine the contribution of single-pass reactor fuel-element failures to the radioactivity of the reactor effluent from 1944 through 1971. Each referenced fuel-element failure occurring in the Hanford single-pass reactors is addressed. The first recorded failure was in 1948, the last in 1970. No records of fuel-element failures were found in documents prior to 1948. Data on the approximately 2000 failures which occurred during the 28 years (1944--1971) of Hanford single-pass reactor operations are provided in this report.

  12. Influence of plutonium contents in MOX fuel on destructive forces at fuel failure in the NSRR experiment

    Nakamura, Jinichi; Sugiyama, Tomoyuki; Nakamura, Takehiko; Kanazawa, Toru; Sasajima, Hideo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    In order to confirm safety margins of the Mixed Oxide (MOX) fuel use in LWRs, pulse irradiation tests are planned in the Nuclear Safety Research Reactor (NSRR) with the MOX fuel with plutonium content up to 12.8%. Impacts of the higher plutonium contents on safety of the reactivity-initiated-accident (RIA) tests are examined in terms of generation of destructive forces to threat the integrity of test capsules. Pressure pulses would be generated at fuel rod failure by releases of high pressure gases. The strength of the pressure pulses, therefore, depends on rod internal - external pressure difference, which is independent to plutonium content of the fuel. The other destructive forces, water hammer, would be generated by thermal interaction between fuel fragments and coolant water. Heat flux from the fragments to the water was calculated taking account of changes in thermal properties of MOX fuels at higher plutonium contents. The results showed that the heat transfer from the MOX fuel would be slightly smaller than that from UO{sub 2} fuel fragments at similar size in a short period to cause the water hammer. Therefore, the destructive forces were not expected to increase in the new tests with higher plutonium content MOX fuels. (author)

  13. Can failure carefully observed become a springboard to success?

    Adrian, Manuella

    2012-01-01

    Since its inception, the addictions field has had a history of failure: failures in conceptualizations, in treatment, in interventions, in policies, in process as well as outcome assessment. Certain actions and activities have had a less than stellar effect which may lead to feelings of personal failure among practitioners, the tagging of processes and programs as being failures when they are not so, as well as an identification of the person being intervened with, by self and others, as being a failure or loser. This paper discusses how to define success and failure and the need to identify both the short(er) and long(er) term, as well as temporary and permanent effects, including the implications of using binary (success or failure; success and failure) and nonbinary (and in addition) categories of assessment. The need to clarify expectations and to establish goals and measurable effects are noted. Being open to accepting results which may be personally disappointing, initially, but which offer opportunities for needed changes may lead to new developments in the field and the establishment of better interventions.

  14. Comparison of a fuel sheath failure model with published experimental data

    Varty, R.L.; Rosinger, H.E.

    1982-01-01

    A fuel sheath failure model has been compared with the published results of experiments in which a Zircaloy-4 fuel sheath was subjected to a temperature ramp and a differential pressure until failure occurred. The model assumes that the deformation of the sheath is controlled by steady-state creep and that there is a relationship between tangential stress and temperature at the instant of failure. The sheath failure model predictions agree reasonably well with the experimental data. The burst temperature is slightly overpredicted by the model. The burst strain is overpredicted for small experimental burst strains but is underpredicted otherwise. The reasons for these trends are discussed and the extremely wide variation in burst strain reported in the literature is explained using the model

  15. A statistical analysis of pellet-clad interaction failures in water reactor fuel

    McDonald, S.G.; Fardo, R.D.; Sipush, P.J.; Kaiser, R.S.

    1981-01-01

    The primary objective of the statistical analysis was to develop a mathematical function that would predict PCI fuel rod failures as a function of the imposed operating conditions. Linear discriminant analysis of data from both test and commercial reactors was performed. The initial data base used encompassed 713 data points (117 failures and 596 non-failures) representing a wide variety of water cooled reactor fuel (PWR, BWR, CANDU, and SGHWR). When applied on a best-estimate basis, the resulting function simultaneously predicts approximately 80 percent of both the failure and non-failure data correctly. One of the most significant predictions of the analysis is that relatively large changes in power can be tolerated when the pre-ramp irradiation power is low, but that only small changes in power can be tolerated when the pre-ramp irradiation power is high. However, it is also predicted that fuel rods irradiated at low power will fail at lower final powers than those irradiated at high powers. Other results of the analysis are that fuel rods with high clad operating temperatures can withstand larger power increases that fuel rods with low clad operating temperatures, and that burnup has only a minimal effect on PCI performance after levels of approximately 10000 MWD/MTU have been exceeded. These trends in PCI performance and the operating parameters selected are believed to be consistent with mechanistic considerations. Published PCI data indicate that BWR fuel usually operates at higher local powers and changes in power, lower clad temperatures, and higher local ramp rates than PWR fuel

  16. Measurement of fission product release during LWR fuel failure

    Osetek, D.J.; King, J.J.

    1979-01-01

    The PBF is a specialized test reactor consisting of an annular core and a central test space 21 cm in diameter and 91 cm high. A test loop circulates coolant through the central experimental section at typical power reactor conditions. Light-water-reactor-type fuel rods are exposed to power bursts simulating reactivity insertion transients, and to power-cooling-mismatch conditions during which the rods are allowed to operate in film boiling. Fission product concentrations in the test loop coolant are continuously monitored during these transients by a Ge(Li) detector based gamma spectrometer. Automatic batch processing of pulse height spectra results in a list of radionuclide concentrations present in the loop coolant as a function of time during the test. Fission product behavior is then correlated to test parameters and posttest examination of the fuel rods. Data are presented from Test PCM-1

  17. The German hydrogen and fuel cell community. Successes and failures

    Canzler, Weert; Marz, Lutz [Wissenschaftszentrum Berlin fuer Sozialforschung gGmbH (WZB), Berlin (Germany); Galich, Ante [Luxembourg Univ. (Luxembourg). Faculty of Languages and Literature, Humanities, Arts and Education

    2013-11-01

    Recently, the German Federal Government made the consequential decision to change its energy program. This not only as a result of the decision to shut down the existing nuclear power plants within the next few years, but also due to vital challenges like climate change and security of energy supply. The shift in the energy-technology paradigm from fossil fuel technologies to regenerative energies constitutes a major technical process but also new economic and social constellations. This paper focuses on hydrogen and fuel cell technologies in Germany. The institutional set up in this field is analysed and the new organizational actors are identified who have actively lobbied towards a political consensus. However, the experts in this field could not attain the required leadership in the public discourse on these technologies. It seems that an attractive guiding vision of a post-fossil energy future and a broad acceptance in daily use would have been major prerequisites for such leadership. (orig.)

  18. Failure Analysis of a Helicopter External Fuel-Tank Pylon

    Newman, John A.; Piascik, Robert S.; Lindenberg, Richard A.

    2002-01-01

    An eight-inch-long (0.2 m) crack was found in an external fuel-tank pylon of a U.S. Coast Guard HH-60 helicopter. The damaged pylon was removed from service and destructively examined at NASA Langley Research Center (LaRC) to determine the cause of the crack. Results of the analysis revealed that crack initiation occurred at corrosion pits in a fastener hole and crack propagation was a result of cyclic loading.

  19. AGR fuel pin pellet-clad interaction failure limits and activity release fractions

    Hughes, H.; Hargreaves, R.

    1985-01-01

    The limiting conditions beyond which pellet-clad interaction can flail AGR fuel are described. They have been determined by many experiments involving post-irradiation examination and testing, loop experiments and cycling and up-rating of both individual fuel stringers and the whole WAGR core. The mechanisms causing this interaction are well understood and are quantitatively expressed in computer codes. Strain concentration effects over fuel cracks determine power cycling endurance while additional strain concentrations at clad ridges and from cross pin temperature gradients contribute to up-rating failures. An equation summarising tube burst test data so as to determine the ductility available at any transient is given. The hollow fuel and more ductile clad of the Civil AGR fuel pins leads to a much improved performance over the original fuel design. The Civil AGRs operate well within these limiting conditions and substantial increases beyond the design burn-up are confidently expected. The activity release on pin failure and its development during continued operation of failed fuel have also been investigated. A retention of radioiodine and caesium of 90-99% compared to the noble gases has been demonstrated. Measured fission gas releases into the free volume of Civil AGR fuel pins have been very low (< 0.1%)

  20. Cladding failure margins for metallic fuel in the integral fast reactor

    Bauer, T.H.; Fenske, G.R.; Kramer, J.M.

    1987-01-01

    The Integral Fast Reactor (IFR) concept being developed at Argonne National Laboratory has prompted a renewed interest in uranium-based metal alloys as a fuel for sodium-cooled fast reactors. In this paper we will present recent measurements of cladding eutectic penetration rates for the ternary IFR alloy and will compare these results with earlier eutectic penetration data for other fuel and cladding materials. A method for calculating failure of metallic fuel pins is developed by combining cladding deformation equations with a large strain analysis where the hoop stress is calculated using the instantaneous wall thickness as determined from correlations of the eutectic penetration-rate data. This method is applied to analyze the results of in-reactor and out-of-reactor fuel pin failure tests on uranium-fissium alloy EBR-II Mark-II driver fuel. In the final section of this paper we extend the calculations to consider the failure of IFR ternary fuel under reactor accident conditions. (orig./GL)

  1. Instrumentation Report No. 2: identification, evaluation, and remedial actions related to transducer failures at the spent fuel test-climax

    Patrick, W.C.; Carlson, R.C.; Rector, N.L.

    1981-11-30

    The Spent Fuel Test-Climax (SFT-C) is a test of the feasibility of safe and reliable short-term storage and retrieval of spent fuel from commercial nuclear reactors. In support of operational and technical goals of the test, about 850 channels of instrumentation have been installed at the SFT-C. Failure of several near-field instruments began less than six months after emplacement of 11 canisters of spent fuel and activation of six thermally similar simulators. The failed units were linear potentiometers (used to make displacement measurements) and vibrating wire stressmeters (used to make change-in-stress measurements). This report discusses the observed problems and remedial actions taken to date.

  2. Instrumentation Report No. 2: identification, evaluation, and remedial actions related to transducer failures at the spent fuel test-climax

    Patrick, W.C.; Carlson, R.C.; Rector, N.L.

    1981-01-01

    The Spent Fuel Test-Climax (SFT-C) is a test of the feasibility of safe and reliable short-term storage and retrieval of spent fuel from commercial nuclear reactors. In support of operational and technical goals of the test, about 850 channels of instrumentation have been installed at the SFT-C. Failure of several near-field instruments began less than six months after emplacement of 11 canisters of spent fuel and activation of six thermally similar simulators. The failed units were linear potentiometers (used to make displacement measurements) and vibrating wire stressmeters (used to make change-in-stress measurements). This report discusses the observed problems and remedial actions taken to date

  3. Water sampling device for fuel rod failure monitoring

    Oogami, Hideaki; Echigoya, Hironori; Matsuoka, Tesshi.

    1991-01-01

    The device of the present invention accurately samples coolants in a channel box as sampling water even if the upper end of the channel box of a fuel assembly is positioned at the same height or lower than the upper end of an upper lattice plate. An existent device comprises an outer cap, an inner cap, an air supply pipe and a water sampling tube. In addition, the device of the present invention comprises a sealing material disposed at the end of the outer cap for keeping liquid sealing with the upper lattice plate and a water level monitoring pipe extended to lower than the inner cap passing through the liquid sealing of the outer cap for sucking the atmosphere in the outer cap. Pressurized air is sent through the air supply pipe, to lower the water level of the coolants in the outer cap and the water level monitoring pipe sucks the pressurized air, by which the inside and the outside of the channel box are partitioned. Subsequently, if the sample water is sampled by a sampling tube, sampling water which enables accurate evaluation for radioactivity concentration in the fuel assembly can be obtained. (I.S.)

  4. Automatic failure identification of the nuclear power plant pellet fuel

    Oliveira, Adriano Fortunato de

    2010-01-01

    This paper proposed the development of an automatic technique for evaluating defects to help in the stage of fabrication of fuel elements. Was produced an intelligent image analysis for automatic recognition of defects in uranium pellets. Therefore, an Artificial Neural Network (ANN) was trained using segments of histograms of pellets, containing examples of both normal (no fault) and of defectives pellets (with major defects normally found). The images of the pellets were segmented into 11 shares. Histograms were made of these segments and trained the ANN. Besides automating the process, the system was able to obtain this classification accuracy of 98.33%. Although this percentage represents a significant advance ever in the quality control process, the use of more advanced techniques of photography and lighting will reduce it to insignificant levels with low cost. Technologically, the method developed, should it ever be implemented, will add substantial value in terms of process quality control and production outages in relation to domestic manufacturing of nuclear fuel. (author)

  5. Power distribution gradients in WWER type cores and fuel failure root causes

    Mikuš, Ján M., E-mail: JanMikus.nrc@hotmail.com

    2014-02-15

    Highlights: • Power (fission rate) distribution gradients can represent fuel failure root causes. • Positions with above gradients were investigated in WWER type cores on reactor LR-0. • Above gradients were evaluated near core heterogeneities and construction materials. • Results can be used for code validation and fuel failure occurrence investigation. - Abstract: Neutron flux non-uniformity and gradients of neutron current resulting in corresponding power (fission rate) distribution changes can represent root causes of the fuel failure. Such situation can be expected in vicinity of some core heterogeneities and construction materials. Since needed data cannot be obtained from nuclear power plant (NPP), results of some benchmark type experiments performed on light water, zero-power research reactor LR-0 were used for investigation of the above phenomenon. Attention was focused on determination of the spatial power distribution changes in fuel assemblies (FAs): Containing fuel rods (FRs) with Gd burnable absorber in WWER-440 and WWER-1000 type cores, Neighboring the core blanket and dummy steel assembly simulators on the periphery of the WWER-440 standard and low leakage type cores, resp., Neighboring baffle in WWER-1000 type cores, and Neighboring control rod (CR) in WWER-440 type cores, namely (a) power peak in axial power distribution in periphery FRs of the adjacent FAs near the area between CR fuel part and butt joint to the CR absorbing part and (b) decrease in radial power distribution in FRs near CR absorbing part. An overview of relevant experimental results from reactor LR-0 and some information concerning leaking FAs on NPP Temelín are presented. Obtained data can be used for code validation and subsequently for the fuel failure occurrence investigation.

  6. Material failures observed in Doublet III neutral beamlines

    Bailey, E.W.; Colleraine, A.; Doll, D.; Grunloh, H.; Kim, J.; Langhorn, A.; Thurgood, B.

    1983-12-01

    The Doublet III neutral beam injectors consist of three separable spools two meters in diameter by four meters long overall when assembled. Contained within these spools are the neutralizers, ion dumps, deflecting magnet, calorimeter dumps, cryogenic panels and beam scraping collimators 3,7. To date three beamlines are in operation on Doublet III, and the beams have accumulated operating time of approximately 32 months, with the oldest having been in operation for 18 months. During this time operation of DIII with the neutral beam sources has demonstrated the following: 7.8 MW injected neutrals from three beamlines (6 sources), high β (4.5%), and non-circular plasma shape. The sources have also exhibited a very reliable injected shot history 4, 5, 6, 8. Material failures encountered during the operation of DIII N.B. injectors and the solutions to these failures are described. Failures include cracking of the neutralizer exit collimator due to heating cycles, failure of cyropanel support rods due to cooling cycles, failure of the sliding drive of the moveable calorimeter due to friction

  7. Material failures observed in the Doublet III neutral beamlines

    Bailey, E.W.; Colleraine, A.; Doll, D.; Grunloh, H.; Kim, J.; Langhorn, A.; Thurgood, B.

    1983-01-01

    The Doublet III neutral beam injectors consist of three separable spools two meters in diameter by four meters long overall when assembled. Contained within these spools are the neutralizers, ion dumps, deflecting magnet, calorimeter dumps, cryogenic panels and beam scraping collimators. To date three beamlines are in operation on Doublet III, and the beams have accumulated operating time of approximately 32 months, with the oldest having been in operation for 18 months. During this time operation of DIII with the neutral beam sources has demonstrated the following: 7.8 MW injected neutrals from three beamlines (6 sources), high β (4.5%), and non-circular plasma shape. The sources have also exibited a very reliable injected shot history. Material failures encountered during the operation of DIII N.B. injectors and the solutions to these failures are described. Failures include cracking of the neutralizer exit collimator due to heating cycles, failure of cyropanel support rods due to cooling cycles, failure of the sliding drive of the moveable calorimeter due to friction

  8. A methodology for the evaluation of fuel rod failures under transportation accidents

    Rashid, J.Y.R.; Machiels, A.J.

    2004-01-01

    Recent studies on long-term behavior of high-burnup spent fuel have shown that under normal conditions of stor-age, challenges to cladding integrity from various postulated damage mechanisms, such as delayed hydride crack-ing, stress-corrosion cracking and long-term creep, would not lead to any significant safety concerns during dry storage, and regulatory rules have subsequently been established to ensure that a compatible level of safety is maintained. However, similar safety assurances for spent fuel transportation have not yet been developed, and further studies are currently being conducted to evaluate the conditions under which transportation-related safety issues can be resolved. One of the issues presently under evaluation is the ability and the extent of the fuel as-semblies to maintain non-reconfigured geometry during transportation accidents. This evaluation may determine whether, or not, the shielding, confinement, and criticality safety evaluations can be performed assuming initial fuel assembly geometries. The degree to which spent fuel re-configuration could occur during a transportation accident would depend to a large degree on the number of fuel rod failures and the type and geometry of the failure modes. Such information can only be developed analytically, as there is no direct experimental data that can provide guidance on the level of damage that can be expected. To this end, the paper focuses on the development of a modeling and analysis methodology that deals with this general problem on a generic basis. First consideration is given to defining acci-dent loading that is equivalent to the bounding, although analytically intractable, hypothetical transportation acci-dent of a 9-meter drop onto essentially unyielding surface, which is effectively a condition for impact-limiters de-sign. Second, an analytically robust material constitutive model, an essential element in a successful structural analysis, is required. A material behavior model

  9. Assessment of Core Failure Limits for Light Water Reactor Fuel under Reactivity Initiated Accidents

    Jernkvist, Lars Olof; Massih, Ali R.

    2004-12-01

    Core failure limits for high-burnup light water reactor UO 2 fuel rods, subjected to postulated reactivity initiated accidents (RIAs), are here assessed by use of best-estimate computational methods. The considered RIAs are the hot zero power rod ejection accident (HZP REA) in pressurized water reactors and the cold zero power control rod drop accident (CZP CRDA) in boiling water reactors. Burnup dependent core failure limits for these events are established by calculating the fuel radial average enthalpy connected with incipient fuel pellet melting for fuel burnups in the range of 30 to 70 MWd/kgU. The postulated HZP REA and CZP CRDA result in lower enthalpies for pellet melting than RIAs that take place at rated power. Consequently, the enthalpy thresholds presented here are lower bounds to RIAs at rated power. The calculations are performed with best-estimate models, which are applied in the FRAPCON-3.2 and SCANAIR-3.2 computer codes. Based on the results of three-dimensional core kinetics analyses, the considered power transients are simulated by a Gaussian pulse shape, with a fixed width of either 25 ms (REA) or 45 ms (CRDA). Notwithstanding the differences in postulated accident scenarios between the REA and the CRDA, the calculated core failure limits for these two events are similar. The calculated enthalpy thresholds for fuel pellet melting decrease gradually with fuel burnup, from approximately 960 J/gUO 2 at 30 MWd/kgU to 810 J/gUO 2 at 70 MWd/kgU. The decline is due to depression of the UO 2 melting temperature with increasing burnup, in combination with burnup related changes to the radial power distribution within the fuel pellets. The presented fuel enthalpy thresholds for incipient UO 2 melting provide best-estimate core failure limits for low- and intermediate-burnup fuel. However, pulse reactor tests on high-burnup fuel rods indicate that the accumulation of gaseous fission products within the pellets may lead to fuel dispersal into the coolant at

  10. Plan of studies on fuel failure detection in Rikkyo Research Reactor

    Matsuura, T.; Nagahara, T.; Hattori, M.; Kawaguchi, K.

    1987-01-01

    Studies on fuel failure detection in Rikkyo Research Reactor have recently been begun in the following four approaches. (1) Accumulation of the data on the concentration of the short-lived radioactivity originating from FP rare gases contained in the air on the water surface of the reactor tank. (2) Accumulation of the data on the concentration of FP (especially 131 I) in the water of the reactor tank. (3) Design and preparation of a ''sniffer'' by which the location of the failed fuel element can be detected, when some anomaly is found in the above two routine measurements. (4) Design and preparation of a vessel containing a fuel element, which can be useful both for ''sipping'' inspection of the fuel element and for storage of the damaged fuel element. In this paper, an outline of the above approaches and the results of some preliminary experiments are reported. (author)

  11. Development of Uranium-Carrying Ball method for calibration of fuel element failure detecting systems

    Liu Yupu; Bao Wanping; Lu Cungang

    1988-01-01

    A Uranium-Carrying Ball method used for the determination of sensitivity, stability of the fuel element failure detecting systems is developed. A special facility for transporting the ball can be carried out by the flow of the cooling water, so that the failure signal can be simulated. Five different types of the Uranium-Carrying Ball have been developed. Type-I to Type-IV may provide failure signal in terms of uranium quantity or exposure area of uranium. Type-V can be used to simulate micro-flaw and examine the detectability of various detective methods for this kind of defect, at the same time it is difficult for the delayed neutron detector to detect micro-flaw. The results of long-time irradiation and washing test show that the working life of the balls is satisfactory. Using the experimentel facility with the balls, detailed study of the capability of various fuel failure detecting systems have been conducted successfully. The operation is easy and safe, the accuracy of this method is higher than that of other methods, the nuclear fuel consumption as well as the radioactive contamination is low. At present, the research on the failure mechanism is being conducted by means of this method

  12. Expert system for identification of simultaneous and sequential reactor fuel failures with gas tagging

    Gross, Kenny C.

    1994-01-01

    Failure of a fuel element in a nuclear reactor core is determined by a gas tagging failure detection system and method. Failures are catalogued and characterized after the event so that samples of the reactor's cover gas are taken at regular intervals and analyzed by mass spectroscopy. Employing a first set of systematic heuristic rules which are applied in a transformed node space allows the number of node combinations which must be processed within a barycentric algorithm to be substantially reduced. A second set of heuristic rules treats the tag nodes of the most recent one or two leakers as "background" gases, further reducing the number of trial node combinations. Lastly, a "fuzzy" set theory formalism minimizes experimental uncertainties in the identification of the most likely volumes of tag gases. This approach allows for the identification of virtually any number of sequential leaks and up to five simultaneous gas leaks from fuel elements.

  13. Expert system for identification of simultaneous and sequential reactor fuel failures with gas tagging

    Gross, K.C.

    1994-01-01

    Failure of a fuel element in a nuclear reactor core is determined by a gas tagging failure detection system and method. Failures are catalogued and characterized after the event so that samples of the reactor's cover gas are taken at regular intervals and analyzed by mass spectroscopy. Employing a first set of systematic heuristic rules which are applied in a transformed node space allows the number of node combinations which must be processed within a barycentric algorithm to be substantially reduced. A second set of heuristic rules treats the tag nodes of the most recent one or two leakers as ''background'' gases, further reducing the number of trial node combinations. Lastly, a ''fuzzy'' set theory formalism minimizes experimental uncertainties in the identification of the most likely volumes of tag gases. This approach allows for the identification of virtually any number of sequential leaks and up to five simultaneous gas leaks from fuel elements. 14 figs

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

    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.

  15. Development of a Computer Code for the Estimation of Fuel Rod Failure

    Rhee, I.H.; Ahn, H.J. [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    1997-12-31

    Much research has already been performed to obtain the information on the degree of failed fuel rods from the primary coolant activities of operating PWRs in the last few decades. The computer codes that are currently in use for domestic nuclear power plants, such as CADE code and ABB-CE codes developed by Westinghouse and ABB-CE, respectively, still give significant overall errors in estimating the failed fuel rods. In addition, with the CADE code, it is difficult to predict the degree of fuel rod failures during the transient period of nuclear reactor operation, where as the ABB-CE codes are relatively more difficult to use for end-users. In particular, the rapid progresses made recently in the area of the computer hardware and software systems that their computer programs be more versatile and user-friendly. While the MS windows system that is centered on the graphic user interface and multitasking is now in widespread use, the computer codes currently employed at the nuclear power plants, such as CADE and ABB-CE codes, can only be run on the DOS system. Moreover, it is desirable to have a computer code for the fuel rod failure estimation that can directly use the radioactivity data obtained from the on-line monitoring system of the primary coolant activity. The main purpose of this study is, therefore, to develop a Windows computer code that can predict the location, the number of failed fuel rods,and the degree of failures using the radioactivity data obtained from the primary coolant activity for PWRs. Another objective is to combine this computer code with the on-line monitoring system of the primary coolant radioactivity at Kori 3 and 4 operating nuclear power plants and enable their combined use for on-line evaluation of the number and degree of fuel rod failures. (author). 49 refs., 85 figs., 30 tabs.

  16. Comparison of US/FRG accident condition models for HTGR fuel failure and radionuclide release

    Verfondern, K.

    1991-03-01

    The objective was to compare calculation models used in safety analyses in the US and FRG which describe fission product release behavior from TRISO coated fuel particles under core heatup accident conditions. The frist step performed is the qualitative comparison of both sides' fuel failure and release models in order to identify differences and similarities in modeling assumptions and inputs. Assumptions of possible particle failure mechanisms under accident conditions (SiC degradation, pressure vessel) are principally the same on both sides though they are used in different modeling approaches. The characterization of a standard (= intact) coated particle to be of non-releasing (GA) or possibly releasing (KFA/ISF) type is one of the major qualitative differences. Similar models are used regarding radionuclide release from exposed particle kernels. In a second step, a quantitative comparison of the calculation models was made by assessing a benchmark problem predicting particle failure and radionuclide release under MHTGR conduction cooldown accident conditions. Calculations with each side's reference method have come to almost the same failure fractions after 250 hours for the core region with maximum core heatup temperature despite the different modeling approaches of SORS and PANAMA-I. The comparison of the results of particle failure obtained with the Integrated Failure and Release Model for Standard Particles and its revision provides a 'verification' of these models in this sense that the codes (SORS and PANAMA-II, and -III, respectively) which were independently developed lead to very good agreement in the predictions. (orig./HP) [de

  17. Potential for fuel melting and cladding thermal failure during a PCM event in LWRs

    El-Genk, M.S.; Croucher, D.W.

    1979-01-01

    The primary concern in nuclear reactor safety is to ensure that no conceivable accident, whether initiated by a failure of the reactor system or by incorrect operation, will lead to a dangerous release of radiation to the environment. A number of hypothesized off-normal power or cooling conditions, generally termed as power-cooling-mismatch (PCM) accidents, are considered in the safety analysis of light water reactors (LWRs). During a PCM accident, film boiling may occur at the cladding surface and cause a rapid temperature increase in the fuel and the cladding, perhaps producing embrittlement of the zircaloy cladding by oxidation. Molten fuel may be produced at the center of the pellets, extrude radially through open cracks in the outer, unmelted portion of the pellet and relocate in the fuel-cladding gap. If the amount of extruded molten fuel is sufficient to establish contact with the cladding, which is at a high temperature during film boiling, the zircaloy cladding may melt. The present work assesses the potential for central fuel melting and thermal failure of the zircaloy cladding due to melting upon being contacted by extruded molten UO 2 -fuel during a PCM event

  18. Fast reactor fuel failures and steam generator leaks: Transient and accident analysis approaches

    1996-10-01

    This report consists of a survey of activities on transient and accident analysis for the LMFR. It is focused on the following subjects: Fuel transient tests and analyses in hypothetical incident/accident situations; sodium-water interaction in steam generators, and sodium fires: test and analyses. There are also sections dealing with the experimental and analytical studies of: fuel subassembly failures; sodium boiling, molten fuel-coolant interaction; molten material movement and relocation in fuel bundles; heat removal after an accident or incident; sodium-water reaction in steam generator; steam generator protection systems; sodium-water contact in steam generator building; fire-fighting methods and systems to deal with sodium fires. Refs, figs, tabs

  19. Fracture Failure Analysis of Fuel Pump Transmission Shaft of Dual-Fuel Engine

    Chen Pei-hong

    2017-01-01

    Full Text Available NTS6ZLCz-129 dual-fuel turbocharged and intercooled engine durability test at 1000h, fuel pump shaft fractured. Fracture analysis, chemical analysis, microstructure examination and finite element stress analysis were carried out on the fractured shaft. The analysis result showed that the shaft fracture cause is forging fold. By improving the forging process, the forging fold was solved, and the durability test can be carried out smoothly.

  20. Alternative Observers for SI Engine Air/Fuel Ratio Control

    Hendricks, Elbert; Poulsen, Jannik; Olsen, Mads Bruun

    1996-01-01

    In earlier work it has been shown that a nonlinear observer based on the use of the manifold pressure state equation and a nonlinear fuel film compensator can maintain accurate A/F ratio control during both steady state and transient operation. This observer may be called a manifold absolute pres...... engine control system designer with a variety of robust control systems which can easily be made redundant in order to satisfy newer engine emissions and diagnosis requirements and legislation...

  1. On the geometry of the fuel rod supports concerning a fretting wear failure

    Kim, Hyung-Kyu; Lee, Young-Ho; Lee, Kang-Hee

    2008-01-01

    Geometrical conditions of spacer grid springs and dimples of a light water reactor fuel assembly are studied in this paper concerning a fuel rod's fretting wear failure. In this framework, the springs/dimples are categorized from the aspects of their orientation with respect to the fuel axis and the contact types. Possible motions on the contacts between the springs/dimples and fuel rods are estimated by conducting a flow-induced vibration test. Features of the wear scar and depth are investigated by independent fretting wear tests carried out with spring and dimple specimens of typical contact geometries. It is also attempted here to apply the contact mechanics theory to a fuel fretting wear analysis such as the prediction of a wear depth profile and its rate, which is influenced by the contact shape of the springs/dimples. It is shown that the theory can be applied to a dimensional control of a coining for the springs/dimples, which is usually carried out in a thin plate fabrication. From the results, the necessary conditions for a spring/dimple geometry for restraining a fretting wear failure are discussed

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

    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

  3. Accidents and failures related to nuclear fuel facilities and nuclear power stations in fiscal 1982

    1983-01-01

    In the chemical preparation room of the reprocessing plant in the Tokai Establishment, Power Reactor and Nuclear Fuel Development Corp., the nasal contamination of small amount (3.6 pCi at maximum) was detected on two workers in June, 1982, but abnormality was not observed in the Lung-monitor of the workers themselves. There was not the effect to the surrounding environment. The failures reported by electric power companies to the Agency of Natural Resources and Energy in accordance with the laws related to atomic energy were 26 cases. The main causes were 5 cases due to improper design management, 2 cases due to improper manufacture management, 4 cases due to improper construction management, 11 cases due to improper maintenance management and 4 other cases. Among those 26 cases, 17 cases occurred in operation, and 9 cases occurred or were detected during shutdown such as regular inspection. Among the 17 cases, 7 cases were the automatic stop by reactor protection system, and 10 cases were the finding by regular in-operation inspection. Among the 9 cases, 5 cases were the breaking of steam generator tubes, 2 cases the breaking of bumper plates at feed heater entrance, and 2 cases other troubles. Moreover, there were 41 minor troubles. (Kako, I.)

  4. Development of advanced fabrication technology for high-temperature gas-cooled reactor fuel. Reduction of coating failure fraction

    Minato, Kazuo; Kikuchi, Hironobu; Fukuda, Kousaku; Tobita, Tsutomu; Yoshimuta, Sigeharu; Suzuki, Nobuyuki; Tomimoto, Hiroshi; Nishimura, Kazuhisa; Oda, Takafumi

    1998-11-01

    The advanced fabrication technology for high-temperature gas-cooled reactor fuel has been developed to reduce the coating failure fraction of the fuel particles, which leads to an improvement of the reactor safety. The present report reviews the results of the relevant work. The mechanisms of the coating failure of the fuel particles during coating and compaction processes of the fuel fabrication were studied to determine a way to reduce the coating failure fraction of the fuel. The coating process was improved by optimizing the mode of the particle fluidization and by developing the process without unloading and loading of the particles at intermediate coating process. The compaction process was improved by optimizing the combination of the pressing temperature and the pressing speed of the overcoated particles. Through these modifications of the fabrication process, the quality of the fuel was improved outstandingly. (author)

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

    Fujishiro, Toshio

    1978-03-01

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

  6. Modification of fuel failure detection system at multi-purpose reactor RSG-GAS, BATAN

    Haruyama, Mitsuo; Shitomi, Hajime; Nakamura, Kiyoshi

    2003-03-01

    As one of the technical cooperation activity based on the Annex III, the Cooperation in the Area of Reactor Physics and Technology, of the Arrangement between the National Energy Agency (BATAN) and the Japan Atomic Energy Research Institute (JAERI), the modification of the Fuel Failure Detection System (FFDS) was carried out by the joint work at the Multi-purpose Reactor RSG-G.A. Siwabessy (RSG-GAS). The system takes the delayed neutron detection method. In normal state, as the background, it measures the gloss delayed neutron concentration emitted in the primary coolant from the fission product (FP) nuclides, which are resulted from a very small amount of fissile material contamination on the fuel plate surface at the fabrication process. When a failure happened at fuel cladding, FP leaks from the fuel meat into the primary coolant. The system shows so higher indication than at normal state, then, the fuel failure can be detected at the early stage and be minimized the damages to the reactor facility and to the environment. The system has been installed at first since November 1994 and applied for reactor operation. However, recently it is not easy to maintain the system for aging degradation and shortage of the spare units and the parts difficult to find in the markets. The modification of FFDS is required for safe and steady reactor operation. The design requirements of the modification are, - To save the system units currently used and the spares on hand as long as practicable, and/or - To replace the system units with those easy to maintain or to obtain at the markets. The modified system obtained around twice of higher sensitivity for delayed neutron detection than before and more reliable monitoring possibility with redundancy. The specification, installation, adjustment methods and characteristics of the modified system and the modus operandi of FFDS at high power reactor operation are described in this paper. (author)

  7. Influence of some fabrication parameters and operating conditions on the PCI failure occurrence in LWR fuel rods

    Bouffioux, P.

    1980-01-01

    In recent LWR designs, the fuel rod failures are induced by a chemically assisted mechanical process, i.e. stress corrosion cracking. The analytical approach towards the analysis of PCI-SCC failures is mainly based on the predictions of the COMETHE code. The failure criteria rely on the concept of a stress threshold together with fission product availability. In the present paper, the use of the COMETHE code to minimize PCI induced clad failure occurrences is illustrated by parametric studies to define acceptable fuel specifications and reactor operating conditions (steady and transient). (author)

  8. Fuel pin failure root causes and power distribution gradients in WWER cores

    Mikus, J.

    2008-01-01

    The purpose of this work is to investigate the influence of some core heterogeneities and reactor construction materials on space power distribution in WWER type cores, especially from viewpoint of the values and gradient occurrence that could result in static loads with some consequences, e.g., fuel pin (FP) or fuel assembly (FA) bowing and possible contribution to the FP failure root causes. Presented information were obtained by means of experiments on research reactor LR-0 concerning the: 1) Power distribution estimation on pellet surface of the FPs neighbouring a FP containing gadolinium (Gd 2 O 3 ) burnable absorber integrated into fuel in WWER-440 and -1000 type cores; 2) Power distribution measurement in periphery FAs neighbouring the baffle in WWER-1000 type cores and 3) Power distribution in FAs neighbouring the control rod absorbing part in a WWER-440 type core. (author)

  9. Prediction of the fuel failure following a large LOCA using modified gap heat transfer model

    Lee, K.M.; Lee, N.H.; Huh, J.Y.; Seo, S.K.; Choi, J.H.

    1995-01-01

    The modified Ross and Stoute gap heat transfer model in the ELOCA.Mk5 code for CANDU safety analysis is based on a simplified thermal deformation model. A review on a series of recent experiments reveals that fuel pellets crack, relocate, and are eccentrically positioned within the sheath rather than solid concentric cylinders. In this study, more realistic offset crap conductance model is implemented in the code to estimate the fuel failure thresholds usincr the transient conditions of a 100% Reactor Outlet Header (ROH) break LOCA. Based on the offset gap conductance model, the total release of I-131 from the failed fuel elements in the core is reduced from 3876 TBq to 3283 TBq to increase margin for dose limit. (author)

  10. A PCI failure in an experimental MOX fuel rod and its sensitivity analysis

    Marino, A.C.

    2000-01-01

    Within our interest in studying MOX fuel performance, the irradiation of the first Argentine prototypes of PHWR MOX fuels began in 1986 with six rods fabricated at the α Facility (CNEA, Argentina). These experiences were made in the HFR-Petten reactor, Holland. The goal of this experience was to study the fuel behaviour with respect to PMCI-SCC. An experiment for extended burnup was performed with the last two MOX rods. During the experiment the final test ramp was interrupted due to a failure in the rod. The post-irradiation examinations indicated that PCI-SCC was a mechanism likely to produce the failure. At the Argentine Atomic Energy Commission (CNEA) the BACO code was developed for the simulation of a fuel rod thermo-mechanical behaviour under stationary and transient power situations. BACO includes a probability analysis within its structure. In BACO the criterion for safe operation of the fuel is based on the maximum hoop stress being below a critical value at the cladding inner surface; this is related to susceptibility to stress corrosion cracking (SCC). The parameters of the MOX irradiation, the preparation of the experiments and post-irradiation analysis were sustained by the BACO code predictions. We present in this paper an overview of the different experiences performed with the MOX fuel rods and the main findings of the post-irradiation examinations. A BACO code description, a wide set of examples which sustain the BACO code validation, and a special calculation for BU15 experiment attained using the BACO code, including a probabilistic analysis of the influence of rod parameters on performance, are included. (author)

  11. Detection of instrument or component failures in a nuclear plant by Luenberger observers

    Wilburn, N.P.; Colley, R.W.; Alexandro, F.J.; Clark, R.N.

    1985-01-01

    A diagnostic system, which will distinguish between instrument failures (flowmeters, etc.) and component failures (valves, filters, etc.) that show the same symptoms, has been developed for nuclear Plants using Luenberger observers. Luenberger observers are online computer based modules constructed following the technology of Clark [3]. A seventh order model of an FFTF subsystem was constructed using the Advanced Continuous Simulation Language (ACSL) and was used to show through simulation that Luenberger observers can be applied to nuclear systems

  12. A model for predicting pellet-cladding interaction induced fuel rod failure, based on nonlinear fracture mechanics

    Jernkvist, L.O.

    1993-01-01

    A model for predicting pellet-cladding mechanical interaction induced fuel rod failure, suitable for implementation in finite element fuel-performance codes, is presented. Cladding failure is predicted by explicitly modelling the propagation of radial cracks under varying load conditions. Propagation is assumed to be due to either iodine induced stress corrosion cracking or ductile fracture. Nonlinear fracture mechanics concepts are utilized in modelling these two mechanisms of crack growth. The novelty of this approach is that the development of cracks, which may ultimately lead to fuel rod failure, can be treated as a dynamic and time-dependent process. The influence of cyclic loading, ramp rates and material creep on the failure mechanism can thereby be investigated. Results of numerical calculations, in which the failure model has been used to study the dependence of cladding creep rate on crack propagation velocity, are presented. (author)

  13. Out-of-reactor experimental study of fuel-pin failure phenomena

    Wrona, B.J.; Galvin, T.M.; Stahl, D.

    1976-01-01

    Fundamental experiments have been performed with a direct-electrical-heating apparatus, on both unclad and quartz-clad UO 2 pellet stacks, to study the effect of a radial constraint on solid and molten-fuel motion during power transients. Results of simulated transient over-power experiments show that molten UO 2 can be quite mobile when the fuel centerline temperature exceeds the boiling point, i.e., fuel vapor pressures become a significant driving force for relocating molten fuel. For radially constrained pellet stacks, when an escape path was provided around the top pellet, significant upward axial fuel motion occurred prior to cladding rupture. Thus, the time sequence of events shows that potential exists for providing a negative reactivity-feedback effect, which would promote nuclear reactor safety. The data tend to support the existence of a ''pressurized-bottle'' effect, which was observed in high-speed movies

  14. Sensitivity and uncertainty evaluation applied to the failure process of nuclear fuel

    Gomes, Daniel S.

    2017-01-01

    Nuclear power plants must operate with minimal risk. The nuclear power plants licensing process is based on a paired model, combining probabilistic and deterministic approaches to improve fuel rod performance during both steady state and transient events. In this study, performance fuel codes were used to simulate the test rod IFA-650-4, with a burnup of 92 GWd/MTU within a Halden reactor. In a loss-of-coolant test, the cladding failed within 336 s after reaching a temperature of 800 °C. Nuclear systems work with many imprecise values that must be quantified and propagated. These sources were separated by physical models or boundary conditions describing fuel thermal conductibility, fission gas release, and creep rates. These factors change output responses. Manufacturing tolerances show dimensional variations for fuel rods, and boundary conditions within the system are characterized using small ranges that can spread throughout the system. To identify the input parameters that produce output effects, we used Pearson coefficients between input and output. These input values represent uncertainties using a stochastic technique that can define the effect of input parameters on the establishment of realistic safety limits. Random sampling provided a set of runs for independent variables proposed by Wilks' formulation. The number of samples required to achieve the 95 th percentile, with 95% confidence, depending on verifying the confidence interval to each output. The FRAPTRAN code utilized a module to reproduce the plastic response, defining the failure limit of the fuel rod. (author)

  15. Sensitivity and uncertainty evaluation applied to the failure process of nuclear fuel

    Gomes, Daniel S., E-mail: dsgomes@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

    2017-07-01

    Nuclear power plants must operate with minimal risk. The nuclear power plants licensing process is based on a paired model, combining probabilistic and deterministic approaches to improve fuel rod performance during both steady state and transient events. In this study, performance fuel codes were used to simulate the test rod IFA-650-4, with a burnup of 92 GWd/MTU within a Halden reactor. In a loss-of-coolant test, the cladding failed within 336 s after reaching a temperature of 800 °C. Nuclear systems work with many imprecise values that must be quantified and propagated. These sources were separated by physical models or boundary conditions describing fuel thermal conductibility, fission gas release, and creep rates. These factors change output responses. Manufacturing tolerances show dimensional variations for fuel rods, and boundary conditions within the system are characterized using small ranges that can spread throughout the system. To identify the input parameters that produce output effects, we used Pearson coefficients between input and output. These input values represent uncertainties using a stochastic technique that can define the effect of input parameters on the establishment of realistic safety limits. Random sampling provided a set of runs for independent variables proposed by Wilks' formulation. The number of samples required to achieve the 95{sup th} percentile, with 95% confidence, depending on verifying the confidence interval to each output. The FRAPTRAN code utilized a module to reproduce the plastic response, defining the failure limit of the fuel rod. (author)

  16. On-line detection of key radionuclides for fuel-rod failure in a pressurized water reactor.

    Qin, Guoxiu; Chen, Xilin; Guo, Xiaoqing; Ni, Ning

    2016-08-01

    For early on-line detection of fuel rod failure, the key radionuclides useful in monitoring must leak easily from failing rods. Yield, half-life, and mass share of fission products that enter the primary coolant also need to be considered in on-line analyses. From all the nuclides that enter the primary coolant during fuel-rod failure, (135)Xe and (88)Kr were ultimately chosen as crucial for on-line monitoring of fuel-rod failure. A monitoring system for fuel-rod failure detection for pressurized water reactor (PWR) based on the LaBr3(Ce) detector was assembled and tested. The samples of coolant from the PWR were measured using the system as well as a HPGe γ-ray spectrometer. A comparison showed the method was feasible. Finally, the γ-ray spectra of primary coolant were measured under normal operations and during fuel-rod failure. The two peaks of (135)Xe (249.8keV) and (88)Kr (2392.1keV) were visible, confirming that the method is capable of monitoring fuel-rod failure on-line. Copyright © 2016 Elsevier Ltd. All rights reserved.

  17. Service Failure and Recovery in a Public Setting: A Comparative Study of Target and Observing Customers

    SEYED SHAHIN SHARIFI

    2017-01-01

    Previous research has examined the effect of service failure and recovery on target and the observing customers separately, with an emphasis on evaluations by target customers. It has been assumed that the reactions of those customers observing the recovery efforts would mirror those of target customers, or perhaps be even more favourable, given that they are not directly affected by the service failure. This research challenges this pre-conception. Through a series of experimental studies, t...

  18. Fuel element failure detection experiments, evaluation of the experiments at KNK II/1 (Intermediate Report)

    Bruetsch, D

    1983-01-01

    In the frame of the fuel element failure detection experiments at KNK II with its first core the measurement devices of INTERATOM were taken into operation in August 1981 and were in operation almost continuously. Since the start-up until the end of the first KNK II core operation plugs with different fuel test areas were inserted in order to test the efficiency of the different measuring devices. The experimental results determined during this test phase and the gained experiences are described in this report and valuated. All three measuring techniques (Xenon adsorption line XAS, gas-chromatograph GC and precipitator PIT) could fulfil the expectations concerning their susceptibility. For XAS and GC the nuclide specific sensitivities as determined during the preliminary tests could be confirmed. For PIT the influences of different parameters on the signal yield could be determined. The sensitivity of the device could not be measured due to a missing reference measuring point.

  19. Shuttle Primary Reaction Control Subsystem Thruster Fuel Valve Pilot Seal Extrusion: A Failure Correlation

    Waller, Jess; Saulsberry, Regor L.

    2003-01-01

    Pilot operated valves (POVs) are used to control the flow of hypergolic propellants monomethylhydrazine (fuel) and nitrogen tetroxide (oxidizer) to the Shuttle orbiter Primary Reaction Control Subsystem (PRCS) thrusters. The POV incorporates a two-stage design: a solenoid-actuated pilot stage, which in turn controls a pressure-actuated main stage. Isolation of propellant supply from the thruster chamber is accomplished in part by a captive polytetrafluoroethylene (PTFE) pilot seal retained inside a Custom 455.1 stainless steel cavity. Extrusion of the pilot seal restricts the flow of fuel around the pilot poppet, thus impeding or preventing the main valve stage from opening. It can also prevent the main stage from staying open with adequate force margin, particularly if there is gas in the main stage actuation cavity. During thruster operation on-orbit, fuel valve pilot seal extrusion is commonly indicated by low or erratic chamber pressure or failure of the thruster to fire upon command (Fail-Off). During ground turnaround, pilot seal extrusion is commonly indicated by slow gaseous nitrogen (GN2) main valve opening times (greater than 38 ms) or slow water main valve opening response times (greater than 33 ms). Poppet lift tests and visual inspection can also detect pilot seal extrusion during ground servicing; however, direct metrology on the pilot seat assembly provides the most quantitative and accurate means of identifying extrusion. Minimizing PRCS fuel valve pilot seal extrusion has become an important issue in the effort to improve PRCS reliability and reduce associated life cycle costs.

  20. Water Chemistry and Clad Corrosion/Deposition Including Fuel Failures. Proceedings of a Technical Meeting

    2013-03-01

    Corrosion is a principal life limiting degradation mechanism in nuclear steam supply systems, particularly taking into account the trends in increasing fuel burnup, thermal ratings and cycle length. Further, many plants have been operating with varying water chemistry regimes for many years, and issues of crud (deposition of corrosion products on other surfaces in the primary coolant circuit) are of significant concern for operators. At the meeting of the Technical Working Group on Fuel Performance and Technology (TWGFPT) in 2007, it was recommended that a technical meeting be held on the subject of water chemistry and clad corrosion and deposition, including the potential consequences for fuel failures. This proposal was supported by both the Technical Working Group on Advanced Technologies for Light Water Reactors (TWG-LWR) and the Technical Working Group on Advanced Technologies for Heavy Water Reactors (TWG-HWR), with a recommendation to hold the meeting at the National Nuclear Energy Generating Company ENERGOATOM, Ukraine. This technical meeting was part of the IAEA activities on water chemistry, which have included a series of coordinated research projects, the most recent of which, Optimisation of Water Chemistry to Ensure Reliable Water Reactor Fuel Performance at High Burnup and in Ageing Plant (FUWAC) (IAEATECDOC-1666), concluded in 2010. Previous technical meetings were held in Cadarache, France (1985), Portland, Oregon, USA (1989), Rez, Czech Republic (1993), and Hluboka nad Vltavou, Czech Republic (1998). This meeting focused on issues associated with the corrosion of fuel cladding and the deposition of corrosion products from the primary circuit onto the fuel assembly, which can cause overheating and cladding failure or lead to unplanned power shifts due to boron deposition in the clad deposits. Crud deposition on other surfaces increases radiation fields and operator dose and the meeting considered ways to minimize the generation of crud to avoid

  1. Sensitivity analysis of fuel pin failure performance under slow-ramp type transient overpower condition by using a fuel performance analysis code FEMAXI-FBR

    Tsuboi, Yasushi; Ninokata, Hisashi; Endo, Hiroshi; Ishizu, Tomoko; Tatewaki, Isao; Saito, Hiroaki

    2012-01-01

    The FEMAXI-FBR is a fuel performance analysis code and has been developed as one module of core disruptive evaluation system, the ASTERIA-FBR. The FEMAXI-FBR has reproduced the failure pin behavior during slow transient overpower. The axial location of pin failure affects the power and reactivity behavior during core disruptive accident, and failure model of which pin failure occurs at upper part of pin is used by reflecting the results of the CABRI-2 test. By using the FEMAXI-FBR, sensitivity analysis of uncertainty of design parameters such as irradiation conditions and fuel fabrication tolerances was performed to clarify the effect on axial location of pin failure during slow transient overpower. The sensitivity analysis showed that the uncertainty of design parameters does not affect the failure location. It suggests that the failure model with which locations of failure occur at upper part of pin can be adopted for core disruptive calculation by taking into consideration of design uncertainties. (author)

  2. Achieving salt-cooled reactor goals: economics, variable electricity, no major fuel failures - 15118

    Forsberg, C.

    2015-01-01

    The Fluoride-salt-cooled High-temperature Reactor (FHR) with a Nuclear air-Brayton Combined Cycle (NACC) and Firebrick Resistance-Heated Energy Storage (FIRES) is a new reactor concept. The FHR uses High-Temperature Gas-cooled Reactor (HTGR) coated-particle fuel and liquid-salt coolants originally developed for molten salt reactors (MSRs) where the fuel was dissolved in the coolant. The FIRES system consists of high-temperature firebrick heated to high temperatures with electricity at times of low electric prices. For a modular FHR operating with a base-load 100 MWe output, the station output can vary from -242 MWe to +242 MWe. The FHR can be built in different sizes. The reactor concept was developed using a top-down approach: markets, requirements, reactor design. The goals are: (1) increase plant revenue by 50 to 100% relative to base-load nuclear plants with capital costs similar to light-water reactors, (2) enable a zero-carbon nuclear renewable electricity grid, and (3) no potential for major fuel failure and thus no potential for major radionuclide offsite releases in a beyond-design-basis accident (BDBA). The basis for the goals and how they may be achieved is described

  3. High-resolution observations of combustion in heterogeneous surface fuels

    E. Louise Loudermilk; Gary L. Achtemeier; Joseph J. O' Brien; J. Kevin Hiers; Benjamin S. Hornsby

    2014-01-01

    In ecosystems with frequent surface fires, fire and fuel heterogeneity at relevant scales have been largely ignored. This could be because complete burns give an impression of homogeneity, or due to the difficulty in capturing fine-scale variation in fuel characteristics and fire behaviour. Fire movement between patches of fuel can have implications for modelling fire...

  4. Visual observations of fuel disruption in in-pile LMFBR accident experiments

    Wright, S.A.; Mast, P.K.

    1982-01-01

    Sandia National Laboratories has been investigating initiation phase phenomena in a series of Fuel Disruption (FD) experiments since 1977. In this program high speed cinematography is used to observe fuel disruption in in-pile experiments that simulate loss of flow accidents. Thus, these experiments provide high resolution measurements of initial fuel and clad motion with prototypic materials and prototypic heating conditions. The main objective of the FD experiment is to determine the timing (relative to fuel temperature) and the mode of fuel disruption under LOF heating conditions. Observed modes of disruption include fuel swelling, solid state breakup, cracking, ejection of a molten fuel jet, slumping, and rapid expansion of small particles. Because the temperature and character of the fuel at disruption are known, disruption can be correlated with the mechanisms driving the disruption such as fuel vapor pressure, molten fuel expansion, fission gases, and impurity gases

  5. PCI fuel failure analysis: a report on a cooperative program undertaken by Pacific Northwest Laboratory and Chalk River Nuclear Laboratories

    Mohr, C.L.; Pankaskie, P.J.; Heasler, P.G.; Wood, J.C.

    1979-12-01

    Reactor fuel failure data sets in the form of initial power (P/sub i/), final power (P/sub f/), transient increase in power (ΔP), and burnup (Bu) were obtained for pressurized heavy water reactors (PHWRs), boiling water reactors (BWRs), and pressurized water reactors (PWRs). These data sets were evaluated and used as the basis for developing two predictive fuel failure models, a graphical concept called the PCI-OGRAM, and a nonlinear regression based model called PROFIT. The PCI-OGRAM is an extension of the FUELOGRAM developed by AECL. It is based on a critical threshold concept for stress dependent stress corrosion cracking. The PROFIT model, developed at Pacific Northwest Laboratory, is the result of applying standard statistical regression methods to the available PCI fuel failure data and an analysis of the environmental and strain rate dependent stress-strain properties of the Zircaloy cladding

  6. Scanning electron microscopy observations of failures of implant overdenture bars: a case series report.

    Waddell, J Neil; Payne, Alan G T; Swain, Michael V; Kieser, Jules A

    2010-03-01

    Soldered or cast bars are used as a standard of care in attachment systems supporting maxillary and mandibular implant overdentures. When failures of these bars occur, currently there is a lack of evidence in relation to their specific etiology, location, or nature. To investigate the failure process of a case series of six failed soldered bars, four intact soldered bars, and one intact cast milled bar, which had been supporting implant overdentures. A total of 11 different overdenture bars were removed from patients with different configuration of opposing arches. A failed bar (FB) group (n = 6) had failed soldered overdenture bars, which were recovered from patients following up to 2 years of wear before requiring prosthodontic maintenance and repair. An intact bar (IB) group (n = 5) had both soldered bars and a single cast milled bar, which had been worn by patients for 2 to 5 years prior to receiving other aspects of prosthodontic maintenance. All bars were examined using scanning electron microscopy to establish the possible mode of failure (FB) or to identify evidence of potential failure in the future (IB). Evidence of a progressive failure mode of corrosion fatigue and creep were observed on all the FB and IB usually around the solder areas and nonoxidizing gold cylinder. Fatigue and creep were also observed in all the IB. Where the level of corrosion was substantial, there was no evidence of wear from the matrices of the attachment system. Evidence of an instantaneous failure mode, ductile and brittle overload, was observed on the fracture surfaces of all the FB, within the solder and the nonoxidizing gold cylinders, at the solder/cylinder interface. Corrosion, followed by corrosion fatigue, appears to be a key factor in the onset of the failure process for overdenture bars in this case series of both maxillary and mandibular overdentures. Limited sample size and lack of standardization identify trends only but prevent broad interpretation of the findings.

  7. A data processing program for transient sodium boiling and fuel failure propagation tests, (2)

    Hasebe, Takeshi; Isozaki, Tadashi; Satoh, Akihiro; Yamaguchi, Katsuhisa; Haga, Kazuo.

    1983-01-01

    Transient Sodium Boiling Tests and Fuel Failure Propagation Tests are being conducted with the out-of-pile test facility, SIENA, in the Core Safety Section of O-arai Engineering Center. The experimental data are recorded using a digital data acquisition system controlled by a HP-1000E computer. The SICILIAN (Speedy Illustration Code for Inspection Line Anomaly) code was developed to obtain quick graphic outputs of data recorded in the magnetic tapes. The program is written in BASIC and Assembler languages and uses a data processing system composed of a desktop computer HP 9845B, a magnetic tape system, a magnetic disc and an eightcolor plotter. The SICILIAN code enables us to get graphic outputs soon after a run. These outputs are very helpful to inspect anomaly in the instrument circuit and to check the experimental conditions of coming runs. (author)

  8. Mechanistic considerations used in the development of the probability of failure in transient increases in power (PROFIT) pellet-zircaloy cladding (thermo-mechanical-chemical) interactions (pci) fuel failure model

    Pankaskie, P.J.

    1980-05-01

    A fuel Pellet-Zircaloy Cladding (thermo-mechanical-chemical) interactions (PCI) failure model for estimating the Probability of Failure in Transient Increases in Power (PROFIT) was developed. PROFIT is based on (1) standard statistical methods applied to available PCI fuel failure data and (2) a mechanistic analysis of the environmental and strain-rate-dependent stress versus strain characteristics of Zircaloy cladding. The statistical analysis of fuel failures attributable to PCI suggested that parameters in addition to power, transient increase in power, and burnup are needed to define PCI fuel failures in terms of probability estimates with known confidence limits. The PROFIT model, therefore, introduces an environmental and strain-rate dependent Strain Energy Absorption to Failure (SEAF) concept to account for the stress versus strain anomalies attributable to interstitial-dislocation interaction effects in the Zircaloy cladding

  9. Behavior and failure of fresh, hydrided and irradiated Zircaloy-4 fuel claddings under RIA conditions

    Le Saux, M.

    2008-01-01

    The purpose of this study is to characterize and simulate the mechanical behaviour and failure of fresh, hydrided and irradiated (in pressurized water reactors) cold-worked stress relieved Zircaloy-4 fuel claddings under reactivity initiated accident conditions. A model is proposed to describe the anisotropic viscoplastic mechanical behavior of the material as a function of temperature (from 20 C up to 1100 C), strain rate (from 3.10 -4 s -1 up to 5 s -1 ), fluence (from 0 up to 1026 n.m -2 ) and irradiation conditions. Axial tensile, hoop tensile, expansion due to compression and hoop plane strain tensile tests are performed at 25 C, 350 C and 480 C in order to analyse the anisotropic plastic and failure properties of the non-irradiated material hydrided up to 1200 ppm. Material strength and strain hardening depend on temperature and hydrogen in solid solution and precipitated hydride contents. Plastic anisotropy is not significantly modified by hydrogen. The material is embrittled by hydrides at room temperature. The plastic strain that leads to hydride cracking decreases with increasing hydrogen content. The material ductility, which increases with increasing temperature, is not deteriorated by hydrogen at 350 C and 480 C. Macroscopic fracture modes and damage mechanisms depend on specimen geometry, temperature and hydrogen content. A Gurson type model is finally proposed to describe both the anisotropic viscoplastic behavior and the ductile fracture of the material as a function of temperature and hydrogen content. (author) [fr

  10. Temperature noise analysis and sodium boiling detection in the fuel failure mockup

    Sides, W.H. Jr.; Fry, D.N.; Leavell, W.H.; Mathis, M.V.; Saxe, R.F.

    1976-01-01

    Sodium temperature noise was measured at the exit of simulated, fast-reactor fuel subassemblies in the Fuel Failure Mockup (FFM) to determine the feasibility of using temperature noise monitors to detect flow blockages in fast reactors. Also, acoustic noise was measured to determine whether sodium boiling in the FFM could be detected acoustically and whether noncondensable gas entrained in the sodium coolant would affect the sensitivity of the acoustic noise detection system. Information from these studies would be applied to the design of safety systems for operating liquid-metal fast breeder reactors (LMFBRs). It was determined that the statistical properties of temperature noise are dependent on the shape of temperature profiles across the subassemblies, and that a blockage upstream of a thermocouple that increases the gradient of the profile near the blockage will also increase the temperature noise at the thermocouple. Amplitude probability analysis of temperature noise shows a skewed amplitude density function about the mean temperature that varies with the location of the thermocouple with respect to the blockage location. It was concluded that sodium boiling in the FFM could be detected acoustically. However, entrained noncondensable gas in the sodium coolant at void fractions greater than 0.4 percent attenuated the acoustic signals sufficiently that boiling was not detected. At a void fraction of 0.1 percent, boiling was indicated only by the two acoustic detectors closest to the boiling site

  11. Program requirements to determine and relate fuel damage and failure thresholds to anticipated conditions in pressurized water reactors

    Loyd, R.F.; Croucher, D.W.

    1980-03-01

    Anticipated transients, licensing criteria, and damage mechanisms for PWR fuel rods are reviewed. Potential mechanistic fuel rod damage limits for PWRs are discussed. An expermental program to be conducted out-of-pile and in the Engineering Test Reactor (ETR) to generate a safety data base to define mechanistic fuel damage and failure thresholds and to relate these thresholds to the thermal-hydraulic and power conditions in a PWR is proposed. The requirements for performing the tests are outlined. Analytical support requirements are defined

  12. Fabrication of an improved tube-to-pipe header heat exchanger for the Fuel Failure Mockup (FFM) Facility

    Prislinger, J.J.; Jones, R.H.

    1977-05-01

    The procedure used in fabricating an improved tube-to-pipe header heat exchanger for the Fuel Failure Mockup (FFM) Facility is described. Superior performance is accomplished at reduced cost with adherence to the ASME Boiler and Pressure Vessel Code. The techniques used and the method of fabrication are described in detail

  13. Pellet-clad interaction observations in boiling water reactor fuel elements

    Sahoo, K.C.; Bahl, J.K.; Sivaramakrishnan, K.S.; Roy, P.R.

    1981-01-01

    Under a programme to assess the performance of fuel elements of Tarapur Atomic Power Station, post-irradiation examination has been carried out on 18 fuel elements in the first phase. Pellet-clad mechanical interaction behaviour in 14 elements with varying burnup and irradiation history has been studied using eddy current testing technique. The data has been analysed to evaluate the role of pellet-clad mechanical interaction in PCI/SCC failure in power reactor operating conditions. (author)

  14. Failure modes observed on worn surfaces of W-C-Co sputtered coatings

    Ramalho, A.; Cavaleiro, A.; Miranda, A.S.; Vieira, M.T.

    1993-01-01

    During scratch testing, the indenter gives rise to a distribution of stresses similar to that observed in tribocontacts. In this work, r.f.-sputtered W-C-Co coatings deposited from sintered WC + Co (6, 10 and 15 wt.% Co) at various substrate biases were scratched and tested tribologically and the morphology of the damaged surfaces was analysed. The cobalt content of the coatings is the main factor determining their tribological characteristics. The failure modes observed on the worn pin-on-disc tested surfaces are explained and compared with those obtained by scratch testing. In spite of it not being possible to establish quantitative results for the wear resistance of W-C-Co coatings from scratch testing, an estimation can be performed based on the observation of the failure modes in the scratch track. Thus scratch testing can be used to predict the tribological behaviour of coated surfaces. This possibility can reduce the number and cost of tribological tests. (orig.)

  15. Observation Likelihood Model Design and Failure Recovery Scheme toward Reliable Localization of Mobile Robots

    Chang-bae Moon

    2011-01-01

    Full Text Available Although there have been many researches on mobile robot localization, it is still difficult to obtain reliable localization performance in a human co-existing real environment. Reliability of localization is highly dependent upon developer's experiences because uncertainty is caused by a variety of reasons. We have developed a range sensor based integrated localization scheme for various indoor service robots. Through the experience, we found out that there are several significant experimental issues. In this paper, we provide useful solutions for following questions which are frequently faced with in practical applications: 1 How to design an observation likelihood model? 2 How to detect the localization failure? 3 How to recover from the localization failure? We present design guidelines of observation likelihood model. Localization failure detection and recovery schemes are presented by focusing on abrupt wheel slippage. Experiments were carried out in a typical office building environment. The proposed scheme to identify the localizer status is useful in practical environments. Moreover, the semi-global localization is a computationally efficient recovery scheme from localization failure. The results of experiments and analysis clearly present the usefulness of proposed solutions.

  16. Observation Likelihood Model Design and Failure Recovery Scheme Toward Reliable Localization of Mobile Robots

    Chang-bae Moon

    2010-12-01

    Full Text Available Although there have been many researches on mobile robot localization, it is still difficult to obtain reliable localization performance in a human co-existing real environment. Reliability of localization is highly dependent upon developer's experiences because uncertainty is caused by a variety of reasons. We have developed a range sensor based integrated localization scheme for various indoor service robots. Through the experience, we found out that there are several significant experimental issues. In this paper, we provide useful solutions for following questions which are frequently faced with in practical applications: 1 How to design an observation likelihood model? 2 How to detect the localization failure? 3 How to recover from the localization failure? We present design guidelines of observation likelihood model. Localization failure detection and recovery schemes are presented by focusing on abrupt wheel slippage. Experiments were carried out in a typical office building environment. The proposed scheme to identify the localizer status is useful in practical environments. Moreover, the semi-global localization is a computationally efficient recovery scheme from localization failure. The results of experiments and analysis clearly present the usefulness of proposed solutions.

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

    Schmuck, I.

    1990-11-01

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

  18. Clinical observation on the treatment of acute liver failure by combined non-biological artificial liver.

    Li, Maoqin; Sun, Jingxi; Li, Jiaqiong; Shi, Zaixiang; Xu, Jiyuan; Lu, Bo; Cheng, Shuli; Xu, Yanjun; Wang, Xiaomeng; Zhang, Xianjiang

    2016-12-01

    The clinical efficacy and safety of different combinations of non-bio artificial liver in the treatment of acute liver failure was examined. A total of 61 cases were selected under blood purification treatment from the patients with severe acute liver failure admitted to the severe disease department of the hospital from December, 2010 to December, 2015. Three types of artificial liver combinations were observed, i.e., plasma exchange plus hemoperfusion plus continuous venovenous hemodiafiltration (PE+HP+CVVHDF), PE+CVVHDF and HP+CVVHDF. The heart rate (HR), mean arterial pressure (MAP), respiratory index (PaO 2 /FiO 2 ), liver and kidney function indicator, as well as platelet and coagulation function were compared. A comparison before and after the treatment using the three methods, showed improvement in the HRs, MAPs, PaO 2 /FiO 2 , total bilirubins (TBIL) and alanine aminotransferases (ALT) (Prate of 62.3% (38/61), and a viral survival rate of 35.0% (7/20); with the non-viral survival rate being 75.6% (31/41). In conclusion, following the treatment of three types of artificial livers, the function was improved to varying degrees, with the PE+HP+CVVHDF and the PE+CVVHDF method being better. By contrast, after the treatment of non-viral liver failure, the survival rate was significantly higher than the patients with viral liver failure.

  19. Is Thrombus With Subcutaneous Edema Detected by Ultrasonography Related to Short Peripheral Catheter Failure? A Prospective Observational Study.

    Takahashi, Toshiaki; Murayama, Ryoko; Oe, Makoto; Nakagami, Gojiro; Tanabe, Hidenori; Yabunaka, Koichi; Arai, Rika; Komiyama, Chieko; Uchida, Miho; Sanada, Hiromi

    Short peripheral catheter (SPC) failure is an important clinical problem. The purpose of this study was to clarify the relationship between SPC failure and etiologies such as thrombus, subcutaneous edema, and catheter dislodgment using ultrasonography and to explore the risk factors associated with the etiologies. Two hundred catheters that were in use for infusion, excluding chemotherapy, were observed. Risk factors were examined by logistic regression analysis. Sixty catheters were removed as the result of SPC failure. Frequency of thrombus with subcutaneous edema in SPC failure cases was significantly greater than in those cases where therapy was completed without complications (P edema. Results suggest that subsurface skin assessment for catheterization could prevent SPC failure.

  20. In-situ observations of stress-induced thin film failures

    Zhao, Z.B., E-mail: zzhao@firstsolar.co [Delphi Research Labs, 51786 Shelby Parkway, Shelby Twp., MI 48315 (United States); Hershberger, J. [Laird Technologies, 4707 Detroit Avenue, Cleveland, Ohio, 44102 (United States); Bilello, J.C. [Department of Materials Science and Engineering, University of Michigan, Ann Arbor, MI 48109-2136 (United States)

    2010-02-01

    In this work, the failure modes of thin films under thermo-mechanical treatments were observed via in-situ white beam X-ray topography. The in-situ experiments were carried out using an experimental setup on Beamline 2-2 at the Stanford Synchrotron Radiation Laboratory. Magnetron sputtered polycrystalline thin films of Ta and CrN on Si substrates were selected for the present study due to their disparate states of intrinsic residual stresses: the Ta film was anisotropically compressive and the CrN film was isotropically tensile. Under a similar heating-cooling cycle in air, the two types of films exhibited distinct failure modes, which were observed in-situ and in a quasi-real-time fashion. The failures of the samples have been interpreted based on their distinctive growth stress states, superimposed on the additional stress development associated with different forms of thermal instabilities upon heating. These included the formation of oxide for the Ta/Si sample, which led to an increase in compressive stress, and a phase change for the CrN/Si sample, which caused the isotropic stress in the film to become increasingly tensile.

  1. In-situ observations of stress-induced thin film failures

    Zhao, Z.B.; Hershberger, J.; Bilello, J.C.

    2010-01-01

    In this work, the failure modes of thin films under thermo-mechanical treatments were observed via in-situ white beam X-ray topography. The in-situ experiments were carried out using an experimental setup on Beamline 2-2 at the Stanford Synchrotron Radiation Laboratory. Magnetron sputtered polycrystalline thin films of Ta and CrN on Si substrates were selected for the present study due to their disparate states of intrinsic residual stresses: the Ta film was anisotropically compressive and the CrN film was isotropically tensile. Under a similar heating-cooling cycle in air, the two types of films exhibited distinct failure modes, which were observed in-situ and in a quasi-real-time fashion. The failures of the samples have been interpreted based on their distinctive growth stress states, superimposed on the additional stress development associated with different forms of thermal instabilities upon heating. These included the formation of oxide for the Ta/Si sample, which led to an increase in compressive stress, and a phase change for the CrN/Si sample, which caused the isotropic stress in the film to become increasingly tensile.

  2. Sharp burnout failure observed in high current-carrying double-walled carbon nanotube fibers

    Song, Li; Toth, Geza; Wei, Jinquan; Liu, Zheng; Gao, Wei; Ci, Lijie; Vajtai, Robert; Endo, Morinobu; Ajayan, Pulickel M.

    2012-01-01

    We report on the current-carrying capability and the high-current-induced thermal burnout failure modes of 5-20 µm diameter double-walled carbon nanotube (DWNT) fibers made by an improved dry-spinning method. It is found that the electrical conductivity and maximum current-carrying capability for these DWNT fibers can reach up to 5.9 × 105 S m - 1 and over 1 × 105 A cm - 2 in air. In comparison, we observed that standard carbon fiber tended to be oxidized and burnt out into cheese-like morphology when the maximum current was reached, while DWNT fiber showed a much slower breakdown behavior due to the gradual burnout in individual nanotubes. The electron microscopy observations further confirmed that the failure process of DWNT fibers occurs at localized positions, and while the individual nanotubes burn they also get aligned due to local high temperature and electrostatic field. In addition a finite element model was constructed to gain better understanding of the failure behavior of DWNT fibers.

  3. A statistical model for prediction of fuel element failure using the Markov process and entropy minimax principles

    Choi, K.Y.; Yoon, Y.K.; Chang, S.H.

    1991-01-01

    This paper reports on a new statistical fuel failure model developed to take into account the effects of damaging environmental conditions and the overall operating history of the fuel elements. The degradation of material properties and damage resistance of the fuel cladding is mainly caused by the combined effects of accumulated dynamic stresses, neutron irradiation, and chemical and stress corrosion at operating temperature. Since the degradation of material properties due to these effects can be considered as a stochastic process, a dynamic reliability function is derived based on the Markov process. Four damage parameters, namely, dynamic stresses, magnitude of power increase from the preceding power level and with ramp rate, and fatigue cycles, are used to build this model. The dynamic reliability function and damage parameters are used to obtain effective damage parameters. The entropy maximization principle is used to generate a probability density function of the effective damage parameters. The entropy minimization principle is applied to determine weighting factors for amalgamation of the failure probabilities due to the respective failure modes. In this way, the effects of operating history, damaging environmental conditions, and damage sequence are taken into account

  4. A Procedure to Address the Fuel Rod Failures during LB-LOCA Transient in Atucha-2 NPP

    Martina Adorni

    2011-01-01

    Full Text Available Depending on the specific event scenario and on the purpose of the analysis, the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes might be required. This may imply the use of a dedicated fuel rod thermomechanical computer code. This paper provides an outline of the methodology for the analysis of the 2A LB-LOCA accident in Atucha-2 NPP and describes the procedure adopted for the use of the fuel rod thermomechanical code. The methodology implies the application of best estimate thermalhydraulics, neutron physics, and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions are provided by core physics and three-dimensional neutron kinetic coupled thermal-hydraulic system codes calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, which are outside the scope of the current paper.

  5. Modeling a failure criterion for U-Mo/Al dispersion fuel

    Oh, Jae-Yong; Kim, Yeon Soo; Tahk, Young-Wook; Kim, Hyun-Jung; Kong, Eui-Hyun; Yim, Jeong-Sik

    2016-05-01

    The breakaway swelling in U-Mo/Al dispersion fuel is known to be caused by large pore formation enhanced by interaction layer (IL) growth between fuel particles and Al matrix. In this study, a critical IL thickness was defined as a criterion for the formation of a large pore in U-Mo/Al dispersion fuel. Specifically, the critical IL thickness is given when two neighboring fuel particles come into contact with each other in the developed IL. The model was verified using the irradiation data from the RERTR tests and KOMO-4 test. The model application to full-sized sample irradiations such as IRISs, FUTURE, E-FUTURE, and AFIP-1 tests resulted in conservative predictions. The parametric study revealed that the fuel particle size and the homogeneity of the fuel particle distribution are influential for fuel performance.

  6. Modeling a failure criterion for U–Mo/Al dispersion fuel

    Oh, Jae-Yong, E-mail: tylor@kaeri.re.kr [Korea Atomic Energy Research Institute, 111, Daedeok-Daero 989 Beon-Gil, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Tahk, Young-Wook; Kim, Hyun-Jung; Kong, Eui-Hyun; Yim, Jeong-Sik [Korea Atomic Energy Research Institute, 111, Daedeok-Daero 989 Beon-Gil, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of)

    2016-05-15

    The breakaway swelling in U–Mo/Al dispersion fuel is known to be caused by large pore formation enhanced by interaction layer (IL) growth between fuel particles and Al matrix. In this study, a critical IL thickness was defined as a criterion for the formation of a large pore in U–Mo/Al dispersion fuel. Specifically, the critical IL thickness is given when two neighboring fuel particles come into contact with each other in the developed IL. The model was verified using the irradiation data from the RERTR tests and KOMO-4 test. The model application to full-sized sample irradiations such as IRISs, FUTURE, E-FUTURE, and AFIP-1 tests resulted in conservative predictions. The parametric study revealed that the fuel particle size and the homogeneity of the fuel particle distribution are influential for fuel performance.

  7. Evolution of thermal stress and failure probability during reduction and re-oxidation of solid oxide fuel cell

    Wang, Yu; Jiang, Wenchun; Luo, Yun; Zhang, Yucai; Tu, Shan-Tung

    2017-12-01

    The reduction and re-oxidation of anode have significant effects on the integrity of the solid oxide fuel cell (SOFC) sealed by the glass-ceramic (GC). The mechanical failure is mainly controlled by the stress distribution. Therefore, a three dimensional model of SOFC is established to investigate the stress evolution during the reduction and re-oxidation by finite element method (FEM) in this paper, and the failure probability is calculated using the Weibull method. The results demonstrate that the reduction of anode can decrease the thermal stresses and reduce the failure probability due to the volumetric contraction and porosity increasing. The re-oxidation can result in a remarkable increase of the thermal stresses, and the failure probabilities of anode, cathode, electrolyte and GC all increase to 1, which is mainly due to the large linear strain rather than the porosity decreasing. The cathode and electrolyte fail as soon as the linear strains are about 0.03% and 0.07%. Therefore, the re-oxidation should be controlled to ensure the integrity, and a lower re-oxidation temperature can decrease the stress and failure probability.

  8. An observational study of the frequency, severity, and etiology of failures in postoperative care after major elective general surgery.

    Symons, Nicholas R A; Almoudaris, Alex M; Nagpal, Kamal; Vincent, Charles A; Moorthy, Krishna

    2013-01-01

    To investigate the nature of process failures in postoperative care, to assess their frequency and preventability, and to explore their relationship to adverse events. Adverse events are common and are frequently caused by failures in the process of care. These processes are often evaluated independently using clinical audit. There is little understanding of process failures in terms of their overall frequency, relative risk, and cumulative effect on the surgical patient. Patients were observed daily from the first postoperative day until discharge by an independent surgeon. Field notes on the circumstances surrounding any nonroutine or atypical event were recorded. Field notes were assessed by 2 surgeons to identify failures in the process of care. Preventability, the degree of harm caused to the patient, and the underlying etiology of process failures were evaluated by 2 independent surgeons. Fifty patients undergoing major elective general surgery were observed for a total of 659 days of postoperative care. A total of 256 process failures were identified, of which 85% were preventable and 51% directly led to patient harm. Process failures occurred in all aspects of care, the most frequent being medication prescribing and administration, management of lines, tubes, and drains, and pain control interventions. Process failures accounted for 57% of all preventable adverse events. Communication failures and delays were the main etiologies, leading to 54% of process failures. Process failures are common in postoperative care, are highly preventable, and frequently cause harm to patients. Interventions to prevent process failures will improve the reliability of surgical postoperative care and have the potential to reduce hospital stay.

  9. Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

    Arkoma, Asko; Hänninen, Markku; Rantamäki, Karin; Kurki, Joona; Hämäläinen, Anitta

    2015-01-01

    Highlights: • The number of failing fuel rods in a LB-LOCA in an EPR is evaluated. • 59 scenarios are simulated with the system code APROS. • 1000 rods per scenario are simulated with the fuel performance code FRAPTRAN-GENFLO. • All the rods in the reactor are simulated in the worst scenario. • Results suggest that the regulations set by the Finnish safety authority are met. - Abstract: In this paper, the number of failing fuel rods in a large break loss-of-coolant accident (LB-LOCA) in EPR-type nuclear power plant is evaluated using statistical methods. For this purpose, a statistical fuel failure analysis procedure has been developed. The developed method utilizes the results of nonparametric statistics, the Wilks’ formula in particular, and is based on the selection and variation of parameters that are important in accident conditions. The accident scenario is simulated with the coupled fuel performance – thermal hydraulics code FRAPTRAN-GENFLO using various parameter values and thermal hydraulic and power history boundary conditions between the simulations. The number of global scenarios is 59 (given by the Wilks’ formula), and 1000 rods are simulated in each scenario. The boundary conditions are obtained from a new statistical version of the system code APROS. As a result, in the worst global scenario, 1.2% of the simulated rods failed, and it can be concluded that the Finnish safety regulations are hereby met (max. 10% of the rods allowed to fail)

  10. Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

    Arkoma, Asko, E-mail: asko.arkoma@vtt.fi; Hänninen, Markku; Rantamäki, Karin; Kurki, Joona; Hämäläinen, Anitta

    2015-04-15

    Highlights: • The number of failing fuel rods in a LB-LOCA in an EPR is evaluated. • 59 scenarios are simulated with the system code APROS. • 1000 rods per scenario are simulated with the fuel performance code FRAPTRAN-GENFLO. • All the rods in the reactor are simulated in the worst scenario. • Results suggest that the regulations set by the Finnish safety authority are met. - Abstract: In this paper, the number of failing fuel rods in a large break loss-of-coolant accident (LB-LOCA) in EPR-type nuclear power plant is evaluated using statistical methods. For this purpose, a statistical fuel failure analysis procedure has been developed. The developed method utilizes the results of nonparametric statistics, the Wilks’ formula in particular, and is based on the selection and variation of parameters that are important in accident conditions. The accident scenario is simulated with the coupled fuel performance – thermal hydraulics code FRAPTRAN-GENFLO using various parameter values and thermal hydraulic and power history boundary conditions between the simulations. The number of global scenarios is 59 (given by the Wilks’ formula), and 1000 rods are simulated in each scenario. The boundary conditions are obtained from a new statistical version of the system code APROS. As a result, in the worst global scenario, 1.2% of the simulated rods failed, and it can be concluded that the Finnish safety regulations are hereby met (max. 10% of the rods allowed to fail)

  11. Failure probabilities of SiC clad fuel during a LOCA in public acceptable simple SMR (PASS)

    Lee, Youho, E-mail: euo@kaist.ac.kr; Kim, Ho Sik, E-mail: hskim25@kaist.ac.kr; NO, Hee Cheon, E-mail: hcno@kaist.ac.kr

    2015-10-15

    Highlights: • Graceful operating conditions of SMRs markedly lower SiC cladding stress. • Steady-state fracture probabilities of SiC cladding is below 10{sup −7} in SMRs. • PASS demonstrates fuel coolability (T < 1300 °C) with sole radiation in LOCA. • SiC cladding failure probabilities of PASS are ∼10{sup −2} in LOCA. • Cold gas gap pressure controls SiC cladding tensile stress level in LOCA. - Abstract: Structural integrity of SiC clad fuels in reference Small Modular Reactors (SMRs) (NuScale, SMART, IRIS) and a commercial pressurized water reactor (PWR) are assessed with a multi-layered SiC cladding structural analysis code. Featured with low fuel pin power and temperature, SMRs demonstrate markedly reduced incore-residence fracture probabilities below ∼10{sup −7}, compared to those of commercial PWRs ∼10{sup −6}–10{sup −1}. This demonstrates that SMRs can serve as a near-term deployment fit to SiC cladding with a sound management of its statistical brittle fracture. We proposed a novel SMR named Public Acceptable Simple SMR (PASS), which is featured with 14 × 14 assemblies of SiC clad fuels arranged in a square ring layout. PASS aims to rely on radiative cooling of fuel rods during a loss of coolant accident (LOCA) by fully leveraging high temperature tolerance of SiC cladding. An overarching assessment of SiC clad fuel performance in PASS was conducted with a combined methodology—(1) FRAPCON-SiC for steady-state performance analysis of PASS fuel rods, (2) computational fluid dynamics code FLUENT for radiative cooling rate of fuel rods during a LOCA, and (3) multi-layered SiC cladding structural analysis code with previously developed SiC recession correlations under steam environments for both steady-state and LOCA. The results show that PASS simultaneously maintains desirable fuel cooling rate with the sole radiation and sound structural integrity of fuel rods for over 36 days of a LOCA without water supply. The stress level of

  12. Effectiveness of Additives in Improving Fuel Lubricity and Preventing Pump Failure at High Temperature

    2013-01-01

    injector nozzle tests were performed in accordance with procedures set forth in an approved 6.5L diesel engine manual using diesel nozzle tester J...Results Fuel injector nozzle tests were performed in accordance with procedures set forth in an approved 6.5L diesel engine manual using diesel nozzle ...UNCLASSIFIED Fuel Injector Results Fuel injector nozzle tests were performed in accordance with procedures set forth in an approved 6.5L

  13. Effects of topical bevacizumab application on early bleb failure after trabeculectomy: observational case series

    Klos-Rola J

    2013-09-01

    Full Text Available Justyna Klos-Rola, Maria Tulidowicz-Bielak, Tomasz Zarnowski Department of Ophthalmology, Medical University of Lublin, Lublin, Poland Background: The aim of this study was to evaluate the influence of topical bevacizumab on the formation and function of filtering blebs in eyes with early bleb failure after antiglaucoma surgery. Methods: Of all patients who underwent mitomycin-augmented trabeculectomy for glaucoma in the Department of Ophthalmology at the Medical University in Lublin, Poland, between March 2009 and March 2010, a total of 21 eyes from 20 patients with injected filtration bleb 9.8 ± 4.7 days after surgery were included in this observational case series. All patients were treated with standard steroid therapy and topical bevacizumab 5 mg/mL five times a day for 20.9 ± 9.8 days. Patients were followed up every other day, and a full eye examination was performed 14, 30, 60, and 180 days after initiation of treatment. Blebs were evaluated for vascularity by slit-lamp examination with concomitant photographic documentation and intraocular pressure measurement. Results: Elevated functional bleb with significantly reduced vascularity was present in 16 eyes, and was flat and nonfunctional in five eyes. Intraocular pressure in all eyes decreased from a mean of 26.6 ± 9.6 mmHg before surgery to 14.6 ± 7.7 mmHg and 15.8 ± 8.3 mmHg at 2 and 6 months after surgery, respectively. Filtration bleb leak was noted in three eyes while on treatment with bevacizumab. Conclusion: Topical application of bevacizumab might favor functional bleb formation after trabeculectomy in eyes with a high risk of failure. Keywords: glaucoma, trabeculectomy, bleb failure, bevacizumab

  14. Field observations and failure analysis of an excavation damaged zone in the Horonobe Underground Research Laboratory

    Aoyagi, Kazuhei; Ishii, Eiichi; Ishida, Tsuyoshi

    2017-01-01

    In the construction of a deep underground facility, the hydromechanical properties of the rock mass around an underground opening are changed significantly due to stress redistribution. This zone is called an excavation damaged zone (EDZ). In high-level radioactive waste disposal, EDZs can provide a shortcut for the escape of radionuclides to the surface environment. Therefore, it is important to develop a method for predicting the detailed characteristics of EDZs. For prediction of the EDZ in the Horonobe Underground Research Laboratory of Japan, we conducted borehole televiewer surveys, rock core analyses, and repeated hydraulic conductivity measurements. We observed that niche excavation resulted in the formation of extension fractures within 0.2 to 1.0 m into the niche wall, i.e., the extent of the EDZ is within 0.2 to 1.0 m into the niche wall. These results are largely consistent with the results of a finite element analysis implemented with the failure criteria considering failure mode. The hydraulic conductivity in the EDZ was increased by 3 to 5 orders of magnitude compared with the outer zone. The hydraulic conductivity in and around the EDZ has not changed significantly in the two years following excavation of the niche. These results show that short-term unloading due to excavation of the niche created a highly permeable EDZ. (author)

  15. A Barrier Options Approach to Modeling Project Failure : The Case of Hydrogen Fuel Infrastructure

    Engelen, P.J.; Kool, C.J.M.; Li, Y.

    2016-01-01

    Hydrogen fuel cell vehicles have the potential to contribute to a sustainable transport system with zero tailpipe emissions. This requires the construction of a network of fuel stations, a long-term, expensive and highly uncertain investment. We contribute to the literature by including a knock-out

  16. Academic success or failure in nursing students: results of a retrospective observational study.

    Lancia, Loreto; Petrucci, Cristina; Giorgi, Fabio; Dante, Angelo; Cifone, Maria Grazia

    2013-12-01

    Nursing student academic failure is a phenomenon of growing international interest, not only because of its economic impact but also because it negatively affects the availability of future nurses in different healthcare systems. To recruit the students with the highest probability of academic success, an open challenge for universities is to recruit students who have previously demonstrated superior scholastic aptitudes that appear to be associated with a greater likelihood of academic success. Documenting the relationship between the selection methods used when selecting nursing students and academic failure will contribute to the international debate concerning the optimisation of the selection strategies. The principal aim of this study was to investigate the role in predicting nursing student academic success of (1) the upper-secondary diploma grades and (2) the score obtained by students in the nursing degree program admission test. A retrospective observational study was conducted. Five cohorts of nursing students, matriculated in consecutive academic years from 2004 to 2008, in an Italian bachelor's degree program were observed retrospectively. Overall, 61.2% of the 1006 considered students concluded their degree within the legal duration allowed for the nursing degree. Students who failed were those who had lowest grades associated with their upper-secondary diploma coursework (p=0.000) and were male (p=0.000). The grades associated with the upper-secondary diploma coursework, unlike the admission test score, correlates positively with the final degree grade and the average value of degree program examination scores. No correlation was found between the upper-secondary diploma coursework grades and the scores obtained in the test for the nursing degree program admission test (r=-0.037). These results suggest that upper-secondary diploma coursework grades are a parameter that should receive great consideration, especially in cases where there are planned

  17. Fuel rod failure due to marked diametral expansion and fuel rod collapse occurred in the HBWR power ramp experiment

    Yanagisawa, Kazuaki

    1985-12-01

    In the power ramp experiment with the BWR type light water loop at the HBWR, the two pre-irradiated fuel rods caused an unexpected pellet-cladding interaction (PCI). One occurred in the fuel rod with small gap of 0.10 mm, which was pre-irradiated up to the burn-up of 14 MWd/kgU. At high power, the diameter of the rod was increased markedly without accompanying significant axial elongation. The other occurred in the rod with a large gap of 0.23 mm, which was pre-irradiated up to the burn-up of 8 MWd/kgU. The diameter of the rod collapsed during a diameter measurement at the maximum power level. The causes of those were investigated in the present study by evaluating in-core data obtained from equipped instruments in the experiment. It was revealed from the investigation that these behaviours were attributed to the local reduction of the coolant flow occurred in the region of a transformer in the ramp rig. The fuel cladding material is seemed to become softened due to temperature increase caused by the local reduction of the coolant flow, and collapsed by the coolant pressure, either locally or wholly depending on the rod diametral gap existed. (author)

  18. Fission product release assessment for end fitting failure in Candu reactor loaded with CANFLEX-NU fuel bundles

    Oh, Dirk Joo; Jeong, Chang Joon; Lee, Kang Moon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle. 4 refs., 1 fig., 4 tabs. (Author)

  19. Fission product release assessment for end fitting failure in Candu reactor loaded with CANFLEX-NU fuel bundles

    Oh, Dirk Joo; Jeong, Chang Joon; Lee, Kang Moon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle. 4 refs., 1 fig., 4 tabs. (Author)

  20. An imprecise Dirichlet model for Bayesian analysis of failure data including right-censored observations

    Coolen, F.P.A.

    1997-01-01

    This paper is intended to make researchers in reliability theory aware of a recently introduced Bayesian model with imprecise prior distributions for statistical inference on failure data, that can also be considered as a robust Bayesian model. The model consists of a multinomial distribution with Dirichlet priors, making the approach basically nonparametric. New results for the model are presented, related to right-censored observations, where estimation based on this model is closely related to the product-limit estimator, which is an important statistical method to deal with reliability or survival data including right-censored observations. As for the product-limit estimator, the model considered in this paper aims at not using any information other than that provided by observed data, but our model fits into the robust Bayesian context which has the advantage that all inferences can be based on probabilities or expectations, or bounds for probabilities or expectations. The model uses a finite partition of the time-axis, and as such it is also related to life-tables

  1. Observer-Based Fuel Control Using Oxygen Measurement

    Andersen, Palle; Bendtsen, Jan Dimon; Mortensen, Jan Henrik

    is constructed and validated against data obtained at the plant. A Kalman filter based on measurements of combustion air flow led into the furnace and oxygen concentration in the flue gas is designed to estimate the actual coal flow. With this estimate, it becomes possible to close an inner loop around the coal......This report describes an attempt to improve the existing control af coal mills used at the Danish power plant Nordjyllandsværket Unit 3. The coal mills are not equipped with coal flow sensors; thus an observer-based approach is investigated. A nonlinear differential equation model of the boiler...

  2. Experimental study of ballooning and failure of WWER-1000 fuel cans during maximum design basis accident

    Karetnikov, G.V.; Bogdanov, A.S.; Semishkin, V.P.; Bezrukov, Yu.A.; Trushin, A.M.; Frizen, E.A.

    2001-01-01

    The processes of ballooning and fracturing in tubular specimens of Eh635 and Eh110 alloy fuel cans are investigated with the use of cinematography. The investigations are carried out under steady-state conditions in the temperature range from 680 to 900 deg C and at pressure drops on the can from 2 to 12 MPa. Time dependences of circumferential strains are plotted for various temperatures of fuel cans at pressure of 2 MPa. It is shown that strain changes are of linear character at an initial portion of the curve and then an accelerated strain development takes place with transition to fracture. Using methods of nonlinear evaluation for time to fracture the approximation dependences are obtained for fuel cans. Experimental data are intended to form the equations of state for fuel can materials and to verify the program TVEL-3 [ru

  3. Observations of in-reactor endurance and rupture life for fueled and unfueled FTR cladding

    Lovell, A.J.; Christensen, B.Y.; Chin, B.A.

    1979-01-01

    Reactor component endurance limits are important to nuclear experimenters and operators. This paper investigates endurance limits of 316 CW fuel pin cladding. The objective of this paper is to compare and analyze two different sets of FTR fuel pin cladding data. The first data set is from unfueled pressurized cladding irradiated in the Experimental Breeder Reactor No. II (EBR-II). This data set was generated in an assembly in which the temperature was monitored and controlled. The second data set contains observations of breached and unbreached EBR-II test fuel pins covering a large range of temperature, power and burnup conditions

  4. Encephalopathy in acute liver failure resulting from acetaminophen intoxication: new observations with potential therapy.

    Brusilow, Saul W; Cooper, Arthur J L

    2011-11-01

    Hyperammonemia is a major contributing factor to the encephalopathy associated with liver disease. It is now generally accepted that hyperammonemia leads to toxic levels of glutamine in astrocytes. However, the mechanism by which excessive glutamine is toxic to astrocytes is controversial. Nevertheless, there is strong evidence that glutamine-induced osmotic swelling, especially in acute liver failure, is a contributing factor: the osmotic gliopathy theory. The object of the current communication is to present evidence for the osmotic gliopathy theory in a hyperammonemic patient who overdosed on acetaminophen. Case report. Johns Hopkins Hospital. A 22-yr-old woman who, 36 hrs before admission, ingested 15 g acetaminophen was admitted to the Johns Hopkins Hospital. She was treated with N-acetylcysteine. Physical examination was unremarkable; her mental status was within normal limits and remained so until approximately 72 hrs after ingestion when she became confused, irritable, and agitated. She was intubated, ventilated, and placed on lactulose. Shortly thereafter, she was noncommunicative, unresponsive to painful stimuli, and exhibited decerebrate posturing. A clinical diagnosis of cerebral edema and increased intracranial pressure was made. She improved very slowly until 180 hrs after ingestion when she moved all extremities. She woke up shortly thereafter. Despite the fact that hyperammonemia is a major contributing factor to the encephalopathy observed in acute liver failure, the patient's plasma ammonia peaked when she exhibited no obvious neurologic deficit. Thereafter, her plasma ammonia decreased precipitously in parallel with a worsening neurologic status. She was deeply encephalopathic during a period when her liver function and plasma ammonia had normalized. Plasma glutamine levels in this patient were high but began to normalize several hours after plasma ammonia had returned to normal. The patient only started to recover as her plasma glutamine began

  5. Fuel assembly reconstitution

    Morgado, Mario M.; Oliveira, Monica G.N.; Ferreira Junior, Decio B.M.; Santos, Barbara O. dos; Santos, Jorge E. dos

    2009-01-01

    Fuel failures have been happened in Nuclear Power Plants worldwide, without lost of integrity and safety, mainly for the public, environment and power plants workers. The most common causes of these events are corrosion (CRUD), fretting and pellet cladding interaction. These failures are identified by increasing the activity of fission products, verified by chemical analyses of reactor coolant. Through these analyses, during the fourth operation cycle of Angra 2 Nuclear Power Plant, was possible to observe fuel failure indication. This indication was confirmed in the end of the cycle during the unloading of reactor core through leakage tests of fuel assembly, using the equipment called 'In Mast Sipping' and 'Box Sipping'. After confirmed, the fuel assembly reconstitution was scheduled, and happened in April, 2007, where was identified the cause and the fuel rod failure, which was substitute by dummy rods (zircaloy). The cause was fretting by 'debris'. The actions to avoid and prevent fuel assemblies failures are important. The goals of this work are to describe the methodology of fuel assembly reconstitution using the FARE (Fuel Assembly Reconstitution Equipment) system, to describe the results of this task in economic and security factors of the company and show how the fuel assembly failures are identified during operation and during the outage. (author)

  6. A Pilot Study to Evaluate California's Fossil Fuel CO2 Emissions Using Atmospheric Observations

    Graven, H. D.; Fischer, M. L.; Lueker, T.; Guilderson, T.; Brophy, K. J.; Keeling, R. F.; Arnold, T.; Bambha, R.; Callahan, W.; Campbell, J. E.; Cui, X.; Frankenberg, C.; Hsu, Y.; Iraci, L. T.; Jeong, S.; Kim, J.; LaFranchi, B. W.; Lehman, S.; Manning, A.; Michelsen, H. A.; Miller, J. B.; Newman, S.; Paplawsky, B.; Parazoo, N.; Sloop, C.; Walker, S.; Whelan, M.; Wunch, D.

    2016-12-01

    Atmospheric CO2 concentration is influenced by human activities and by natural exchanges. Studies of CO2 fluxes using atmospheric CO2 measurements typically focus on natural exchanges and assume that CO2 emissions by fossil fuel combustion and cement production are well-known from inventory estimates. However, atmospheric observation-based or "top-down" studies could potentially provide independent methods for evaluating fossil fuel CO2 emissions, in support of policies to reduce greenhouse gas emissions and mitigate climate change. Observation-based estimates of fossil fuel-derived CO2 may also improve estimates of biospheric CO2 exchange, which could help to characterize carbon storage and climate change mitigation by terrestrial ecosystems. We have been developing a top-down framework for estimating fossil fuel CO2 emissions in California that uses atmospheric observations and modeling. California is implementing the "Global Warming Solutions Act of 2006" to reduce total greenhouse gas emissions to 1990 levels by 2020, and it has a diverse array of ecosystems that may serve as CO2 sources or sinks. We performed three month-long field campaigns in different seasons in 2014-15 to collect flask samples from a state-wide network of 10 towers. Using measurements of radiocarbon in CO2, we estimate the fossil fuel-derived CO2 present in the flask samples, relative to marine background air observed at coastal sites. Radiocarbon (14C) is not present in fossil fuel-derived CO2 because of radioactive decay over millions of years, so fossil fuel emissions cause a measurable decrease in the 14C/C ratio in atmospheric CO2. We compare the observations of fossil fuel-derived CO2 to simulations based on atmospheric modeling and published fossil fuel flux estimates, and adjust the fossil fuel flux estimates in a statistical inversion that takes account of several uncertainties. We will present the results of the top-down technique to estimate fossil fuel emissions for our field

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

    Luxat, J.C.

    2002-01-01

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

  8. The effect of oxygen on the failure of reactor fuel sheaths during a postulated loss-of-coolant accident

    Ferner, J.; Rosinger, H.E.

    1983-09-01

    The failure model for Zircaloy-4 reactor fuel sheaths was used to study the effect of steam oxidation on sheath burst strain. The model, in the form of a computer program called BURST-3, was used to calculate burst strain for a Zircaloy-4 sheath under arbitrary pressure and temperature sequences in an oxidizing (steam) atmosphere. In particular, BURST-3 was used in a parametric study to predict the sheath behaviour in steam as compared to an inert atmosphere, the effect of heating rate, and the effect of circumferential temperature variations on burst strain. It was found that fuel sheath oxidation, which decreases burst strain, becomes increasingly important with increasing temperature and/or time. An effective oxygen concentration of greater than 0.27 wt. percent will cause the sheath to fail with a negligible strain. The hottest region of a sheath will have the highest oxygen concentration, the largest localized strain, and will be the site of failure. The model predictions were compared to experimental data in the range 900 to 1600 K. Agreement between theory and experiment for all three heating rates (5, 25, and 100 K.s -1 ) was very good

  9. Determination of end-of-life-failure fractions of HTGR-fuel particles by postirradiation annealing and beta autoradiography

    Thiele, B.A.; Herren, M.

    1978-11-01

    Fission-product contamination of the helium coolant of High-Temperature Gas-Cooled Reactors (HTGR) is strongly influenced by the end-of-life (EOL) failed-particle fraction. Knowledge of the EOL-failure fraction is the basis for model calculations to predict the total fission product release from the reactor core. After disintegration of irradiation fuel rods, fuel particles are placed in individual holes of a graphite tray. During a 5-h heat treatment at 1000 0 C in a helium atmosphere failed particles leak fission products, especially the volatile cesium, into the graphite. After unloading a β-autoradiograph of the tray is made. Holes that housed defective particles are identified from black spots on the β-sensitive film. The EOL-failure fraction is the ratio of defective particles to the total number of particles tested. The technique is called PIAA, PostIrradiation Annealing and Autoradiography. The PIAA technique was applied to particles of a Trisocoated highly-enriched UO 2 fissile batch irradiated to a burnup of 35% FIMA at an irradiation temperature of 1250 0 C. Visual examination showed all particles to be intact. From 11 to 47% of the particles had failed, as determined by PIAA. Further, postirradiation examination showed that localized corrosion of the silicon carbide coating by fission-product rare-earth chlorides had occurred

  10. In-pile observations of fuel and clad relocation during LMFBR initiation phase accident experiments - the STAR experiments

    Wright, S.A.; Schumacher, G.; Henkel, P.R.; Royl, P.

    1987-01-01

    A series of seven in-pile experiments (the STAR experiments) were performed in which clad motion and fuel dispersal were observed in small pin bundles with high-speed cinematography. The experimental heating conditions reproduced a range of Loss of Flow (LOF) accident scenarios for the lead subassemblies in LMFBRs. The experiments show strong tendencies for limited clad motion in multiple pin bundles, early fuel disruption and dispersal (prior to fuel melting) in moderate power transients having simultaneous clad melting and fuel disruption. The more recent experiments indicate a possibility of steel vapor driven fuel dispersal after fuel breakup and intimate fuel/steel mixing. (author)

  11. Tracking of fuel particles after pin failure in nominal, loss-of-flow and shutdown conditions in the MYRRHA reactor

    Buckingham, Sophia; Planquart, Philippe [von Karman Institute, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Van Tichelen, Katrien [SCK- CEN, Boeretang 200, 2400 Mol (Belgium)

    2017-02-15

    Highlights: • Quantification of the design and safety of the MYRRHA reactor in the event of a pin failure. • Simulation of different accident scenarios in both forced and natural convection regime. • The accumulation areas at the free-surface in case of the least dense particles depend on the flow regime. • The densest particles form an important deposit at the bottom of the vessel. • Further study of the risk of core blockage requires a detailed model of the core. - Abstract: This work on fuel dispersion aims at quantifying the design and safety of the MYRRHA nuclear reactor. A number of accidents leading to the release of a secondary phase into the primary coolant loop are investigated. Among these scenarios, an incident leading to the failure of one or more of the fuel pins is simulated while the reactor is operating in nominal conditions, but also in natural convection regime either during accident transients such as loss-of-flow or during the normal shut-down of the reactor. Two single-phase CFD models of the MYRRHA reactor are constructed in ANSYS Fluent to represent the reactor in nominal and natural convection conditions. An Euler–Lagrange approach with one-way coupling is used for the flow and particle tracking. Firstly, a steady state RANS solution is obtained for each of the three conditions. Secondly, the particles are released downstream from the core outlet and particle distributions are provided over the coolant circuit. Their size and density are defined such that test cases represent potential extremes that may occur. Analysis of the results highlights different particle behaviors, depending essentially on gravity forces and kinematic effects. Statistical distributions highlight potential accumulation regions that may form at the free-surfaces, on top of the upper diaphragm plate or at the bottom of the vessel. These results help to localize regions of fuel accumulation in order to provide insight for development of strategies for

  12. Acute renal failure in critically ill newborns increases the risk of death: a prospective observational study from India.

    Patel, Ankur; Sharma, Deepak; Shastri, Sweta; Sharma, Pradeep

    2016-09-01

    To determine the incidence and risk factors of acute renal failure (ARF) in hospitalized critically ill neonates and analyze outcome of all neonates with renal failure in relation to risk factors. In this prospective observational study 815 infants were enrolled. Renal profile (blood urea and serum creatinine) was done after 12 h of life (or at the time of admission for outborn babies) and then every 12 hourly. Daily 24 h urine output was evaluated. Incidence of renal failure in critically ill neonates was 10.67%. Out of 87 ARF neonates 52 (60%) expired. Mortality in the renal failure group was significantly higher in comparison to control group (p renal failure was more common than oliguric renal failure, but mortality in the oliguric renal failure group was more. Neonatal sepsis was the most common cause of ARF. Eight neonates underwent peritoneal dialysis (PD) out of which there were seven neonatal deaths. Prognosis of neonates with ARF requiring PD was very poor. It can thus be concluded that the health care personal should do rapid diagnosis of ARF in neonates with potential risk factors and also goal at an early and effective treatment of these risk factors in neonates with ARF.

  13. Mesenchymal Stem Cell Benefits Observed in Bone Marrow Failure and Acquired Aplastic Anemia

    Gonzaga, Vivian Fonseca; Lisboa, Gustavo Sabino; Frare, Eduardo Osório

    2017-01-01

    Acquired aplastic anemia (AA) is a type of bone marrow failure (BMF) syndrome characterized by partial or total bone marrow (BM) destruction resulting in peripheral blood (PB) pancytopenia, which is the reduction in the number of red blood cells (RBC) and white blood cells (WBC), as well as platelets (PLT). The first-line treatment option of AA is given by hematopoietic stem cell (HSCs) transplant and/or immunosuppressive (IS) drug administration. Some patients did not respond to the treatment and remain pancytopenic following IS drugs. The studies are in progress to test the efficacy of adoptive cellular therapies as mesenchymal stem cells (MSCs), which confer low immunogenicity and are reliable allogeneic transplants in refractory severe aplastic anemia (SAA) cases. Moreover, bone marrow stromal cells (BMSC) constitute an essential component of the hematopoietic niche, responsible for stimulating and enhancing the proliferation of HSCs by secreting regulatory molecules and cytokines, providing stimulus to natural BM microenvironment for hematopoiesis. This review summarizes scientific evidences of the hematopoiesis improvements after MSC transplant, observed in acquired AA/BMF animal models as well as in patients with acquired AA. Additionally, we discuss the direct and indirect contribution of MSCs to the pathogenesis of acquired AA. PMID:29333168

  14. Reactor fueling system

    Hattori, Noriaki; Hirano, Haruyoshi.

    1983-01-01

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

  15. Failure analysis of top nozzle holddown spring screw for nuclear fuel assembly

    Koh, S. K.; Ryu, C. H.; Na, E. G.; Baek, T. H.; Jeon, K. L.

    2003-01-01

    A failure analysis of holddown spring screw was performed using fracture mechanics approach. The spring screw was designed such that it was capable of sustaining the loads imposed by the initial tensile preload and operational loads. In order to investigate the cause of failure, a stress analysis of the top nozzle spring assembly was done using finite element analysis and a life prediction of the screw was made using a fracture mechanics approach. The elastic-plastic finite element analysis showed that the local stresses at the critical regions of head-shank fillet and thread root significantly exceeded than the yield strength of the screw material, resulting in local plastic deformation. Primary water stress corrosion cracking life of the Inconel 600 screw was predicted by using integration of the Scott model and resulted in 1.42 years, which was fairly close to the actual service life of the holddown spring screw

  16. Fuel failure at the Laguna Verde unit 1- during Cycle 4

    Espinosa Vega, Juan Manuel

    1996-01-01

    The present work describes the event occurred at the Laguna Verde nuclear power plants Unit 1 during its fourth cycle ensembles; the first failure, by means of a test of power suppression, and the second one, during the sipping accomplished in the four refuelling of the unit. Also it describes the re-evaluation of the event accomplished by the licenser, the manufacturer and the Mexican agency

  17. Fuzzy-based failure mode and effect analysis (FMEA) of a hybrid molten carbonate fuel cell (MCFC) and gas turbine system for marine propulsion

    Ahn, Junkeon; Noh, Yeelyong; Park, Sung Ho; Choi, Byung Il; Chang, Daejun

    2017-10-01

    This study proposes a fuzzy-based FMEA (failure mode and effect analysis) for a hybrid molten carbonate fuel cell and gas turbine system for liquefied hydrogen tankers. An FMEA-based regulatory framework is adopted to analyze the non-conventional propulsion system and to understand the risk picture of the system. Since the participants of the FMEA rely on their subjective and qualitative experiences, the conventional FMEA used for identifying failures that affect system performance inevitably involves inherent uncertainties. A fuzzy-based FMEA is introduced to express such uncertainties appropriately and to provide flexible access to a risk picture for a new system using fuzzy modeling. The hybrid system has 35 components and has 70 potential failure modes, respectively. Significant failure modes occur in the fuel cell stack and rotary machine. The fuzzy risk priority number is used to validate the crisp risk priority number in the FMEA.

  18. Clinical characteristics and outcome of heart failure and captagon amphetamine use: An observational prospective study

    Abdelfatah A. Elasfar

    2014-03-01

    Conclusios: Captagon use was found to be an independent risk factor of death and other morbidities in patients presented with cardiomyopathy and acute heart failure. Our study underscores the importance of improving education concerning the cardiac risks of captagon use.

  19. The evaluation of failure stress and released amount of fission product gas of power ramped rod by fuel behaviour analysis code 'FEMAXI-III'

    Yanagisawa, Kazuaki; Fujita, Misao

    1984-01-01

    Pellet-Cladding Interaction(PCI) related in-pile failure of Zircaloy sheathed fuel rod is in general considered to be caused by combination of pellet-cladding mechanical interaction(PCMI) with fuel-cladding chemical interaction(FCCI). An understanding of a basic mechanism of PCI-related fuel failure is therefore necessary to get actual cladding hoop stress from mechanical interaction and released amounts of fission product(FP) gas of aggressive environmental agency from chemical interaction. This paper describes results of code analysis performed on fuel failure to cladding hoop stress and amounts of FP gas released under the condition associated with power ramping. Data from Halden(HBWR) and from Studsvik(R2) are used for code analysis. The fuel behaviour analysis code ''FEMAXI-III'' is used as an analytical tool. The followings are revealed from the study: (1) PCI-related fuel failure is dependent upon cladding hoop stress and released amounts of FP gas at power ramping. (2) Preliminary calculated threshold values of hoop stress and of released amounts of FP gas to PCI failure are respectively 330MPa, 10% under the Halden condition, 190MPa, 5% under the Inter ramp(BWR) condition, and 270MPa, 14% under the Over ramp(PWR) condition. The values of hoop stress calculated are almost in the similar range of those obtained from ex-reactor PCI simulated tests searched from references published. (3) The FEMAXI-III code verification is made in mechanical manner by using in-pile deformation data(diametral strain) obtained from power ramping test undertaken by JAERI. While, the code verification is made in thermal manner by using punctured FP gas data obtained from post irradiation examination performed on non-defected power ramped fuel rods. The calculations are resulted in good agreements to both, mechanical and thermal experimental data suggesting the validity of the code evaluation. (J.P.N.)

  20. Materials properties utilization in a cumulative mechanical damage function for LMFBR fuel pin failure analysis

    Jacobs, D.C.

    1977-01-01

    An overview is presented of one of the fuel-pin analysis techniques used in the CRBRP program, the cumulative mechanical damage function. This technique, as applied to LMFBR's, was developed along with the majority of models used to describe the mechanical properties and environmental behavior of the cladding (i.e., 20 percent cold-worked, 316 stainless steel). As it relates to fuel-pin analyses the Cumulative Mechanical Damage Function (CDF) continually monitors cladding integrity through steady state and transient operation; it is a time dependent function of temperature and stress which reflects the effects of both the prior mechanical history and the variations in mechanical properties caused by exposure to the reactor environment

  1. Clinical characteristics and outcome of heart failure and captagon amphetamine use: An observational prospective study

    Abdelfatah A. Elasfar; Kamal Eldein Ahmad; Waleed AlShaghaa

    2014-01-01

    The fenetylline (captagon) tablets (an amphetamine like substance) are a stimulant drugs which are widely used in the Arabian Peninsula. Objectives: The aim of this study was to evaluate the clinical characteristics and outcome of acute heart failure in patients using captagon tablets. Methods: From September, 2009, through December, 2011, 280 consecutive patients with acute dilated cardiomyopathy and acute heart failure syndrome presented to emergency department in one tertiary care ce...

  2. Austrian contributions to fuel rod failure models shown at the International Standard Problem ISP-14

    Sdouz, G.

    1984-04-01

    The computer code BALON-2A was improved to perform the International Standard Problem ISP-14. The main extensions are the implementation of input-options and the development of a model to predict the pressure in the fuel rod gap. With these improvements and some calculations for input values satisfying results have been obtained. This is remarkable because loss of coolant accident analyses are performed usually with larger computer codes. (Author) [de

  3. Retrieval system for emplaced spent unreprocessed fuel (SURF) in salt bed depository. Baseline concept criteria specifications and mechanical failure probabilities

    Hudson, E.E.; McCleery, J.E.

    1979-05-01

    One of the integral elements of the Nuclear Waste Management Program is the material handling task of retrieving Canisters containing spent unreprocessed fuel from their emplacement in a deep geologic salt bed Depository. A study of the retrieval concept data base predicated this report. In this report, alternative concepts for the tasks are illustrated and critiqued, a baseline concept in scenario form is derived and basic retrieval subsystem specifications are presented with cyclic failure probabilities predicted. The report is based on the following assumptions: (a) during retrieval, a temporary radiation seal is placed over each Canister emplacement; (b) a sleeve, surrounding the Canister, was initially installed during the original emplacement; (c) the emplacement room's physical and environmental conditions established in this report are maintained while the task is performed

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

    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.

  5. Comparison of SCDAP/RELAP5/MOD3 to TRAC-PF1/MOD1 for timing analysis of PWR fuel pin failures

    Jones, K.R.; Katsma, K.R.; Wade, N.L.; Siefken, L.J.; Straka, M.

    1991-01-01

    A comparison has been made of SCDAP/RELAP5/MOD3- and TRAC-PF1/MOD1- based calculations of the fuel pin failure timing (time from containment isolation signal to first fuel pin failure) in a loss-of-coolant accident (LOCA). The two codes were used to calculate the thermal-hydraulic boundary conditions for a complete, double-ended, offset-shear break of a cold leg in a Westinghouse 4-loop pressurized water reactor. Both calculations used the FRAPCON-2 code to calculate the steady-state fuel rod behavior and the FRAP-T6 code to calculate the transient fuel rod behavior. The analysis was performed for 16 combinations of fuel burnups and power peaking factors extending up to the Technical Specifications limits. While all calculations were made on a best-estimate basis, the SCDAP/RELAP5/MOD3 code has not yet been fully assessed for large-break LOCA analysis. The results indicate that SCDAP/RELAP5/MOD3 yields conservative fuel pin failure timing results in comparison to those generated using TRAC-PF1/MOD1. 7 refs., 5 figs

  6. He's Skilled, She's Lucky: A Meta-analysis of Observers' Attributions for Women's and Men's Successes and Failures.

    Swim, Janet K.; Sanna, Lawrence J.

    1996-01-01

    This meta-analysis builds on past qualitative reviews examining different attributions that observers give for other women's and men's successes and failures. Results suggest the greatest support for the argument that differences in expectations for women's and men's performances on masculine tasks influence the selection of stable or unstable…

  7. Favorable bed utilization and readmission rates for emergency department observation unit heart failure patients.

    Schrager, Justin; Wheatley, Matthew; Georgiopoulou, Vasiliki; Osborne, Anwar; Kalogeropoulos, Andreas; Hung, Olivia; Butler, Javed; Ross, Michael

    2013-06-01

    The objective was to compare readmission rates and hospital bed-days between acute decompensated heart failure (AHF) patients admitted or discharged following accelerated treatment protocol (ATP)-driven care in an emergency department observation unit (OU). This was a retrospective cohort study conducted at two urban university-affiliated hospitals. A total of 358 selected AHF patients received treatment on an ATP in the OU between October 1, 2007, and June 30, 2011. The comparison of interest was admission or discharge following OU treatment. The outcome of interest was readmission within 30 and 90 days of hospital discharge following care in the OU. We also examined resource use (inpatient, inpatient plus outpatient-days) between the admitted and discharged groups. Time to readmission analysis was performed with Cox proportional hazards regression. Discharged and admitted patients were similar with respect to age, race, sex, ED length of stay (LOS), and OU LOS. Patients admitted from the OU had a higher median B-type natriuretic peptide (BNP; 1,063 pg/mL [interquartile range {IQR} = 552 to 2,067 pg/mL] vs. 708 pg/mL [IQR = 254 to 1,683 pg/mL]; p = 0.002) and blood urea nitrogen (BUN; 19 mg/dL [IQR = 14 to 26 mg/dL] vs. 17 mg/dL [IQR = 13 to 23 mg/dL]) than those discharged (p = 0.04) and a lower median ejection fraction (EF; 22.5% [15% to 43%] vs. 35% [IQR 20% to 55%]; p = 0.002). In models controlling for age, race, sex, clinical site, BNP, BUN, creatinine, and EF, the 30-day readmission rate (13.8% in the study population as a whole) was not significantly different between the patients discharged or admitted following OU care (hazard ratio [HR] = 0.99; 95% confidence interval [CI] = 0.47 to 2.10). The readmission rates were also not significantly different at 90 days (HR = 1.07; 95% CI = 0.65 to 1.77). Within 30 days of discharge from the OU, patients spent a median of 1.7 days (IQR = 0.0 to 5.1 days) as inpatients, compared to 3.5 days (IQR = 2.3 to 5.8 days

  8. Statistical analysis of failure time in stress corrosion cracking of fuel tube in light water reactor

    Hirao, Keiichi; Yamane, Toshimi; Minamino, Yoritoshi

    1991-01-01

    This report is to show how the life due to stress corrosion cracking breakdown of fuel cladding tubes is evaluated by applying the statistical techniques to that examined by a few testing methods. The statistical distribution of the limiting values of constant load stress corrosion cracking life, the statistical analysis by making the probabilistic interpretation of constant load stress corrosion cracking life, and the statistical analysis of stress corrosion cracking life by the slow strain rate test (SSRT) method are described. (K.I.)

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

    Herszage, A.; Toren, M.

    1998-01-01

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

  10. A CFD model for analysis of performance, water and thermal distribution, and mechanical related failure in PEM fuel cells

    Maher A.R. Sadiq Al-Baghdadi

    2016-07-01

    Full Text Available This paper presents a comprehensive three–dimensional, multi–phase, non-isothermal model of a Proton Exchange Membrane (PEM fuel cell that incorporates significant physical processes and key parameters affecting the fuel cell performance. The model construction involves equations derivation, boundary conditions setting, and solution algorithm flow chart. Equations in gas flow channels, gas diffusion layers (GDLs, catalyst layers (CLs, and membrane as well as equations governing cell potential and hygro-thermal stresses are described. The algorithm flow chart starts from input of the desired cell current density, initialization, iteration of the equations solution, and finalizations by calculating the cell potential. In order to analyze performance, water and thermal distribution, and mechanical related failure in the cell, the equations are solved using a computational fluid dynamic (CFD code. Performance analysis includes a performance curve which plots the cell potential (Volt against nominal current density (A/cm2 as well as losses. Velocity vectors of gas and liquid water, liquid water saturation, and water content profile are calculated. Thermal distribution is then calculated together with hygro-thermal stresses and deformation. The CFD model was executed under boundary conditions of 20°C room temperature, 35% relative humidity, and 1 MPA pressure on the lower surface. Parameters values of membrane electrode assembly (MEA and other base conditions are selected. A cell with dimension of 1 mm x 1 mm x 50 mm is used as the object of analysis. The nominal current density of 1.4 A/cm2 is given as the input of the CFD calculation. The results show that the model represents well the performance curve obtained through experiment. Moreover, it can be concluded that the model can help in understanding complex process in the cell which is hard to be studied experimentally, and also provides computer aided tool for design and optimization of PEM

  11. Healthy lifestyle and decreasing risk of heart failure in women: the Women's Health Initiative observational study.

    Agha, Golareh; Loucks, Eric B; Tinker, Lesley F; Waring, Molly E; Michaud, Dominique S; Foraker, Randi E; Li, Wenjun; Martin, Lisa W; Greenland, Philip; Manson, JoAnn E; Eaton, Charles B

    2014-10-28

    The impact of a healthy lifestyle on risk of heart failure (HF) is not well known. The objectives of this study were to evaluate the effect of a combination of lifestyle factors on incident HF and to further investigate whether weighting each lifestyle factor has additional impact. Participants were 84,537 post-menopausal women from the WHI (Women's Health Initiative) observational study, free of self-reported HF at baseline. A healthy lifestyle score (HL score) was created wherein women received 1 point for each healthy criterion met: high-scoring Alternative Healthy Eating Index, physically active, healthy body mass index, and currently not smoking. A weighted score (wHL score) was also created in which each lifestyle factor was weighted according to its independent magnitude of effect on HF. The incidence of hospitalized HF was determined by trained adjudicators using standardized methodology. There were 1,826 HF cases over a mean follow-up of 11 years. HL score was strongly associated with risk of HF (multivariable-adjusted hazard ratio [HR] [95% confidence interval (CI)] 0.49 [95% CI: 0.38 to 0.62], 0.36 [95% CI: 0.28 to 0.46], 0.24 [95% CI: 0.19 to 0.31], and 0.23 [95% CI: 0.17 to 0.30] for HL score of 1, 2, 3, and 4 vs. 0, respectively). The HL score and wHL score were similarly associated with HF risk (HR: 0.46 [95% CI: 0.41 to 0.52] for HL score; HR: 0.48 [95% CI: 0.42 to 0.55] for wHL score, comparing the highest tertile to the lowest). The HL score was also strongly associated with HF risk among women without antecedent coronary heart disease, diabetes, or hypertension. An increasingly healthy lifestyle was associated with decreasing HF risk among post-menopausal women, even in the absence of antecedent coronary heart disease, hypertension, and diabetes. Weighting the lifestyle factors had minimal impact. Copyright © 2014 American College of Cardiology Foundation. Published by Elsevier Inc. All rights reserved.

  12. WWER identification and analysis of dominant factors affecting the fuel failure rates in WWER-1000 units in Czech Republic, Bulgaria, Ukraine and Russia

    Evdokimov, I.; Likhanskii, V.; Afanasieva, E.; Kanukova, V.; Kozhakin, A.; Maslova, L.; Chernetskiy, M.; Zborovskii, V.; Sorokin, A.

    2015-01-01

    The paper reviews the major findings of the study in the frame of the “Zero Failure Rate” project for WWER. The study included analysis and systematization of available data on leaking fuel assemblies found in 2003 through 2014 in WWER-1000 nuclear units in Russia, Ukraine, Czech Republic and Bulgaria. The study was intended to be used in preparation of recommendations and elaboration of corrective measures for enhancement of reliability and decrease of the failure rates for the WWER-1000 fuel. One of the key areas in successful implementation of the industry ‘zero failure’ goal is a challenge of significant increase of inspections of WWER-1000 fuel assemblies. It may be reasonable (with account taken for international experience) to think of development of more effective equipment for prompt fuel inspections & repair in WWER-1000 spent fuel pool. Another challenge is the elaboration of unified fuel inspection guidelines to ensure that limited industry resources are spent in the most productive way. In the frame of this work it may be helpful to implement in practice the criteria for safe removal of defective fuel rods from the leaking FA under repair

  13. Decentralized Sliding Mode Observer Based Dual Closed-Loop Fault Tolerant Control for Reconfigurable Manipulator against Actuator Failure.

    Bo Zhao

    Full Text Available This paper considers a decentralized fault tolerant control (DFTC scheme for reconfigurable manipulators. With the appearance of norm-bounded failure, a dual closed-loop trajectory tracking control algorithm is proposed on the basis of the Lyapunov stability theory. Characterized by the modularization property, the actuator failure is estimated by the proposed decentralized sliding mode observer (DSMO. Moreover, the actuator failure can be treated in view of the local joint information, so its control performance degradation is independent of other normal joints. In addition, the presented DFTC scheme is significantly simplified in terms of the structure of the controller due to its dual closed-loop architecture, and its feasibility is highly reflected in the control of reconfigurable manipulators. Finally, the effectiveness of the proposed DFTC scheme is demonstrated using simulations.

  14. Decentralized Sliding Mode Observer Based Dual Closed-Loop Fault Tolerant Control for Reconfigurable Manipulator against Actuator Failure

    Zhao, Bo; Li, Yuanchun

    2015-01-01

    This paper considers a decentralized fault tolerant control (DFTC) scheme for reconfigurable manipulators. With the appearance of norm-bounded failure, a dual closed-loop trajectory tracking control algorithm is proposed on the basis of the Lyapunov stability theory. Characterized by the modularization property, the actuator failure is estimated by the proposed decentralized sliding mode observer (DSMO). Moreover, the actuator failure can be treated in view of the local joint information, so its control performance degradation is independent of other normal joints. In addition, the presented DFTC scheme is significantly simplified in terms of the structure of the controller due to its dual closed-loop architecture, and its feasibility is highly reflected in the control of reconfigurable manipulators. Finally, the effectiveness of the proposed DFTC scheme is demonstrated using simulations. PMID:26181826

  15. Chaos emerging in soil failure patterns observed during tillage: Normalized deterministic nonlinear prediction (NDNP) and its application.

    Sakai, Kenshi; Upadhyaya, Shrinivasa K; Andrade-Sanchez, Pedro; Sviridova, Nina V

    2017-03-01

    Real-world processes are often combinations of deterministic and stochastic processes. Soil failure observed during farm tillage is one example of this phenomenon. In this paper, we investigated the nonlinear features of soil failure patterns in a farm tillage process. We demonstrate emerging determinism in soil failure patterns from stochastic processes under specific soil conditions. We normalized the deterministic nonlinear prediction considering autocorrelation and propose it as a robust way of extracting a nonlinear dynamical system from noise contaminated motion. Soil is a typical granular material. The results obtained here are expected to be applicable to granular materials in general. From a global scale to nano scale, the granular material is featured in seismology, geotechnology, soil mechanics, and particle technology. The results and discussions presented here are applicable in these wide research areas. The proposed method and our findings are useful with respect to the application of nonlinear dynamics to investigate complex motions generated from granular materials.

  16. Heart rate awareness in patients with chronic stable heart failure. A multi-center observational study.

    Moran, D

    2014-08-23

    We assessed adherence to European Society of Cardiology heart rate guidelines (i.e. heart rates less than 70bpm) in patients with chronic stable heart failure. We also investigated the percent of patients on target doses of rate controlling drugs.

  17. Hemodynamic changes as a result of experimental heart failure in sheep as observed with technetium 99m

    Van Rooyen, J.M.

    1978-01-01

    The aim of this project is to study heart failure in sheep that have gousiekte by using technetium 99m to determine the blood flow. Gousiekte is a congestive cardiomyopathy that occurs in ruminant animals. It is characterised by a latent period of 2-6 weeks followed by a sudden death. It appears that during gousiekte the stroke volume decreases with about 40% and the PFI increases with more than a 100%. The decrease in stroke volume is observed by means of an electromagnetic bloodflow meter. The greater change in PFI than stroke volume during gousiekte is a sign that there is congestive failure of the left ventricle. A decrease in ejection fraction has been observed by a loss of value lower than 30% after the final congestive phase has been reached. Normally 7,4 plus minus 0,3 contractions of a sheep's heart are needed to pump the blood from the right to left side of the heart. Complete congestive failure during gousiekte can decrease the effectiveness of the heart so that 50 contractions are needed, in other words a PFI value of 50. Two phases can be distinguished during the development of gousiekte namely a compensation phase and a decompensation phase. The PFI as criterion is used to establish the influence of certain drugs with an inotropic effect and to establish whether a gousiekte heart can protentiate after administration of the drugs. The findings are positive which shows than energy abnormalities are not primary causes of gousiekte. As a model for heart failure gousiekte can be compared with other well known models of heart failure such a volume overload, pressure overload and coronary ligatures in sheep. The by-product of this research is the development and possible application of the technetium isotope method to diagnose heart failure in sheep

  18. Control device for can failures of liquid cooled nuclear reactor fuel elements

    Althaus, D.; Mohm, F.; Nyhof, M.

    1974-01-01

    Checking of the fuel or breeding elements of, e.g., sodium-cooled reactors is done by detecting fission products in the coolant, with a flushing gas like argon removing the fission products from the coolant and carrying them to a detector system. In order to increase the escape rate of the fission products from the elements, these are lifted by a hoisting unit into a pit reaching down below the coolant level and are then heated. Heating is achieved by the decay heat or by an additional heating in the receiving pit which at this point is thermically insulated from the exterior. The flushing gas is blown radially into the receiving pit from below. The rising bubbles take along the fission products to a scintillation counter mounted on the head of the receiving vessel. This vessel may have double walls with coolant flowing through the interspace. (DG) [de

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

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

    1979-10-01

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

  20. Retrieval system for emplaced spent unreprocessed fuel (SURF) in salt bed depository: accident event analysis and mechanical failure probabilities. Final report

    Bhaskaran, G.; McCleery, J.E.

    1979-10-01

    This report provides support in developing an accident prediction event tree diagram, with an analysis of the baseline design concept for the retrieval of emplaced spent unreprocessed fuel (SURF) contained in a degraded Canister. The report contains an evaluation check list, accident logic diagrams, accident event tables, fault trees/event trees and discussions of failure probabilities for the following subsystems as potential contributors to a failure: (a) Canister extraction, including the core and ram units; (b) Canister transfer at the hoist area; and (c) Canister hoisting. This report is the second volume of a series. It continues and expands upon the report Retrieval System for Emplaced Spent Unreprocessed Fuel (SURF) in Salt Bed Depository: Baseline Concept Criteria Specifications and Mechanical Failure Probabilities. This report draws upon the baseline conceptual specifications contained in the first report

  1. Serum Procalcitonin and Peripheral Venous Lactate for Predicting Dengue Shock and/or Organ Failure: A Prospective Observational Study.

    Vipa Thanachartwet

    2016-08-01

    Full Text Available Currently, there are no biomarkers that can predict the incidence of dengue shock and/or organ failure, although the early identification of risk factors is important in determining appropriate management to reduce mortality. Therefore, we sought to determine the factors associated with dengue shock and/or organ failure and to evaluate the prognostic value of serum procalcitonin (PCT and peripheral venous lactate (PVL levels as biomarkers of dengue shock and/or organ failure.A prospective observational study was conducted among adults hospitalized for confirmed viral dengue infection at the Hospital for Tropical Diseases in Bangkok, Thailand between October 2013 and July 2015. Data, including baseline characteristics, clinical parameters, laboratory findings, serum PCT and PVL levels, management, and outcomes, were recorded on pre-defined case report forms. Of 160 patients with dengue, 128 (80.0% patients had dengue without shock or organ failure, whereas 32 (20.0% patients developed dengue with shock and/or organ failure. Using a stepwise multivariate logistic regression analysis, PCT ≥0.7 ng/mL (odds ratio [OR]: 4.80; 95% confidence interval [CI]: 1.60-14.45; p = 0.005 and PVL ≥2.5 mmol/L (OR: 27.99, 95% CI: 8.47-92.53; p <0.001 were independently associated with dengue shock and/or organ failure. A combination of PCT ≥0.7 ng/mL and PVL ≥2.5 mmol/L provided good prognostic value for predicting dengue shock and/or organ failure, with an area under the receiver operating characteristics curve of 0.83 (95% CI: 0.74-0.92, a sensitivity of 81.2% (95% CI: 63.6-92.8%, and a specificity of 84.4% (95% CI: 76.9-90.2%. Dengue shock patients with non-clearance of PCT and PVL expired during hospitalization.PCT ≥0.7 ng/mL and PVL ≥2.5 mmol/L were independently associated with dengue shock and/or organ failure. The combination of PCT and PVL levels could be used as prognostic biomarkers for the prediction of dengue shock and/or organ failure.

  2. A simulation of the temperature overshoot observed at high burnup in annular fuel pellets

    Baron, D [Electricite de France, Moret-sur-Loing (France); Couty, J C [Electricite de France (EDF), 69 - Villeurbanne (France)

    1997-08-01

    Instrumented experiments have been carried out in recent years to calibrate and improve temperature calculations at high burnup in PWR nuclear fuel rods. The introduction of a thermocouple in the fuel stack allows the experiment to record the centre-line temperature all along the irradiation or re-irradiation. The results obtained on fresh fuel have not revealed any abnormal behavior as have observations done on high burnup rods. In this case, a sudden overshoot has been recorded on the thermocouple temperature above an average power threshold. Several hypotheses have been suggested. Only two seem to be acceptable: one in relation to an effect of grain decohesion, another based on a modification of fuel chemistry. The apparent reversibility of the phenomena when power decreases led us to prefer the first explanation. Indeed, the introduction of a thermocouple means that annular fuel pellets must be used. These are either initially manufactured with a central hole or drilled after base irradiation, using the ``RISOE`` technique. One must bear in mind that the use of such annular pellets drastically changes the crack pattern as irradiation proceeds. This is due to a different stress field which, combined with a weakening of the grain binding energy, leads to a partial grain decohesion on the inner face of the annular pellet. Modification of the grain binding energy is related to the presence of an increasing local population of gas bubbles and metallic precipitates at grain boundaries, as swelling creates intergranular local stresses which also could probably enhance the grain decohesion process. This grain decohesion concerns a 250 to 350 {mu}m depth and shows a narrow cracks network through which released fission gas can flow, temporarily pushing the resident helium gas out. The low conductivity of these gaseous fission products and the numerous gas layers created this way could partly explain the unexpected temperatures measured in high burnup fuels. (Abstract

  3. Mechanical failure of SKB spent fuel disposal canisters. Mathematical modelling and scoping calculations

    Takase, Hiroyasu; Benbow, S.; Grindrod, P.

    1998-10-01

    According to the current design of SKB, a copper overpack with a cast steel inner component will be used as the disposal canister for spent nuclear fuel. A recent study considered the case of a breach in the copper overpack, through which groundwater could enter the canister. It has pointed out that hydrogen gas generated by an anaerobic corrosion could cushion the system and reduce or eventually stop further infiltration of water into the breached canister, and thence the spent fuel. One potential pitfall in this previous study lies in the fact that it did not consider any processes which might violate the following assumptions which are essential for the gas 'cushioning': 1. Hydrogen gas accumulated in the annular gap in the canister forms a free gas phase which is stable indefinitely into future; 2. Elevated gas pressure in the canister prevents further supply of groundwater except for diffusion of vapour. In the current study we developed a set of mathematical models for the above problem and applied it to carry out an independent assessment of the long-term behaviour of the canister. A key aim in this study was to clarify whether there are any alternative processes which may affect the result obtained by the previous study by violating one of the assumptions listed above. For this purpose, a scenario development exercise was conducted. The result supported the concept described in the previous study. One exception is that possible intrusion of bentonite gel followed by its desaturation could leave paths both for the gas and water simultaneously without forming a gas cushion. This is summarised in the first part of the report. In the second part, development of mathematical models and their applications are described. The key results are: 1. The model describing behaviour of gas and pore water in the canister and the buffer material reproduced the main results of the previous study; 2. The model considering intrusion of the bentonite gel pointed out possibility

  4. Visual observations of a degraded bundle of irradiated fuel: the Phebus FPT1 test

    Barrachin, M.; Bottomley, P.D.

    1999-01-01

    The international Phebus-FP (Fission Product) project is managed by the Institut de Protection et Surete Nucleaire in collaboration with Electricite de France (EDF), the European Commission (EC), the USNRC (USA), COG (Canada), NUPEC and JAERI (Japan), KAERI (South Korea), PSI and HSK (Switzerland). It is designed to measure the source-term and to study the degradation of irradiated UO 2 fuel in conditions typical of a severe loss of coolant accident in a pressurised water reactor (PWR). In the first test (FPT0), performed in December '93, a bundle of 20 fresh fuel rods and a central Ag-In-Cd control rod underwent a short 15-day irradiation to generate fission products before testing in the Phebus reactor in Cadarache. The second test (FPT1) was performed in July '96, in the same conditions and geometry, but using irradiated fuel (-23 GWd/tU). In the FPT1 test, the bundle was heated to an estimated 3000 K over a period of 30 minutes in order to induce a substantial liquefaction of the bundle. After the test, the bundle was embedded in epoxy and cut at different levels to investigate the mechanisms of the core degradation. This paper reports the visual observations of the degraded FPT1 bundle, very preliminary interpretations about the scenario of degradation and a comparison between the behaviour of the fuel in the FPT0 and FPT1 tests. (author)

  5. Behavior of metallic fuel in treat transient overpower tests

    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

  6. Salvage Radiation Therapy Dose Response for Biochemical Failure of Prostate Cancer After Prostatectomy—A Multi-Institutional Observational Study

    Pisansky, Thomas M., E-mail: pisansky.thomas@mayo.edu [Department of Radiation Oncology, Mayo Clinic, Rochester, Minnesota (United States); Agrawal, Shree [Case Western Reserve University School of Medicine, Cleveland, Ohio (United States); Hamstra, Daniel A. [Department of Radiation Oncology, University of Michigan, Ann Arbor, Michigan (United States); Koontz, Bridget F. [Department of Radiation Oncology, Duke Cancer Institute, Durham, North Carolina (United States); Liauw, Stanley L. [Department of Radiation and Cellular Oncology, University of Chicago, Chicago, Illinois (United States); Efstathiou, Jason A. [Department of Radiation Oncology, Massachusetts General Hospital, Boston, Massachusetts (United States); Michalski, Jeff M. [Department of Radiation Oncology, Washington University, St. Louis, Missouri (United States); Feng, Felix Y. [Department of Radiation Oncology, University of Michigan, Ann Arbor, Michigan (United States); Abramowitz, Matthew C.; Pollack, Alan [Department of Radiation Oncology, University of Miami, Miami, Florida (United States); Anscher, Mitchell S. [Department of Radiation Oncology, Virginia Commonwealth University, Richmond, Virginia (United States); Moghanaki, Drew [Department of Radiation Oncology, Virginia Commonwealth University, Richmond, Virginia (United States); Hunter Holmes McGuire Veterans Administration Medical Center, Richmond, Virginia (United States); Den, Robert B. [Department of Radiation Oncology, Thomas Jefferson University, Philadelphia, Pennsylvania (United States); Stephans, Kevin L. [Department of Radiation Oncology, Cleveland Clinic, Cleveland, Ohio (United States); Zietman, Anthony L. [Department of Radiation Oncology, Massachusetts General Hospital, Boston, Massachusetts (United States); Lee, W. Robert [Department of Radiation Oncology, Duke Cancer Institute, Durham, North Carolina (United States); Kattan, Michael W. [Department of Quantitative Health Sciences, Cleveland Clinic, Cleveland, Ohio (United States); and others

    2016-12-01

    Purpose: To determine whether a dose-response relationship exists for salvage radiation therapy (RT) of biochemical failure after prostatectomy for prostate cancer. Methods and Materials: Individual data from 1108 patients who underwent salvage RT at 10 academic centers were pooled. The cohort was enriched for selection criteria more likely associated with tumor recurrence in the prostate bed (margin positive and pre-RT prostate-specific antigen [PSA] level of ≤2.0 ng/mL) and without the confounding of planned androgen suppression. The cumulative incidence of biochemical failure and distant metastasis over time was computed, and competing risks hazard regression models were used to investigate the association between potential predictors and these outcomes. The association of radiation dose with outcomes was the primary focus. Results: With a 65.2-month follow-up duration, the 5- and 10-year estimates of freedom from post-RT biochemical failure (PSA level >0.2 ng/mL and rising) was 63.5% and 49.8%, respectively, and the cumulative incidence of distant metastasis was 12.4% by 10 years. A Gleason score of ≥7, higher pre-RT PSA level, extraprostatic tumor extension, and seminal vesicle invasion were associated with worse biochemical failure and distant metastasis outcomes. A salvage radiation dose of ≥66.0 Gy was associated with a reduced cumulative incidence of biochemical failure, but not of distant metastasis. Conclusions: The use of salvage radiation doses of ≥66.0 Gy are supported by evidence presented in the present multicenter pooled analysis of individual patient data. The observational reporting method, limited sample size, few distant metastasis events, modest follow-up duration, and elective use of salvage therapy might have diminished the opportunity to identify an association between the radiation dose and this endpoint.

  7. Deferred modification of antiretroviral regimen following documented treatment failure in Asia: results from the TREAT Asia HIV Observational Database (TAHOD)

    Zhou, J; Li, PCK; Kumarasamy, N; Boyd, M; Chen, YMA; Sirisanthana, T; Sungkanuparph, S; Oka, S; Tau, G; Phanuphak, P; Saphonn, V; Zhang, FJ; Omar, SFS; Lee, CKC; Ditangco, R; Merati, TP; Lim, PL; Choi, JY; Law, MG; Pujari, S

    2010-01-01

    Objective The aim of the study was to examine the rates and predictors of treatment modification following combination antiretroviral therapy (cART) failure in Asian patients with HIV enrolled in the TREAT Asia HIV Observational Database (TAHOD). Methods Treatment failure (immunological, virological and clinical) was defined by World Health Organization criteria. Countries were categorized as high or low income by World Bank criteria. Results Among 2446 patients who initiated cART, 447 were documented to have developed treatment failure over 5697 person-years (7.8 per 100 person-years). A total of 253 patients changed at least one drug after failure (51.6 per 100 person-years). There was no difference between patients from high- and low-income countries [adjusted hazard ratio (HR) 1.02; P = 0.891]. Advanced disease stage [Centers for Disease Control and Prevention (CDC) category C vs. A; adjusted HR 1.38, P = 0.040], a lower CD4 count (≥ 51 cells/μL vs. ≤ 50 cells/μL; adjusted HR 0.61, P = 0.022) and a higher HIV viral load (≥ 400 HIV-1 RNA copies/mL vs. failure. Compared with patients from low-income countries, patients from high-income countries were more likely to change two or more drugs (67% vs. 49%; P = 0.009) and to change to a protease-inhibitor-containing regimen (48% vs. 16%; Pfailure. This deferred modification is likely to have negative implications for accumulation of drug resistance and response to second-line treatment. There is a need to scale up the availability of second-line regimens and virological monitoring in this region. PMID:19601993

  8. Salvage Radiation Therapy Dose Response for Biochemical Failure of Prostate Cancer After Prostatectomy—A Multi-Institutional Observational Study

    Pisansky, Thomas M.; Agrawal, Shree; Hamstra, Daniel A.; Koontz, Bridget F.; Liauw, Stanley L.; Efstathiou, Jason A.; Michalski, Jeff M.; Feng, Felix Y.; Abramowitz, Matthew C.; Pollack, Alan; Anscher, Mitchell S.; Moghanaki, Drew; Den, Robert B.; Stephans, Kevin L.; Zietman, Anthony L.; Lee, W. Robert; Kattan, Michael W.

    2016-01-01

    Purpose: To determine whether a dose-response relationship exists for salvage radiation therapy (RT) of biochemical failure after prostatectomy for prostate cancer. Methods and Materials: Individual data from 1108 patients who underwent salvage RT at 10 academic centers were pooled. The cohort was enriched for selection criteria more likely associated with tumor recurrence in the prostate bed (margin positive and pre-RT prostate-specific antigen [PSA] level of ≤2.0 ng/mL) and without the confounding of planned androgen suppression. The cumulative incidence of biochemical failure and distant metastasis over time was computed, and competing risks hazard regression models were used to investigate the association between potential predictors and these outcomes. The association of radiation dose with outcomes was the primary focus. Results: With a 65.2-month follow-up duration, the 5- and 10-year estimates of freedom from post-RT biochemical failure (PSA level >0.2 ng/mL and rising) was 63.5% and 49.8%, respectively, and the cumulative incidence of distant metastasis was 12.4% by 10 years. A Gleason score of ≥7, higher pre-RT PSA level, extraprostatic tumor extension, and seminal vesicle invasion were associated with worse biochemical failure and distant metastasis outcomes. A salvage radiation dose of ≥66.0 Gy was associated with a reduced cumulative incidence of biochemical failure, but not of distant metastasis. Conclusions: The use of salvage radiation doses of ≥66.0 Gy are supported by evidence presented in the present multicenter pooled analysis of individual patient data. The observational reporting method, limited sample size, few distant metastasis events, modest follow-up duration, and elective use of salvage therapy might have diminished the opportunity to identify an association between the radiation dose and this endpoint.

  9. Why has the introduction of natural gas vehicles failed in Germany?—Lessons on the role of market failure in markets for alternative fuel vehicles

    Peters von Rosenstiel, Dirk; Heuermann, Daniel F.; Hüsig, Stefan

    2015-01-01

    Despite private investments exceeding two billion Euros and tax incentives of more than 500 million Euros, the market share of natural gas vehicles (NGVs) in Germany has lagged far behind expectations and behind market developments in other countries. With total cost of ownership being on average lower for NGVs than for gasoline and diesel vehicles this raises the question of the existence of market failure in the German NGV-market. We use a case study approach where we combine quantitative data with insights from a multi-industry expert panel and in-depth interviews with experts from industry, government and civil society in order to examine whether and how different types of market failure contribute to the status quo in the German market for NGVs. We conclude that coordination failure in complementary markets, an artificially created monopoly of service stations at motorways, imperfect information, bounded consumer rationality, and principle-agent-problems are the most prominent market failures inhibiting the development of a functioning market for NGVs. Our results are instructive for the design of effective public policies and investor strategies aiming to create markets for alternative fuel vehicles. - Highlights: • We analyze market failure in the German market for natural gas vehicles. • Coordination failure is the most important reason for market failure to arise. • Minor factors: regulatory deficits, imperfect information, bounded rationality. • Policies encompass stabilizing expectations and supporting actor coordination. • Our results are instructive for policies and investor strategies in AFV-markets

  10. Sliding mode observer for proton exchange membrane fuel cell: automotive application

    Piffard, Maxime; Gerard, Mathias; Fonseca, Ramon Da; Massioni, Paolo; Bideaux, Eric

    2018-06-01

    This work proposes a state observer as a tool to manage cost and durability issues for PEMFC (Proton Exchange Membrane Fuel Cell) in automotive applications. Based on a dead-end anode architecture, the observer estimates the nitrogen build-up in the anode side, as well as relative humidities in the channels. These estimated parameters can then be used at fuel cell management level to enhance the durability of the stack. This observer is based on transport equations through the membrane and it reconstructs the behavior of the water and nitrogen inside the channels without the need of additional humidity sensors to correct the estimate. The convergence of the output variables is proved with Lyapunov theory for dynamic operating conditions. The validation is made with a high-fidelity model running a WLTC (Worldwide harmonized Light vehicles Test Cycle). This observer provides the average values of nitrogen and relative humidities with sufficient precision to be used in a global real-time control scheme.

  11. Application of the Nelson model to four timelag fuel classes using Oklahoma field observations: Model evaluation and comparison with national Fire Danger Rating System algorithms

    J. D. Carlson; Larry S. Bradshaw; Ralph M. Nelson; Randall R Bensch; Rafal Jabrzemski

    2007-01-01

    The application of a next-generation dead-fuel moisture model, the 'Nelson model', to four timelag fuel classes using an extensive 21-month dataset of dead-fuel moisture observations is described. Developed by Ralph Nelson in the 1990s, the Nelson model is a dead-fuel moisture model designed to take advantage of frequent automated weather observations....

  12. A simulation of the temperature overshoot observed at high burnup in annular fuel pellets

    Baron, D.; Couty, J.C.

    1997-01-01

    Instrumented experiments have been carried out in recent years to calibrate and improve temperature calculations at high burnup in PWR nuclear fuel rods. The introduction of a thermocouple in the fuel stack allows the experiment to record the centre-line temperature all along the irradiation or re-irradiation. The results obtained on fresh fuel have not revealed any abnormal behavior as have observations done on high burnup rods. In this case, a sudden overshoot has been recorded on the thermocouple temperature above an average power threshold. Several hypotheses have been suggested. Only two seem to be acceptable: one in relation to an effect of grain decohesion, another based on a modification of fuel chemistry. The apparent reversibility of the phenomena when power decreases led us to prefer the first explanation. Indeed, the introduction of a thermocouple means that annular fuel pellets must be used. These are either initially manufactured with a central hole or drilled after base irradiation, using the ''RISOE'' technique. One must bear in mind that the use of such annular pellets drastically changes the crack pattern as irradiation proceeds. This is due to a different stress field which, combined with a weakening of the grain binding energy, leads to a partial grain decohesion on the inner face of the annular pellet. Modification of the grain binding energy is related to the presence of an increasing local population of gas bubbles and metallic precipitates at grain boundaries, as swelling creates intergranular local stresses which also could probably enhance the grain decohesion process. This grain decohesion concerns a 250 to 350 μm depth and shows a narrow cracks network through which released fission gas can flow, temporarily pushing the resident helium gas out. The low conductivity of these gaseous fission products and the numerous gas layers created this way could partly explain the unexpected temperatures measured in high burnup fuels. The purpose of

  13. Experimental observations on electrorefining spent nuclear fuel in molten LiCl-KCl/liquid cadmium system

    Johnson, T. A.; Laug, D. V.; Li, S. X.; Sofu, T.

    1999-01-01

    Argonne National Laboratory (ANL) is currently performing a demonstration program for the Department of Energy (DOE) which processes spent nuclear fuel from the Experimental Breeder Reactor (EBR-II). One of the key steps in this demonstration program is electrorefining of the spent fuel in a molten LiCl-KCl/liquid cadmium system using a pilot scale electrorefiner (Mk-IV ER). This article summarizes experimental observations and engineering aspects for electrorefining spent fuel in the molten LiCl-KCl/liquid cadmium system. It was found that the liquid cadmium pool acted as an intermediate electrode during the electrorefining process in the ER. The cadmium level was gradually decreased due to its high vapor pressure and vaporization rate at the ER operational temperature. The low cadmium level caused the anode assembly momentarily to touch the ER vessel hardware, which generated a periodic current change at the salt/cathode interface and improved uranium recovery efficiency for the process. The primary current distributions calculated by numerical simulations were used in interpreting the experimental results

  14. Observations of crud deposits, corrosion and erosion of BWR and PWR fuel

    Bairiot, H.

    1983-01-01

    The BWR experience is limited to one reactor but the PWR experience covers a wide range of successive generations of power plants (7 in total). The systems are described and their water chemistry briefly commented. Some R and D performed on the effects of the operating regimes (steady state and transients) are summarized. Observations made by pool-side inspections and postirradiation examinations of fuel are outlined concerning water chemistry effects (crud deposits and corrosion) and ''mechanical'' coolant-cladding interaction (chip deposits and baffle jetting). (author)

  15. Defining the Brittle Failure Envelopes of Individual Reaction Zones Observed in CO2-Exposed Wellbore Cement.

    Hangx, Suzanne J T; van der Linden, Arjan; Marcelis, Fons; Liteanu, Emilia

    2016-01-19

    To predict the behavior of the cement sheath after CO2 injection and the potential for leakage pathways, it is key to understand how the mechanical properties of the cement evolves with CO2 exposure time. We performed scratch-hardness tests on hardened samples of class G cement before and after CO2 exposure. The cement was exposed to CO2-rich fluid for one to six months at 65 °C and 8 MPa Ptotal. Detailed SEM-EDX analyses showed reaction zones similar to those previously reported in the literature: (1) an outer-reacted, porous silica-rich zone; (2) a dense, carbonated zone; and (3) a more porous, Ca-depleted inner zone. The quantitative mechanical data (brittle compressive strength and friction coefficient) obtained for each of the zones suggest that the heterogeneity of reacted cement leads to a wide range of brittle strength values in any of the reaction zones, with only a rough dependence on exposure time. However, the data can be used to guide numerical modeling efforts needed to assess the impact of reaction-induced mechanical failure of wellbore cement by coupling sensitivity analysis and mechanical predictions.

  16. Performance limits of coated particle fuel. Part II. Mechanical failure of coated particles due to internal gas pressure and kernel swelling

    Hick, H.; Nabielek, H.; Harrison, T. A.

    1973-10-15

    This report presents a summary of experimental results and their theoretical explanation with regard to the "Pressure Failure" of coated particle fuel. While the experimental results refer mainly to the Dragon Reference Particle as proposed for typical Low Enriched Homogeneous Prismatic Steam Cycle HTR Power Reactors, the theoretical understanding of the phenomena and the mathematical models for their description are not limited to a specific design line.

  17. Observations of indirect filial cannibalism in response to nest failure of Black-crowned Night-Herons (Nycticorax nycticorax)

    Brussee, Brianne E.; Coates, Peter S.; Dwight, Ian; Young, Laura G.

    2017-01-01

    During 2011, four separate instances of indirect filial cannibalism, whereby adults consumed their young that died from unknown causes, were observed using video-monitoring techniques in a nesting colony of Black-crowned Night-Herons (Nycticorax nycticorax) on Alcatraz Island. Though they were not observed actively killing their young, in all four observations adult Black-crowned Night-Herons consumed their young following death (i.e., indirect filial cannibalism). We could not determine cause of chick mortality, but parental neglect was likely a contributing factor in at least two instances. Indirect filial cannibalism is not commonly documented among birds, and understanding how cannibalism contributes to nest failure can help researchers better understand factors that limit nesting populations.

  18. Observations on the treatment of mediastinal masses in Hodgkin's disease emphasizing site of failure

    Ryoo, M.C.; Kagan, A.R.; Wollin, M.; Nussbaum, H.; Chan, P.Y.; Hintz, B.L.; Rao, A.R.; McMahon, J.

    1987-01-01

    Of 244 patients with Hodgkin's disease, 126 (52%) had an abnormal mediastinum. Sixty-four patients were treated with radiation, 36 with radiation and chemotherapy, and 25 with chemotherapy alone as an initial treatment. Twenty of 52 (38%) with stage I or II who received initially radiation alone relapsed, and 70% (14 of 20) of them were salvaged with chemotherapy. Therefore, the ultimate failure rate was 12% (6 of 52). Forty percent (8 of 20) of these patients failed within or at the margin of the radiation portal, and 60% failed predominantly outside of the radiation field. Even though we did not treat the whole lung prophylactically, there was only one true peripheral lung recurrence. Nine of 20 (45%) recurred in more than one site. Of 36 patients treated with combined radiation and chemotherapy, 21 patients had stage I, II, or IIIA disease. Of these, two patients relapsed. Of 86 patients with accessible x-ray films, 30 patients had large masses with a ratio of mass to transverse diameter greater than .33 at the broadest level. Fifty-six patients had small masses. Survival at 96 months in patients with stages I-IIIA with either large or small masses is 94% (p = 0.80). Their relapse-free survival at 96 months is 79% for large masses and 95% for small masses (p = 0.18). The site of relapse is discussed in detail in the text. There were five treatment-related deaths; three patients died of acute myelogenous leukemia. Our data do not support the role of whole-lung prophylactic irradiation or initial combined radiotherapy and chemotherapy in patients with large mediastinal masses

  19. A Puzzle Unsolved: Failure to Observe Different Effects of God and Religion Primes on Intergroup Attitudes.

    Ramsay, Jonathan E; Tong, Eddie M W; Pang, Joyce S; Chowdhury, Avijit

    2016-01-01

    Religious priming has been found to have both positive and negative consequences, and recent research suggests that the activation of God-related and community-related religious cognitions may cause outgroup prosociality and outgroup derogation respectively. The present research sought to examine whether reminders of God and religion have different effects on attitudes towards ingroup and outgroup members. Over two studies, little evidence was found for different effects of these two types of religious primes. In study 1, individuals primed with the words "religion", "God" and a neutral control word evaluated both ingroup and outgroup members similarly, although a marginal tendency towards more negative evaluations of outgroup members by females exposed to religion primes was observed. In study 2, no significant differences in attitudes towards an outgroup member were observed between the God, religion, and neutral priming conditions. Furthermore, the gender effect observed in study 1 did not replicate in this second study. Possible explanations for these null effects are discussed.

  20. Fuel failure detection device

    Katagiri, Masaki.

    1979-01-01

    Purpose: To improve the SN ratio in the detection. Constitution: Improved precipitator method is provided. Scintillation detectors of a same function are provided respectively by each one before and after a gas reservoir for depositing fission products in the cover gas to detecting wires. The outputs from the two detectors (output from the wire not deposited with the fission products and the output from the wire after deposition) are compared to eliminate background noises resulted from not-decayed nucleides. A subtraction circuit is provided for the elimination. Since the background noises of the detecting wire can thus be measured and corrected on every detection, the SN ratio can be increased. (Ikeda, J.)

  1. Sunscreen use and failures--on site observations on a sun-holiday

    Petersen, Bibi; Datta, Pameli; Philipsen, Peter Alshede

    2013-01-01

    With this observation study we aimed to determine how and when sunscreen was used. 20 sun seekers were observed during a one-week sun holiday in Hurghada, Egypt. The sunscreen application thickness was related to part of body, time outdoors, exposure to ultraviolet radiation and to sunburning. Skin...... and the sunscreen application thickness was in average 0.79 mg cm(-)2 giving an approximated effect of SPF3. For different body parts either the total UVR exposure dose or the UVR exposure time and UVR exposure dose before the first sunscreen application were higher for sunburned than non-sunburned skin sites....... In the final model gender, skin type and UVR to skin (adjusted for SPF and sunscreen application thickness) were significant predictors of sunburning. The sunscreen application thickness of 0.79 mg cm(-)2 was less than the 2 mg cm(-2) used for testing SPF. The late start of sunscreen use and improper...

  2. Estimation of failure probability on real structure utilized by earthquake observation data

    Matsubara, Masayoshi

    1995-01-01

    The objective of this report is to propose the procedure which estimates the structural response on a real structure by utilizing earthquake observation data using Neural network system. We apply the neural network system to estimate the ground motion of the site by enormous earthquake data published from Japan Meteorological Agency. The proposed procedure has some possibility to estimate the correlation between earthquake and response adequately. (author)

  3. In situ X-ray observation of semi-solid deformation and failure in Al-Cu alloys

    Phillion, A.B., E-mail: andre.phillion@ubc.ca [School of Engineering, University of British Columbia, 3333 University Way, Kelowna, BC, V1V 1V7 (Canada); Hamilton, R.W.; Fuloria, D.; Leung, A.C.L.; Rockett, P. [Department of Materials, Imperial College London, Prince Consort Road, London SW7 2BP (United Kingdom); Connolley, T. [Diamond Light Source Ltd., Harwell Science and Innovation Campus, Didcot OX11 0DE (United Kingdom); Lee, P.D. [Department of Materials, Imperial College London, Prince Consort Road, London SW7 2BP (United Kingdom)

    2011-02-15

    Semi-solid deformation has been directly observed in an Al-12 wt.% Cu alloy through the combination of real-time synchrotron X-ray radiography and a bespoke high-temperature tensile tester over a range of fraction solid from 0.35 to 0.98. During deformation at low and moderate fraction solids, the X-ray radiographs indicate that there is significant feeding of interdendritic liquid in the region of strain localization prior to crack formation. Furthermore, the measured load required to initiate localized tensile deformation was found to be similar over the range of fraction solid 0.35 to 0.66. At higher fraction solids, the radiographic observations are consistent with classical hot tearing behaviour: limited liquid flow due to low permeability; void nucleation and coalescence; and final failure. Based on these results, a three-stage mechanism for semi-solid failure is proposed which includes the effects of liquid flow and micro-neck formation.

  4. Meteorological Satellites (METSAT) and Earth Observing System (EOS) Advanced Microwave Sounding Unit-A (AMSU-A) Failure Modes and Effects Analysis (FMEA) and Critical Items List (CIL)

    1996-01-01

    This Failure Modes and Effects Analysis (FMEA) is for the Advanced Microwave Sounding Unit-A (AMSU-A) instruments that are being designed and manufactured for the Meteorological Satellites Project (METSAT) and the Earth Observing System (EOS) integrated programs. The FMEA analyzes the design of the METSAT and EOS instruments as they currently exist. This FMEA is intended to identify METSAT and EOS failure modes and their effect on spacecraft-instrument and instrument-component interfaces. The prime objective of this FMEA is to identify potential catastrophic and critical failures so that susceptibility to the failures and their effects can be eliminated from the METSAT/EOS instruments.

  5. Quantifying global fossil-fuel CO2 emissions: from OCO-2 to optimal observing designs

    Ye, X.; Lauvaux, T.; Kort, E. A.; Oda, T.; Feng, S.; Lin, J. C.; Yang, E. G.; Wu, D.; Kuze, A.; Suto, H.; Eldering, A.

    2017-12-01

    Cities house more than half of the world's population and are responsible for more than 70% of the world anthropogenic CO2 emissions. Therefore, quantifications of emissions from major cities, which are only less than a hundred intense emitting spots across the globe, should allow us to monitor changes in global fossil-fuel CO2 emissions, in an independent, objective way. Satellite platforms provide favorable temporal and spatial coverage to collect urban CO2 data to quantify the anthropogenic contributions to the global carbon budget. We present here the optimal observation design for future NASA's OCO-2 and Japanese GOSAT missions, based on real-data (i.e. OCO-2) experiments and Observing System Simulation Experiments (OSSE's) to address different error components in the urban CO2 budget calculation. We identify the major sources of emission uncertainties for various types of cities with different ecosystems and geographical features, such as urban plumes over flat terrains, accumulated enhancements within basins, and complex weather regimes in coastal areas. Atmospheric transport errors were characterized under various meteorological conditions using the Weather Research and Forecasting (WRF) model at 1-km spatial resolution, coupled to the Open-source Data Inventory for Anthropogenic CO2 (ODIAC) emissions. We propose and discuss the optimized urban sampling strategies to address some difficulties from the seasonality in cloud cover and emissions, vegetation density in and around cities, and address the daytime sampling bias using prescribed diurnal cycles. These factors are combined in pseudo data experiments in which we evaluate the relative impact of uncertainties on inverse estimates of CO2 emissions for cities across latitudinal and climatological zones. We propose here several sampling strategies to minimize the uncertainties in target mode for tracking urban fossil-fuel CO2 emissions over the globe for future satellite missions, such as OCO-3 and future

  6. Analysis of terminated TOP accidents in the FTR using the Los Alamos failure model

    Mast, P.K.; Scott, J.H.

    1978-01-01

    A new fuel pin failure model (the Los Alamos Failure Model), based on a linear life fraction rule failure criterion, has been developed and is reported herein. Excellent agreement between calculated and observed failure time and location has been obtained for a number of TOP TREAT tests. Because of the nature of the failure criterion used, the code has also been used to investigate the extent of cladding damage incurred in terminated as well as unterminated TOP transients in the FTR

  7. Effect of assembly error of bipolar plate on the contact pressure distribution and stress failure of membrane electrode assembly in proton exchange membrane fuel cell

    Liu, Dong' an; Peng, Linfa; Lai, Xinmin [State Key Laboratory of Mechanical System and Vibration, Shanghai Jiao Tong University, Shanghai 200240 (China)

    2010-07-01

    In practice, the assembly error of the bipolar plate (BPP) in a PEM fuel cell stack is unavoidable based on the current assembly process. However its effect on the performance of the PEM fuel cell stack is not reported yet. In this study, a methodology based on FEA model, ''least squares-support vector machine (LS-SVM)'' simulation and statistical analysis is developed to investigate the effect of the assembly error of the BPP on the pressure distribution and stress failure of membrane electrode assembly (MEA). At first, a parameterized FEA model of a metallic BPP/MEA assembly is established. Then, the LS-SVM simulation process is conducted based on the FEA model, and datasets for the pressure distribution and Von Mises stress of MEA are obtained, respectively for each assembly error. At last, the effect of the assembly error is obtained by applying the statistical analysis to the LS-SVM results. A regression equation between the stress failure and the assembly error is also built, and the allowed maximum assembly error is calculated based on the equation. The methodology in this study is beneficial to understand the mechanism of the assembly error and can be applied to guide the assembly process for the PEM fuel cell stack. (author)

  8. Post-irradiation examination of a failed PHWR fuel bundle of KAPS-2

    Mishra, Prerna; Unnikrishnan, K.; Viswanathan, U.K.; Shriwastaw, R.S.; Singh, J.L.; Ouseph, P.M.; Alur, V.D.; Singh, H.N.; Anantharaman, S.; Sah, D.N.

    2006-08-01

    Detailed post irradiation examination was carried out on a PHWR fuel bundle irradiated at Kakrapar Atomic Power Station unit 2 (KAPS-2). The fuel bundle had failed early in life at a low burnup of 387 MWd/T. Non destructive and destructive examination was carried out to identify the cause of fuel failure. Visual examination and leak testing indicated failure in two fuel pins of the outer ring of the bundle in the form of axial cracks near the end plug location. Ultrasonic testing of the end cap weld indicated presence of lack of fusion type defect in the two fuel pins. No defect was found in other fuel pins of the bundle. Metallographic examination of fuel sections taken from the crack location in the failed fuel pin showed extensive restructuring of fuel. The centre temperature of the fuel had exceeded 1700 degC at this location in the failed fuel pin, whereas fuel centre temperature in the un-failed fuel pin was only about 1300 degC. Severe fuel clad interaction was observed in the failed fuel pin at and near the location of failure but no such interaction was observed in the un-failed fuel pins. Several incipient cracks originating from the inside surface were found in the cladding near failure location in addition to the main through wall crack. The incipient cracks were filled with interaction products and hydride platelets were present at tip of the cracks. It was concluded from the observations that the primary cause of failure was the presence of a part-wall defect in the end cap weld of the fuel pins. These defects opened up during reactor operation leading to steam ingress into the fuel, which caused high fuel centre temperature and severe fuel-cladding interaction resulting in secondary failures. A more stringent inspection and quality control of end plug weld during fabrication using ultrasonic test has been recommended to avoid such failure. (author)

  9. Atmospheric observations of carbon monoxide and fossil fuel CO2 emissions from East Asia

    Turnbull, Jocelyn C.; Tans, Pieter P.; Lehman, Scott J.

    2011-01-01

    Flask samples from two sites in East Asia, Tae-Ahn Peninsula, Korea (TAP), and Shangdianzi, China (SDZ), were measured for trace gases including CO2, CO and fossil fuel CO2(CO(2)ff, derived from Delta(CO2)-C-14 observations). The five-year TAP record shows high CO(2)ff when local air comes from...... the Korean Peninsula. Most samples, however, reflect air masses from Northeastern China with lower CO(2)ff. Our small set of SDZ samples from winter 2009/2010 have strongly elevated CO(2)ff. Biospheric CO2 contributes substantially to total CO2 variability at both sites, even in winter when non-fossil CO2....../ppm respectively, consistent with recent bottom-up inventory estimates and other observational studies. Locally influenced TAP samples fall into two distinct data sets, ascribed to air sourced from South Korea and North Korea. The South Korea samples have low R-CO:CO2ff of 13 +/- 3 ppb/ppm, slightly higher than...

  10. High Burnup Fuel Behaviour under LOCA Conditions as Observed in Halden Reactor Experiments

    Kolstad, E.; Wiesenack, W.; Oberlander, B.; Tverberg, T.

    2013-01-01

    In the context of assessing the validity of safety criteria for loss of coolant accidents with high burnup fuel, the OECD Halden Reactor Project has implemented an integral in-pile LOCA test series. In this series, fuel fragmentation and relocation, axial gas communication in high burnup rods as affected by gap closure and fuel- clad bonding, and secondary cladding oxidation and hydriding are of major interest. In addition, the data are being used for code validation as well as model development and verification. So far, nine tests with irradiated fuel segments (burnup 40-92 MW.d.kg -1 ) from PWR, BWR and VVER commercial nuclear power plants have been carried out. The in-pile measurements and the PIE results show a good repeatability of the experiments. The paper describes the experimental setup as well as the principal features and main results of these tests. Fuel fragmentation and relocation have occurred to varying degrees in these tests. The paper compares the conditions leading to the presence or absence of fuel fragmentation, e.g., burnup and loss of constraint. Axial gas flow is an important driving force for clad ballooning, fuel relocation and fuel expulsion. The experiments have provided evidence that such gas flow can be impeded in high burnup fuel with a potential impact on the ballooning and fuel dispersal. Although the results of the Halden LOCA tests are, to some extent, amplified by conditions and features deliberately introduced into the test series, the fuel behaviour identified in the Halden tests has an impact on the safety assessment of high burnup fuel and should give rise to improvements of the predictive capabilities of LOCA modelling codes. (author)

  11. The role of genetic nephropathies in the formation of chronic renal failure in children (a clinical observation of a child with acrorenal syndrome

    M.A. Gonchar

    2017-05-01

    Full Text Available In the article, the authors analyze the literature, as well as the results of their own long-term experience in the diagnosis of genetically determined renal diseases accompanied by the development of chronic renal failure. The main causes of diseases, the principles of their diagnosis and treatment were outlined. Clinical observation of a patient with acrorenal syndrome with complicated development of chronic renal failure is given.

  12. Researches of real observation geometry in monitoring fuel-containing materials' subcriticality

    Vysotskij, E.D.; Shevchenko, V.G.; Shevchenko, M.V.

    2002-01-01

    The effectiveness of fuel-containing materials monitoring is discussed in the part related to the feasibilities of researches and realization of optimal geometry (detectors - source) of survey of neutron activity dynamics in nuclearly hazardous areas with clusters of fuel-containing materials concentrated in the premises 305/2

  13. Measurements and observations on microscopic swelling in MX-type fuels

    Ronchi, C.; Ray, I.L.F.; Thiele, H.; Laar, J. van de.

    1978-01-01

    Microscopic swelling has been investigated by electron microscopy in several MX-type fuels, irradiated in fast and thermal neutron flux. The results show that fission gas bubbles in these compounds grow to large sizes if the in-pile fuel temperature rises above a critical value (swelling critical temperature Tsub(C)). A comparison has been made of the swelling rates in fuels of different composition, showing that Tsub(C) increases from carbides to nitrides. In fuels subjected to in-pile restructuring (highly rated) He-bonded pins microscopic swelling is affected by pore and grain boundary migration. The influence of these phenomena on the fuel swelling performance has been discussed

  14. Updating of adventitious fuel pin failure frequency in sodium-cooled fast reactors and probabilistic risk assessment on consequent severe accident in Monju

    Fukano, Yoshitaka; Kurisaka, Kenichi; Nishimura, Masahiro; Naruto, Kenichi

    2015-01-01

    Experimental studies, deterministic approaches and probabilistic risk assessments (PRAs) on local fault (LF) propagation in sodium-cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious-fuel-pin-failures (AFPFs) have been considered to be the most dominant initiators of LFs in these PRAs because of their high frequency of occurrence during reactor operation and possibility of fuel-element-failure-propagation (FEFP). A PRA on FEFP from AFPF (FEFPA) in the Japanese prototype SFR (Monju) was performed in this study based on the state-of-the-art knowledge, reflecting the most recent operation procedures under off-normal conditions. Frequency of occurrence of AFPF in SFRs which was the initiating event of the event tree in this PRA was updated using a variety of methods based on the above-mentioned latest review on experiences of this phenomenon. As a result, the frequency of occurrence of, and the core damage frequency (CDF) from, AFPF in Monju was significantly reduced to a negligible magnitude compared with those in the existing PRAs. It was, therefore concluded that the CDF of FEFPA in Monju could be comprised in that of anticipated transient without scram or protected loss of heat sink events from both the viewpoint of occurrence probability and consequences. (author)

  15. Direct ultimate disposal of spent fuel. Simulation of shaft transport. Tests for removing operating failures (TA 6)

    Filbert, W.; Heda, M.; Neydek, J.

    1994-03-01

    Probable operating failures were analysed and appropriate means for recovery were planned. Based on the analysis of probable failures, three major countermeasures were defined and planned in detail for subsequent demonstration tests: recovery of a fully functioning plateau car, recovery of a plateau car that got stuck, and recovery of a plateau car that ran off the rails. All failures investigated can be repaired by one countermeasure, or a combination of the measures provided for. Trained personnel will be able to restore full service of a loaded POLLUX plateau car that ran off the track and is damaged within approx. 50 minutes. Dose calculations on a conservative basis indicate personal doses resulting from recovery and repair work to be between 200 μSv (20 mrem) and 52 μSv (5.2 mrem). The collective dose is calculated to be approx. 250 μSv (25 mrem). (orig./HP) [de

  16. Fission gas release modelling: developments arising from instrumented fuel assemblies, out-of-pile experiments and microstructural observations

    Leech, N.A.; Smith, M.R.; Pearce, J.H.; Ellis, W.E.; Beatham, N.

    1990-01-01

    This paper reviews the development of fission gas release modelling in thermal reactor fuel (both steady-state and transient) and in particular, illustrates the way in which experimental data have been, and continue to be, the main driving force behind model development. To illustrate this point various aspects of fuel performance are considered: temperature calculation, steady-state and transient fission gas release, grain boundary gas atom capacity and microstructural phenomena. The sources of experimental data discussed include end-of-life fission gas release measurements, instrumented fuel assemblies (e.g. rods with internal pressure transducers, fuel centre thermocouples), swept capsule experiments, out-of-pile annealing experiments and microstructural techniques applied during post-irradiation evaluation. In the case of the latter, the benefit of applying many observation and analysis techniques on the same fuel samples (the approach adopted at NRL Windscale) is emphasized. This illustrates a shift of emphasis in the modelling field from the development of large, complex thermo-mechanical computer codes to the assessment of key experimental data in order to develop and evaluate sub-models which correctly predict the observed behaviour. (author)

  17. Observation on the irradiation behavior of U-Mo alloy dispersion fuel

    Hofman, Gerard L.; Meyer, Mitchell K.; Park, Jong-Man

    2000-01-01

    Initial results from the postirradiation examination of high-density dispersion fuel test RERTR-3 are discussed. The U-Mo alloy fuels in this test were irradiated to 40% U-235 burnup at temperature ranging from 140 0 C to 240 0 C. Temperature has a significant effect on overall swelling of the test plates. The magnitude of the swelling appears acceptable and no unstable irradiation behavior is evident. (author)

  18. BI/TRI-dimensional effects observed in PWR fuel during transient conditions and their numerical simulation

    Linet, B; Hourdequin, N [Departement de Mecanique et Technologie, CEA Centre d` Etudes Nucleaires de Saclay, Gif-sur-Yvette (France)

    1997-08-01

    TOUTATIS is the modular program (both modules 2D and 3D are included) from the METERO project developed by the French Atomic Energy Commission ``CEA``. The model allows the user to calculate the deformations connected to the pellet-clad systems, and hence the Pellet-Cladding Interactions ``PCI`` induced by unilateral contact. Furthermore TOUTATIS provides sufficient versatility to allow the simulation of almost any phenomena, from creep and plasticity to the stress corrosion (residual stresses, dish filling of the pellets from the center, thermo-mechanical feedback) or fuel cracking (3D). The general approach provides a unique capability for understanding different phenomena, some of which remain still unexplained. The first example is related to rod bending, since this phenomenon has been observed in some experimental reactors. Several possible explanations have been put forward, such as flux dipping, buckling or thermohydraulic perturbations. Indeed a spatial parabolic distribution of the flux induces a shift of the isopower area in the pellets, but its effect decreases progressively as the distance from the center of the pellet is increased. So the variations on the clad temperature are just a few degrees and cannot produce the stated rod bending. The second hypothesis was based on a thermohydraulic perturbation. Both chosen configurations (azymutal area/small spot), which induced a thermal perturbation (corroborated by shift of the bubble area), are nevertheless insufficient to bring about the recorded strains. Lastly the calculations performed with the 3D model showed clearly that this rod bending was caused by single buckling induced itself by the immobilization of the rod in experimental channel. 19 figs.

  19. BI/TRI-dimensional effects observed in PWR fuel during transient conditions and their numerical simulation

    Linet, B.; Hourdequin, N.

    1997-01-01

    TOUTATIS is the modular program (both modules 2D and 3D are included) from the METERO project developed by the French Atomic Energy Commission ''CEA''. The model allows the user to calculate the deformations connected to the pellet-clad systems, and hence the Pellet-Cladding Interactions ''PCI'' induced by unilateral contact. Furthermore TOUTATIS provides sufficient versatility to allow the simulation of almost any phenomena, from creep and plasticity to the stress corrosion (residual stresses, dish filling of the pellets from the center, thermo-mechanical feedback) or fuel cracking (3D). The general approach provides a unique capability for understanding different phenomena, some of which remain still unexplained. The first example is related to rod bending, since this phenomenon has been observed in some experimental reactors. Several possible explanations have been put forward, such as flux dipping, buckling or thermohydraulic perturbations. Indeed a spatial parabolic distribution of the flux induces a shift of the isopower area in the pellets, but its effect decreases progressively as the distance from the center of the pellet is increased. So the variations on the clad temperature are just a few degrees and cannot produce the stated rod bending. The second hypothesis was based on a thermohydraulic perturbation. Both chosen configurations (azymutal area/small spot), which induced a thermal perturbation (corroborated by shift of the bubble area), are nevertheless insufficient to bring about the recorded strains. Lastly the calculations performed with the 3D model showed clearly that this rod bending was caused by single buckling induced itself by the immobilization of the rod in experimental channel. 19 figs

  20. System failure, innovation policy and patents: Fuel cells and related hydrogen technology in Norway 1990-2002

    Godoe, Helge; Nygaard, Stian

    2006-01-01

    The empirical focus of this article is technological innovation activities in the emerging field of fuel cells and related hydrogen technology in Norway from 1990 to 2002. In this period, four comparatively large-scale research and development projects and a number of smaller projects aimed at development of fuel cells technology were undertaken, resulting in many inventions that were subsequently patented. Although this creativity may be considered an indication of success, only one of the projects became successful in an innovation perspective. All the large projects were initiated and funded for divergent political and economic reasons. An important reason in the late 1980s was the prospect of using Norway's abundant supply of natural gas in fuel cells for electric power generation. The large R and D projects that attempted to develop fuel cells based on natural gas as energy source failed. In contrast, the successful project was undertaken by military R and D, i.e. in a different system of innovation than the projects that failed. Analysis of these cases points to the importance of a systemic approach to innovations-and to policy making. One challenge for policy makers is to decide how they should promote this development which is crucial for the vision of a future 'Hydrogen Economy', i.e. what kind of policy incentives should be introduced to spur efficiency in technological development and diffusion. Theoretically, many options are available; however, understanding the innovation dynamics in this sector is fundamental for making choices. In this article, focus will be set on policy aspects using an innovation systemic approach to analyze development of fuel cells and related hydrogen technology in Norway

  1. Interpretation of the CABRI-RAFT LTX test up to pin failure based on detailed data evaluation and PARAS-2S code analysis

    Fukano, Yoshitaka; Sato, Ikken

    2001-09-01

    The CABRI-RAFT LTX test aims at a study on the fuel-pin-failure mechanism, in-pin fuel motion and post-failure fuel relocation with an annular fuel pin which was pre-irradiated up to peak burn-up of 6.4 at%. The transient test conditions similar to those of the LT4 test were selected in the LTX test using the same type of fuel pin, allowing an effective direct comparison between the two tests. In contrast to the LT4 test which showed a large PCMI-mitigation potential of the annular fuel-pin design, early pin failure occurred in the LTX test when fuel does not seem to have molten. In order to clarify the fuel pin failure mechanism, interpretation of the LTX test up to pin failure is performed in this study, through an experimental data evaluation and a PAPAS-2S-code analysis. The PAPAS-2S code simulates reasonably the fuel thermal conditions such as transient fuel-pin heat-up and fuel melting. The present detailed data evaluation shows that the earlier cladding failure compared with the LT4 test is mainly attributed to the local cladding heat-up. Under the high-temperature condition, plenum gas pressure has a certain potential to explain the observed failure. Fuel swelling-induced PCMI does not seem significant in the LTX test and it may have contributed to the early pin failure only to a limited extent, if any. (author)

  2. UK experience on fuel and cladding interaction in oxide fuels

    Batey, W [Dounreay Experimental Reactor Establishment, Thurso, Caithness (United Kingdom); Findlay, J R [AERE, Harwell, Didcot, Oxon (United Kingdom)

    1977-04-01

    The occurrence of fuel cladding interactions in fast reactor fuels has been observed in UK irradiations over a period of years. Chemical incompatibility between fuel and clad represents a potential source of failure and has, on this account, been studied using a variety of techniques. The principal fuel of interest to the UK for fast reactor application is mixed uranium plutonium oxide clad in stainless steel and it is in this field that the majority of work has been concentrated. Some consideration has been given to carbide fuels, because of their application as an advanced fuel. This experience is described in the accompanying paper. Several complementary initiatives have been followed to investigate the interactions in oxide fuel. The principal source of experimental information is from the experimental fuel irradiation programme in the Dounreay Fast Reactor (DFR). Supporting information has been obtained from irradiation programmes in Materials Testing Reactors (MTR). Conditions approaching those in a fast reactor are obtained and the effects of specific variables have been examined in specifically designed experiments. Out-of-reactor experiments have been used to determine the limits of fuel and cladding compatibility and also to give indications of corrosion The observations from all experiments have been examined in the light of thermo-dynamic predictions of fuel behaviour to assess the relative significance of various observations and operating conditions. An experimental programme to control and limit the interactions in oxide fuel is being followed.

  3. UK experience on fuel and cladding interaction in oxide fuels

    Batey, W.; Findlay, J.R.

    1977-01-01

    The occurrence of fuel cladding interactions in fast reactor fuels has been observed in UK irradiations over a period of years. Chemical incompatibility between fuel and clad represents a potential source of failure and has, on this account, been studied using a variety of techniques. The principal fuel of interest to the UK for fast reactor application is mixed uranium plutonium oxide clad in stainless steel and it is in this field that the majority of work has been concentrated. Some consideration has been given to carbide fuels, because of their application as an advanced fuel. This experience is described in the accompanying paper. Several complementary initiatives have been followed to investigate the interactions in oxide fuel. The principal source of experimental information is from the experimental fuel irradiation programme in the Dounreay Fast Reactor (DFR). Supporting information has been obtained from irradiation programmes in Materials Testing Reactors (MTR). Conditions approaching those in a fast reactor are obtained and the effects of specific variables have been examined in specifically designed experiments. Out-of-reactor experiments have been used to determine the limits of fuel and cladding compatibility and also to give indications of corrosion The observations from all experiments have been examined in the light of thermo-dynamic predictions of fuel behaviour to assess the relative significance of various observations and operating conditions. An experimental programme to control and limit the interactions in oxide fuel is being followed

  4. Observations of in-reactor strain for fueled and unfueled FTR cladding

    Gilbert, E.R.; Makenas, B.J.; Wilson, D.R.

    1979-01-01

    It has been demonstrated that equations derived from in-reactor creep and swelling in unfueled pressurized tubes of 20% CW AISI 316 stainless steel can be used to predict strains in prototypic FTR mixed-oxide (UO 2 --PuO 2 ) fuel pins. For fast neutron fluences below 6 x 10 22 n/cm 2 the strains were small (less than one percent) and good agreement was found (within 0.1 percent diametral strain) between the equations and the fuel pin strains. This paper describes an extension of the earlier study to fast neutron fluences up to 11 x 10 22 n/cm 2

  5. Case study on unusual occurrence of wire rope failure of 15/5t EOT crane at one of the plant, Nuclear Fuel Complex

    Siddu, P.V.; Subhakar, A.; Devender Rao, D.; Srichandan, T.T.; Harilingam, S.; Mahesh Kumar, V.V.; Kiran Kumar, I.; Thakur, Sudhir

    2016-01-01

    At NFC, Production of fuel bundles and other structural involves a large number of material handling activities for transfer of materials from one plant to another. One of the widely used materials handling equipment at NFC is the Electric Overhead Travelling (EOT) Crane. These EOT cranes will be tested yearly as per the statutory requirements in the presence of the competent person. Also, regular inspection and maintenance activities will be carried by the plant maintenance crew. Even with preventive maintenance and monitoring, there could be accidents due to unsafe conditions, unsafe acts or both. However, all the unusual occurrence, incidents and accidents can be prevented. This paper deals about the unusual occurrence of wire rope failure of 15/5T EOT crane at one of the plant in NFC; recommendations given by the Safety Committee, AERB and action taken by the plant to prevent reoccurrence of unusual occurrence. (author)

  6. Spent fuel's behavior under dynamic drip tests

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

    1995-01-01

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

  7. Plasma barodiffusion in inertial-confinement-fusion implosions: application to observed yield anomalies in thermonuclear fuel mixtures.

    Amendt, Peter; Landen, O L; Robey, H F; Li, C K; Petrasso, R D

    2010-09-10

    The observation of large, self-generated electric fields (≥10(9)  V/m) in imploding capsules using proton radiography has been reported [C. K. Li, Phys. Rev. Lett. 100, 225001 (2008)]. A model of pressure gradient-driven diffusion in a plasma with self-generated electric fields is developed and applied to reported neutron yield deficits for equimolar D3He [J. R. Rygg, Phys. Plasmas 13, 052702 (2006)] and (DT)3He [H. W. Herrmann, Phys. Plasmas 16, 056312 (2009)] fuel mixtures and Ar-doped deuterium fuels [J. D. Lindl, Phys. Plasmas 11, 339 (2004)]. The observed anomalies are explained as a mild loss of deuterium nuclei near capsule center arising from shock-driven diffusion in the high-field limit.

  8. Mechanical behaviour and failure of fuel cladding zirconium alloys in nuclear power plants under accidental RIA-type situation

    Doan, D.T.

    2009-01-01

    In French Nuclear Pressurized Water Reactors (PWRs), most of structural parts of the fuel assembly consist of zirconium alloy tubes and plates. Optimizing the management of fuel in nuclear power plants led to the increase in the duration of fuel cycles and power. The use of high fuel burnups requires drastic changes in the rules for reactor design in the nuclear safety. The evaluation of nuclear reactors in accident situations is based on reference accident scenarios. One of these hypothetical accidents, examined in this study, is the 'Reactivity Initiated Accident'. In order to assess the structural integrity of these parts it is necessary to characterize both the plastic flow and fracture behaviour of the materials at various stages of the life cycle, (i.e. at increasing levels of hydriding, irradiation, oxidation or thermal mechanical loading). The purpose of this work is to provide experimental data and to develop a model of the thermo-mechanical behaviour and to propose a design analysis method in the case of non-irradiated clads, in RIA-type situations. Mechanical tests were conducted on Cold-Worked-Stress-Relieved and on Recrystallized Zircaloy-4 sheets using various kinds of samples including smooth and notched tensile specimens and small punch tests. Temperature was set to 25, 250 and 600 C with hydrogen contents between 0 and 1000 ppm. The model is based on a simplified description of a Zircaloy polycrystal in which scalar isotropic ductile damage including void nucleation and growth is added. The model is also physically based to easily transfer parameters determined for one material state to another (e.g. transfer between sheet and tube or between different levels of irradiation). The model was implemented in the Finite Element software Zebulon using either an explicit or an implicit time integration scheme. Uniaxial tension tests were used to tune the model parameters for both materials, considering various values of temperature and hydrogen levels

  9. Driving force of PCMI failure under reactivity initiated accident conditions and influence of hydrogen embrittlement on failure limit

    Tomiyasu, Kunihiko; Sugiyama, Tomoyuki; Nakamura, Takehiko; Fuketa, Toyoshi

    2005-09-01

    In order to clarify the driving force of PCMI (Pellet/Cladding Mechanical Interaction) failure on high burnup fuels and to investigate the influence of hydrogen embrittlement on failure limit under RIA (Reactivity Initiated Accident) conditions, RIA-simulation experiments were performed on fresh fuel rods in the NSRR (Nuclear Safety Research Reactor). The driving force of PCMI was restricted only to thermal expansion of pellet by using fresh UO 2 pellets. Fresh claddings were pre-hydrided to simulate hydrogen absorption of high burnup fuel rods. In seven experiments out of fourteen, test rods resulted in PCMI failure, which has been observed in the NSRR tests on high burnup PWR fuels, in terms of the transient behavior and the fracture configuration. This indicates that the driving force of PCMI failure is sufficiently explained with thermal expansion of pellet and a contribution of fission gas on it is small. A large number of incipient cracks were generated in the outer surface of the cladding even on non-failed fuel rods, and they stopped at the boundary between hydride rim, which was a hydride layer localized in the periphery of the cladding, and metallic layer. It suggests that the integrity of the metallic layer except for the hydride rim has particular importance for failure limit. Fuel enthalpy at failure correlates with the thickness of hydride rim, and tends to decrease with thicker hydride layer. (author)

  10. Global Partitioning of NOx Sources Using Satellite Observations: Relative Roles of Fossil Fuel Combustion, Biomass Burning and Soil Emissions

    Jaegle, Lyatt; Steinberger, Linda; Martin, Randall V.; Chance, Kelly

    2005-01-01

    This document contains the following abstract for the paper "Global partitioning of NOx sources using satellite observations: Relative roles of fossil fuel combustion, biomass burning and soil emissions." Satellite observations have been used to provide important new information about emissions of nitrogen oxides. Nitrogen oxides (NOx) are significant in atmospheric chemistry, having a role in ozone air pollution, acid deposition and climate change. We know that human activities have led to a three- to six-fold increase in NOx emissions since pre-industrial times, and that there are three main surface sources of NOx: fuel combustion, large-scale fires, and microbial soil processes. How each of these sources contributes to the total NOx emissions is subject to some doubt, however. The problem is that current NOx emission inventories rely on bottom-up approaches, compiling large quantities of statistical information from diverse sources such as fuel and land use, agricultural data, and estimates of burned areas. This results in inherently large uncertainties. To overcome this, Lyatt Jaegle and colleagues from the University of Washington, USA, used new satellite observations from the Global Ozone Monitoring Experiment (GOME) instrument. As the spatial and seasonal distribution of each of the sources of NOx can be clearly mapped from space, the team could provide independent topdown constraints on the individual strengths of NOx sources, and thus help resolve discrepancies in existing inventories. Jaegle's analysis of the satellite observations, presented at the recent Faraday Discussion on "Atmospheric Chemistry", shows that fuel combustion dominates emissions at northern mid-latitudes, while fires are a significant source in the Tropics. Additionally, she discovered a larger than expected role for soil emissions, especially over agricultural regions with heavy fertilizer use. Additional information is included in the original extended abstract.

  11. Association of ambient particulate matter with heart failure incidence and all-cause readmissions in Tasmania: an observational study.

    Huynh, Quan L; Blizzard, Christopher Leigh; Marwick, Thomas H; Negishi, Kazuaki

    2018-05-10

    We sought to investigate the relationship between air quality and heart failure (HF) incidence and rehospitalisation to elucidate whether there is a threshold in this relationship and whether this relationship differs for HF incidence and rehospitalisation. This retrospective observational study was performed in an Australian state-wide setting, where air pollution is mainly associated with wood-burning for winter heating. Data included all 1246 patients with a first-ever HF hospitalisation and their 3011 subsequent all-cause readmissions during 2009-2012. Daily particulate matter <2.5 µm (PM 2.5 ), temperature, relative humidity and influenza infection were recorded. Poisson regression was used, with adjustment for time trend, public and school holiday and day of week. Tasmania has excellent air quality (median PM 2.5 =2.9 µg/m 3 (IQR: 1.8-6.0)). Greater HF incidences and readmissions occurred in winter than in other seasons (p<0.001). PM 2.5 was detrimentally associated with HF incidence (risk ratio (RR)=1.29 (1.15-1.42)) and weakly so with readmission (RR=1.07 (1.02-1.17)), with 1 day time lag. In multivariable analyses, PM 2.5 significantly predicted HF incidence (RR=1.12 (1.01-1.24)) but not readmission (RR=0.96 (0.89-1.04)). HF incidence was similarly low when PM <4 µg/m 3 and only started to rise when PM 2.5 ≥4 µg/m 3 . Stratified analyses showed that PM 2.5 was associated with readmissions among patients not taking beta-blockers but not among those taking beta-blockers (p interaction =0.011). PM 2.5 predicted HF incidence, independent of other environmental factors. A possible threshold of PM 2.5 =4 µg/m 3 is far below the daily Australian national standard of 25 µg/m 3 . Our data suggest that beta-blockers might play a role in preventing adverse association between air pollution and patients with HF. © Article author(s) (or their employer(s) unless otherwise stated in the text of the article) 2018. All rights reserved. No commercial

  12. Experience with nuclear fuel utilization in Bulgaria

    Harizanov, Y [Committee on the Use of Atomic Energy for Peaceful Purposes, Sofia (Bulgaria)

    1997-12-01

    The presentation on experience with nuclear fuel utilization in Bulgaria briefly reviews the situation with nuclear energy in Bulgaria and then discusses nuclear fuel performance (amount of fuel loaded, type of fuel, burnup, fuel failures, assemblies deformation). 2 tabs.

  13. HTGR Fuel performance basis

    Shamasundar, B.I.; Stansfield, O.M.; Jensen, D.D.

    1982-05-01

    The safety characteristics of the high-temperature gas-cooled reactor (HTGR) during normal and accident conditions are determined in part by HTGR fuel performance. During normal operation, less than 0.1% fuel failure occurs, primarily from defective particles. This low fuel failure fraction limits circulating activity to acceptable levels. During severe accidents, the radiological consequence is influenced by high-temperature fuel particle behavior. An empirical fuel failure model, supported by recent experimental data, is presented. The onset of significant fuel particle failure occurs at temperatures in excess of 1600 0 C, and complete fuel failure occurs at 2660 0 C. This indicates that the fuel is more retentive at higher temperatures than previously assumed. The more retentive nature of the fuel coupled with the high thermal capacitance of the core results in slow release of fission products from the core during severe accidents

  14. Observation of methanol behavior in fuel cells in situ by NMR spectroscopy.

    Han, Oc Hee; Han, Kee Sung; Shin, Chang Woo; Lee, Juhee; Kim, Seong-Soo; Um, Myung Sup; Joh, Han-Ik; Kim, Soo-Kil; Ha, Heung Yong

    2012-04-16

    The chemical conversion of methanol in direct methanol fuel cells was followed in situ by NMR spectroscopy. Comparing data of the methanol oxidation on Pt and PtRu anode catalysts allowed the role of Ru in both Faradaic and non-Faradaic reactions to be investigated. The spatial distributions of chemicals could also be determined. (Picture: T1-T4=inlet and outlet tubes.). Copyright © 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  15. Severe fuel damage in steam and helium environments observed in in-reactor experiments

    Saito, S.; Shiozawa, S.

    1984-01-01

    The bahavior of severe fuel damages has been studied in gaseous environments simulating core uncovery accidents in the in-reactor experiments utilizing the NSRR. Two types of cladding relocation modes, azimuthal flow and melt-down, were revealed through the parametric experiments. The azimuthal flow was evident in an oxidizing environment in case of no oxide film break. The melt-down can be categorized into flow-down and move-down, according to the velocity of the melt-down. Cinematographies showed that the flow-down was very fast as water flows down while the move-down appeared to be much slower. The flow-down was possible in an unoxidizing environment, whereas the move-down of molten cladding occured through a crack induced in an oxide film in an oxidizing environment. The criterion of the relocation modes was developed as a function of peak cladding temperature and oxidation condition. It was also found that neither immediate quench nor fuel fracture occurred upon flooding when cladding temperature was about 1800 0 C at water injection. The external mechanical force is needed for fuel fracture. (orig.)

  16. Chinese Herbs Containing Aristolochic Acid Associated with Renal Failure and Urothelial Carcinoma: A Review from Epidemiologic Observations to Causal Inference

    Yang, Hsiao-Yu; Chen, Pau-Chung; Wang, Jung-Der

    2014-01-01

    Herbal remedies containing aristolochic acid (AA) have been designated to be a strong carcinogen. This review summarizes major epidemiologic evidence to argue for the causal association between AA exposure and urothelial carcinoma as well as nephropathy. The exposure scenarios include the following: Belgian women taking slimming pills containing single material Guang Fang Ji, consumptions of mixtures of Chinese herbal products in the general population and patients with chronic renal failure ...

  17. Data Assimilation of Dead Fuel Moisture Observations from Remote automated Weather Stations

    Vejmelka, Martin; Kochanski, A.; Mandel, Jan

    2016-01-01

    Roč. 25, č. 5 (2016), s. 558-568 ISSN 1049-8001 R&D Projects: GA ČR GA13-34856S Grant - others:National Science Foundation(US) AGS-0835579 and DMS-1216481; NASA (US) NNX12AQ85G and NNX13AH9G. Institutional support: RVO:67985807 Keywords : data assimilation * dead fuel moisture * equilibrium * Kalman filter * remote automated weather stations * time lag model * trend surface model Subject RIV: DG - Athmosphere Sciences, Meteorology Impact factor: 2.748, year: 2016

  18. Nuclear fuel preheating system

    Andrea, C.

    1975-01-01

    A nuclear reactor new fuel handling system which conveys new fuel from a fuel preparation room into the reactor containment boundary is described. The handling system is provided with a fuel preheating station which is adaptd to heat the new fuel to reactor refueling temperatures in such a way that the fuel is heated from the top down so that fuel element cladding failure due to thermal expansions is avoided. (U.S.)

  19. Fuel performance experience

    Sofer, G.A.

    1986-01-01

    The history of LWR fuel supply has been characterized by a wide range of design developments and fuel cycle cost improvements. Exxon Nuclear Company, Inc. has pursued an aggressive fuel research and development program aimed at improved fuel performance. Exxon Nuclear has introduced many design innovations which have improved fuel cycle economics and operating flexibility while fuel failures remain at very low levels. The removable upper tie plate feature of Exxon Nuclear assemblies has helped accelerate this development, enabling repeated inspections during successive plant outages. Also, this design feature has made it possible to repair damaged fuel assemblies during refueling outages, thereby minimizing the economic impact of fuel failure from all causes

  20. Observation of the failure mechanism of brick masonry doublets with cement and lime mortars by microfocus X-ray computed tomography

    Hendrickx, Roel; Bruyninckx, Katrien; Schueremans, Luc; Kerckhofs, Greet; Verstrynge, Els; Wevers, Martine; Van Balen, Koenraad

    2010-01-01

    The nature of the failure mechanism of masonry under compression depends on the properties of the brick and mortar. It is well-known that the ratio of stiffness of both materials has an important effect. Furthermore the pattern of crack development and propagation and the occurrence of local compaction of soft mortars have been the subject of some study, but remained difficult to observe. This study aims at the visualisation of these phenomena by using a hydraulic press inside a microfocus X-...

  1. A New Family of Nonlinear Observers for SI Engine Air/Fuel Ratio Control

    Jensen, P. B.; Olsen, M. B.; Poulsen, J.

    1997-01-01

    The paper treats a newly developed set of nonlinear observers for advanced spark ignition engine control.......The paper treats a newly developed set of nonlinear observers for advanced spark ignition engine control....

  2. Chinese Herbs Containing Aristolochic Acid Associated with Renal Failure and Urothelial Carcinoma: A Review from Epidemiologic Observations to Causal Inference

    Hsiao-Yu Yang

    2014-01-01

    Full Text Available Herbal remedies containing aristolochic acid (AA have been designated to be a strong carcinogen. This review summarizes major epidemiologic evidence to argue for the causal association between AA exposure and urothelial carcinoma as well as nephropathy. The exposure scenarios include the following: Belgian women taking slimming pills containing single material Guang Fang Ji, consumptions of mixtures of Chinese herbal products in the general population and patients with chronic renal failure in Taiwan, occupational exposure in Chinese herbalists, and food contamination in farming villages in valleys of the Danube River. Such an association is corroborated by detecting specific DNA adducts in the tumor tissue removed from affected patients. Preventive actions of banning such use and education to the healthcare professionals and public are necessary for the safety of herbal remedies.

  3. Chinese herbs containing aristolochic acid associated with renal failure and urothelial carcinoma: a review from epidemiologic observations to causal inference.

    Yang, Hsiao-Yu; Chen, Pau-Chung; Wang, Jung-Der

    2014-01-01

    Herbal remedies containing aristolochic acid (AA) have been designated to be a strong carcinogen. This review summarizes major epidemiologic evidence to argue for the causal association between AA exposure and urothelial carcinoma as well as nephropathy. The exposure scenarios include the following: Belgian women taking slimming pills containing single material Guang Fang Ji, consumptions of mixtures of Chinese herbal products in the general population and patients with chronic renal failure in Taiwan, occupational exposure in Chinese herbalists, and food contamination in farming villages in valleys of the Danube River. Such an association is corroborated by detecting specific DNA adducts in the tumor tissue removed from affected patients. Preventive actions of banning such use and education to the healthcare professionals and public are necessary for the safety of herbal remedies.

  4. Planar LIF observation of unburned fuel escaping the upper ring-land crevice in an SI engine

    Green, R.M.; Cloutman, L.D.

    1997-01-01

    PLIF has been used to observe the in-cylinder transport of unburned fuel that, while trapped in the ring-land and ring-groove crevices, survives combustion in the propagating flame. Away from the top-ring gap, we detect a wall-jet comprised of unburned charge exiting the top ring-land crevice opening. At the location of the top-ring gap, we observed unburned fuel lying in the cool boundary layer along the cylinder wall during the later stages of the expansion stroke. This layer is scraped into the roll-up vortex during the exhaust stroke. These data lead us to conclude that away from the end gap, unburned, high pressure charge, trapped between the two compression rings escapes as a wall jet after ring-reversal near the bottom center. Conversely, at the ring gap, when the cylinder pressure drops below the pressure between the compression rings, the trapped charge escapes through the gap and forms a thin layer on the cylinder wall.

  5. Nuclear fuels

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

    2009-01-01

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO 2 pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO 2 and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under irradiation

  6. Nuclear fuels

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

    2009-07-01

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO{sub 2} pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO{sub 2} and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under

  7. KMRR fuel design

    Son, D.S.; Sim, B.S.; Kim, T.R.; Hwang, W.; Kim, B.G.; Ku, Y.H.; Lee, C.B.; Lim, I.C.

    1992-06-01

    KMRR fuel rod design criteria on fuel swelling, blistering and oxide spallation have been reexamined. Fuel centerline temperature limit of 250deg C in normal operation condition and fuel swelling limit of 12 % at the end of life have been proposed to prevent fuel failure due to excessive fuel swelling. Fuel temperature limit of 485deg C has been proposed to exclude the possibility of fuel failures during transients or under accident condition. Further analyses are needed to decide the fuel cladding temperature limit to preclude the oxide spallation. Design changes in fuel assembly structure and their effects on related systems have been reviewed from a structural integrity viewpoint. The remained works in fuel mechanical design area have been identified and further efforts of fuel design group will be focused on these aspects. (Author)

  8. BISON Investigation of the Effect of the Fuel- Cladding Contact Irregularities on the Peak Cladding Temperature and FCCI Observed in AFC-3A Rodlet 4

    Medvedev, Pavel G. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    The primary objective of this report is to document results of BISON analyses supporting Fuel Cycle Research and Development (FCRD) activities. Specifically, the present report seeks to provide explanation for the microstructural features observed during post irradiation examination of the helium-bonded annular U-10Zr fuel irradiated during the AFC-3A experiment. Post irradiation examination of the AFC-3A rodlet revealed microstructural features indicative of the fuel-cladding chemical interaction (FCCI) at the fuel-cladding interface. Presence of large voids was also observed in the same locations. BISON analyses were performed to examine stress and temperature profiles and to investigate possible correlation between the voids and FCCI. It was found that presence of the large voids lead to a formation of circumferential temperature gradients in the fuel that may have redirected migrating lanthanides to the locations where fuel and cladding are in contact. Resulting localized increase of lanthanide concentration is expected to accelerate FCCI. The results of this work provide important guidance to the post irradiation examination studies. Specifically, the hypothesis of lanthanides being redirected from the voids to the locations where the fuel and the cladding are in contact should be verified by conducting quantitative electron microscopy or Electron Probe Micro-Analyzer (EPMA). The results also highlight the need for computer models capable of simulating lanthanide diffusion in metallic fuel and establish a basis for validation of such models.

  9. Prognosis and serum creatinine levels in acute renal failure at the time of nephrology consultation: an observational cohort study

    de Irala Jokin

    2007-09-01

    Full Text Available Abstract Background The aim of this study is to evaluate the association between acute serum creatinine changes in acute renal failure (ARF, before specialized treatment begins, and in-hospital mortality, recovery of renal function, and overall mortality at 6 months, on an equal degree of ARF severity, using the RIFLE criteria, and comorbid illnesses. Methods Prospective cohort study of 1008 consecutive patients who had been diagnosed as having ARF, and had been admitted in an university-affiliated hospital over 10 years. Demographic, clinical information and outcomes were measured. After that, 646 patients who had presented enough increment in serum creatinine to qualify for the RIFLE criteria were included for subsequent analysis. The population was divided into two groups using the median serum creatinine change (101% as the cut-off value. Multivariate non-conditional logistic and linear regression models were used. Results A ≥ 101% increment of creatinine respect to its baseline before nephrology consultation was associated with significant increase of in-hospital mortality (35.6% vs. 22.6%, p Conclusion In this cohort, patients who had presented an increment in serum level of creatinine of ≥ 101% with respect to basal values, at the time of nephrology consultation, had increased mortality rates and were discharged from hospital with a more deteriorated renal function than those with similar Liano scoring and the same RIFLE classes, but with a

  10. Identifying the European fossil fuel plumes in the atmosphere over the Northeast Atlantic Region through isotopic observations and numerical modelling

    Geels, C.; Christensen, J.H.; Hansen, A.W.

    2006-01-01

    Atmospheric transport, C-14. fossil fuel CO_2, numerical modeling, the north East Atlantic Region Udgivelsesdato: 18 August......Atmospheric transport, C-14. fossil fuel CO_2, numerical modeling, the north East Atlantic Region Udgivelsesdato: 18 August...

  11. Experimental observations to the electrical field for electrorefining of spent nuclear fuel in the Mark-IV electrorefiner

    Li, S. X.

    1998-01-01

    Experimental results from the pilot scale electrorefiner (Mark-IV ER) treating spent nuclear fuel are reported in this article. The electrorefining processes were carried out in a LiCl-KCl-UCl 3 electrolyte. It has been noted that spool of molten cadmium below the electrolyte plays an important role in the electrorefining operations. In addition, formations of electrical shorting path between anode baskets and the electrorefiner vessel were observed, which lessened the uranium dissolution process from anode baskets, however appeared to improve the morphology of cathode deposit. The FIDAP simulation code was used to calculate the electrical potential field distributions and the potential gradient near the cathode. The effect of the electrical shorting between anode baskets and electrorefiner vessel on the morphology of cathode products is discussed

  12. Observation of ionomer in catalyst ink of polymer electrolyte fuel cell using cryogenic transmission electron microscopy

    Takahashi, Shinichi; Shimanuki, Junichi; Mashio, Tetsuya; Ohma, Atsushi; Tohma, Hajime; Ishihara, Ayumi; Ito, Yoshiko; Nishino, Yuri; Miyazawa, Atsuo

    2017-01-01

    Optimizing the catalyst layer structure is one of the key issues for improving performance despite lower platinum loading. The catalyst ink, consisting of platinum-loaded carbon particles and ionomer dispersed in an aqueous solvent, is a key factor for controlling the structure of the catalyst layer because the catalyst layer is prepared in a wet coating process. For that purpose, we visualized the nanostructure of the ionomer in the catalyst ink by cryogenic electron microscopy, especially cryogenic transmission electron microscopy (cryo-TEM). By cryo-TEM, it was revealed that ionomer molecules formed rod-like aggregates macro-homogeneously in the solvent, and a similar morphology was observed in a carbon-particle-containing solvent. In contrast, ionomer aggregates in the catalyst ink containing platinum nanoparticles loaded on carbon particles were denser in the vicinity of the platinum-loaded carbon particles. That can be attributed to strong interaction between platinum nanoparticles and sulfonic acid groups in the ionomer. It also implies that a good understanding of ionomer morphology in the catalyst ink can play an important role in controlling the catalyst layer microstructure for reducing platinum loading.

  13. Fuel performance update at Japanese BWRs

    Otsuka, Yasuyuki; Abe, Moriyasu

    2008-01-01

    Fuel Performance of Japanese BWRs has shown excellency with a very low fuel failure rate over the past couple of decades. In order to eliminate debris-fretting failures, which is considered as a dominant cause, efforts have been made to work out strict FME program and introduce lower tie plate with debris filters. Regarding a measure to detect failures without delay, an online off-gas monitor is installed, and there occurred no significant secondary failures leading to unnecessary plant shutdowns thanks to the monitor. Current standard fuel designs are 9x9 lattice configuration with the design maximum burnup set at 55 GWd/t and the bundle average discharge burnup at 45 GWd/t. Satisfactory performance of many fuels has gained so far to the burnup level. However recent data show an increase in hydrogen pickup of fuel claddings and spacers in the region of high burnup or high irradiation period, and this is one of the issues that should be addressed to maintain and improve fuel reliability for the reasons that hydrogen concentration may affect ductility of zirconium alloy. Furthermore, a recent study showed abnormally high hydrogen concentration in fuel cladding. Comprehensive root cause analyses were made, which include a poolside measurement for a number of high burnup fuels and development of new technique for tenacious crud removal, with the result that it is confirmed that a unique combination of cladding material, irradiation period and environment caused the phenomenon and it is not expected to occur again. New iron enhanced zirconium alloys, which have been investigated in for a long time, are expected to have greater resistance to hydrogen pickup, and introduction of the alloys is considered as the future materials. Regarding extensive inspections of fuel assemblies and channel boxes at Kashiwazaki-Kariwa NPS, no abnormality due to Niigata-ken Chuetsu-oki earthquake was observed for all plants. This paper deals with the current fuel performance at Japanese

  14. Nuclear fuel quality assurance

    1976-01-01

    the full application of the quality assurance concept in the purchase of fuel and fuel manufacturing services will depend to a large extent on the availability of fuel specification data. On the part of fuel purchasers, there is an obvious interest in getting as many details of fuel specification as possible in order to be able to establish a proper level of control over the quality of their purchases. On the other hand, if such specifications are set up in advance by the purchasers, there are often complaints by the manufacturers that the specifications were set up without proper regard for the latest technical information on fuel performance and for the realities of manufacturing processes and technical capabilities. This problem may be resolved when fuel design activities are properly meshed with a full quality assurance system. Discussions during the seminar showed that the operation of acceptable quality assurance systems is a well-established practice at most of the fuel manufacturers. The fuel purchaser may monitor such a system through quality assurance programme auditing as agreed to the individual vendor-purchaser contracts. In this way confidence may be obtained in the quality of the purchased product. However, it is considered that the further improvement of the relations between fuel manufacturers and purchasers could be achieved through the following actions undertaken at the international level: (1) standardization of fuel specifications and testing procedures; (2) dissemination of information on fuel specifications and their connections with observed fuel failure rate; (3) Establishment of a standardized quality assurance programme for fuel fabrication; (4) establishment of a central information service to assist utility groups in preparing documents and procedures to be used in quality assurance activities

  15. Experience in the manufacture and performance of CANDU fuel for KANUPP

    Salim, M.; Ahmed, I.; Butt, P.

    1995-01-01

    Karachi Nuclear Power Plant (KANUPP) a 137 MWe CANDU unit is In operation since 1971. Initially, it was fueled with Canadian fuel bundles. In July 1980 Pakistani manufactured fuel was introduced in the reactor core, irradiated to a burnup of about 7500 MWd-teU -1 and successfully discharged in May 1984. The core was progressively fuelled with Pakistani fuel and in August 1990 the reactor core contained all Pakistani made fuel. As of the present, 3 core equivalent Pakistani fuel bundles have been successfully discharged at an average bumup of 6500 MWd-teU -1 . with a maximum burnup of ∼ 10,200 MWd-teU -1 . No fuel failure of Pakistani bundles has been observed so far. This paper presents the indigenous efforts towards manufacture and operational aspects of KANUPP fuel and compares its behaviour with that of Canadian supplied fuel. The Pakistani fuel has performed well and is as good as the Canadian fuel. (author)

  16. Evaluation of the effects of chronic biomass fuel smoke exposure on peripheral endothelial functions: an observational study.

    Buturak, Ali; Genç, Ahmet; Ulus, Ozden Sıla; Duygu, Egemen; Okmen, Arda Sanlı; Uyarel, Hüseyin

    2011-09-01

    To evaluate the effect of chronic biomass fuel (BMF) smoke exposure on peripheral endothelial functions. Forty-seven healthy subjects who have been exposed to BMF smoke since birth (mean age 31.6±6.8 years, 21 male) were enrolled in the present cross-sectional observational study. The control group consisted of 32 healthy subjects (mean age 27.9±4.4 years, 11 male). The carotid intima media thickness (CIMT), flow associated dilatation (FAD %) and endothelium independent vasodilatation (GTN %) were assessed in all subjects. The carotid CIMT was defined as the distance between the leading edge of the lumen-intima and the media-adventitia interfaces. FAD % was defined as the percentage change in the internal diameter of the brachial artery during reactive hyperemia related to the baseline. GTN % was defined as the change in diameter in response to the application of 400 µg of glyceril trinitrate relative to the baseline scan at the end of the fourth minute. Statistical analysis was performed using Student's t-test, Chi-square test and Spearman rank order correlation analysis. The average exposure time of the subjects to biomass fuel smoke was 31.7±6.6 years. They have been exposed to dung inhalation products meanly 8.3±1.8 months in a year seasonally. The average daily exposure time was 15.7±3.3 hours. CIMT values of the two groups were not statistically different from each other (0.47±0.09 vs. 0.49±0.06 mm, p=0.138). However, a markedly reduced FAD % was determined in the study group (5.06±4.95 vs. 10.7±4.64, pinhalation products. Therefore, chronic BMF smoke exposure may be a risk factor for the development of endothelial dysfunction.

  17. Final summary report of fuel-dynamics tests H2 and E4

    Doerner, R.C.; Rothman, A.B.; De Volpi, A.; Dickerman, C.E.; Deitrich, L.W; Stahl, D.; Murphy, W.F.

    1976-02-01

    Results of two failure experiments using LMFBR-type fuel during simulated unprotected transient overpower accidents are reported and analyzed. In both experiments, a single fresh fuel pin in a Mark-IIA loop was subjected to a temperature-limited, step-reactivity irradiation in the TREAT reactor. Total energy was 490 MJ in Test H2 and 690 MJ in Test E4. Except for their timing, the sequence of events in the failure scenario was the same for both tests. Local coolant boiling began 25-50 msec before failure. Significant upward fuel flow in the center of the pin started as early as 100 msec before cladding failure. Cladding failure was due to melting after contact with molten fuel and occurred at the top of the fuel column. Formation of an outlet flow-channel blockage began about 10 msec after failure and was complete by 50 msec. Inlet blockage began later and was less extensive. No significant amount of fuel sweepout was observed. Fuel remains separated into a small group of 50-1000-μm fragments and a macroscopic group of chunks and clinkers. The final distribution of fuel remains may have resulted from a delayed fuel/steel interaction in the inlet region

  18. GENUSA Fuel Evolution

    Choithramani, Sylvia; Malpica, Maria [ENUSA Industrias Avanzadas, GENUSA, Josefa Valcarcel, 26 28027 Madrid (Spain); Fawcett, Russel [Global Nuclear Fuel (United States)

    2009-06-15

    deliver improved performance. Relative to the 8x8 fuel operated in the 1980's, today's designs provide {approx}25% more efficiency and power capability and twice as much energy. Because of GENUSA's evolutionary design commitment, these product improvements have been successfully rolled out to our customers with no design or fabrication-related performance surprises. Additionally, this has been accomplished with an accompanying steady improvement in fuel reliability. In the past three decades, fuel reliability has improved by approximately three orders of magnitude. That is, the fuel rod leaker rate has been reduced from over five hundred rods per million operating, to less than ten. In past decades, most plants experienced failures each cycle, and fleet-wide failure mechanisms drove reliability statistics. Today, a small minority of our customers' plants experience failures in any cycle, mainly recurrent, low level debris fretting failures in a handful of plants. GENUSA is committed to providing the most robust, and balanced, fuel solutions to our customers based on our extensive experience and technological capabilities. Identifying and successfully mitigating the mechanisms that cause fuel failures has been instrumental in this observed improvement in fuel reliability. GENUSA systematically identified and eliminated mechanisms leading to failure through pool-side and hot cell examinations, and feedback of lessons learned into the design and fabrication of the fuel. Some of the highly successful mitigating actions during this history include: - Improved pellet fabrication in the 1970's to eliminate cladding primary hydride failures; - Corrosion-resistant cladding, with a chemistry and microstructure specifically targeted to protect against crud-induced corrosion (CILC) failures; - Improved cladding and welding fabrication and inspection techniques that assured the hermeticity and quality of the delivered fuel rod; - Tightened pellet missing

  19. Analysis of trastuzumab and chemotherapy in advanced breast cancer after the failure of at least one earlier combination: An observational study

    Locker Gottfried J

    2006-03-01

    Full Text Available Abstract Background Combining trastuzumab and chemotherapy is standard in her2/neu overexpressing advanced breast cancer. It is not established however, whether trastuzumab treatment should continue after the failure of one earlier combination. In this trial, we report our experience with continued treatment beyond disease progression. Methods Fifty-four patients, median age 46 years, range 25–73 years, were included. We analysed for time to tumour progression (TTP for first, second and beyond second line treatment, response rates and overall survival. Results Median time of observation was 24 months, range 7–51. Response rates for first line treatment were 7.4% complete remission (CR, 35.2% partial remissions (PR, 42.6% stable disease > 6 months (SD and 14.8% of patients experienced disease progression despite treatment (PD. Corresponding numbers for second line were 3.7% CR, 22.2% PR, 42.6% SD and 31.5% PD; numbers for treatment beyond second line (60 therapies, 33 pts 3rd line, 18 pts 4th line, 6 pts 5th line, 2 pts 6th line and 1 patient 7th line were 1.7% CR, 28.3% PR, 28.3% SD and 41.6% PD respectively. Median TTP was 6 months (m in the first line setting, and also 6 m for second line and beyond second line. An asymptomatic drop of left ventricular ejection fraction below 50% was observed in one patient. No case of symptomatic congestive heart failure was observed. Conclusion The data presented clearly strengthen evidence that patients do profit from continued trastuzumab treatment. The fact that TTP did not decrease significantly from first line to beyond second line treatment is especially noteworthy. Still, randomized trials are warranted.

  20. Vortex-ring mixing as a measure of diastolic function of the human heart: Phantom validation and initial observations in healthy volunteers and patients with heart failure.

    Töger, Johannes; Kanski, Mikael; Arvidsson, Per M; Carlsson, Marcus; Kovács, Sándor J; Borgquist, Rasmus; Revstedt, Johan; Söderlind, Gustaf; Arheden, Håkan; Heiberg, Einar

    2016-06-01

    To present and validate a new method for 4D flow quantification of vortex-ring mixing during early, rapid filling of the left ventricle (LV) as a potential index of diastolic dysfunction and heart failure. 4D flow mixing measurements were validated using planar laser-induced fluorescence (PLIF) in a phantom setup. Controls (n = 23) and heart failure patients (n = 23) were studied using 4D flow at 1.5T (26 subjects) or 3T (20 subjects) to determine vortex volume (VV) and inflowing volume (VVinflow ). The volume mixed into the vortex-ring was quantified as VVmix-in = VV-VVinflow . The mixing ratio was defined as MXR = VVmix-in /VV. Furthermore, we quantified the fraction of the end-systolic volume (ESV) mixed into the vortex-ring (VVmix-in /ESV) and the fraction of the LV volume at diastasis (DV) occupied by the vortex-ring (VV/DV). PLIF validation of MXR showed fair agreement (R(2) = 0.45, mean ± SD 1 ± 6%). MXR was higher in patients compared to controls (28 ± 11% vs. 16 ± 10%, P Vortex-ring mixing can be quantified using 4D flow. The differences in mixing parameters observed between controls and patients motivate further investigation as indices of diastolic dysfunction. J. Magn. Reson. Imaging 2016;43:1386-1397. © 2015 Wiley Periodicals, Inc.

  1. Potential of European 14CO2 observation network to estimate the fossil fuel CO2 emissions via atmospheric inversions

    Wang, Yilong; Broquet, Grégoire; Ciais, Philippe; Chevallier, Frédéric; Vogel, Felix; Wu, Lin; Yin, Yi; Wang, Rong; Tao, Shu

    2018-03-01

    Combining measurements of atmospheric CO2 and its radiocarbon (14CO2) fraction and transport modeling in atmospheric inversions offers a way to derive improved estimates of CO2 emitted from fossil fuel (FFCO2). In this study, we solve for the monthly FFCO2 emission budgets at regional scale (i.e., the size of a medium-sized country in Europe) and investigate the performance of different observation networks and sampling strategies across Europe. The inversion system is built on the LMDZv4 global transport model at 3.75° × 2.5° resolution. We conduct Observing System Simulation Experiments (OSSEs) and use two types of diagnostics to assess the potential of the observation and inverse modeling frameworks. The first one relies on the theoretical computation of the uncertainty in the estimate of emissions from the inversion, known as posterior uncertainty, and on the uncertainty reduction compared to the uncertainty in the inventories of these emissions, which are used as a prior knowledge by the inversion (called prior uncertainty). The second one is based on comparisons of prior and posterior estimates of the emission to synthetic true emissions when these true emissions are used beforehand to generate the synthetic fossil fuel CO2 mixing ratio measurements that are assimilated in the inversion. With 17 stations currently measuring 14CO2 across Europe using 2-week integrated sampling, the uncertainty reduction for monthly FFCO2 emissions in a country where the network is rather dense like Germany, is larger than 30 %. With the 43 14CO2 measurement stations planned in Europe, the uncertainty reduction for monthly FFCO2 emissions is increased for the UK, France, Italy, eastern Europe and the Balkans, depending on the configuration of prior uncertainty. Further increasing the number of stations or the sampling frequency improves the uncertainty reduction (up to 40 to 70 %) in high emitting regions, but the performance of the inversion remains limited over low

  2. Potential of European 14CO2 observation network to estimate the fossil fuel CO2 emissions via atmospheric inversions

    Y. Wang

    2018-03-01

    Full Text Available Combining measurements of atmospheric CO2 and its radiocarbon (14CO2 fraction and transport modeling in atmospheric inversions offers a way to derive improved estimates of CO2 emitted from fossil fuel (FFCO2. In this study, we solve for the monthly FFCO2 emission budgets at regional scale (i.e., the size of a medium-sized country in Europe and investigate the performance of different observation networks and sampling strategies across Europe. The inversion system is built on the LMDZv4 global transport model at 3.75°  ×  2.5° resolution. We conduct Observing System Simulation Experiments (OSSEs and use two types of diagnostics to assess the potential of the observation and inverse modeling frameworks. The first one relies on the theoretical computation of the uncertainty in the estimate of emissions from the inversion, known as posterior uncertainty, and on the uncertainty reduction compared to the uncertainty in the inventories of these emissions, which are used as a prior knowledge by the inversion (called prior uncertainty. The second one is based on comparisons of prior and posterior estimates of the emission to synthetic true emissions when these true emissions are used beforehand to generate the synthetic fossil fuel CO2 mixing ratio measurements that are assimilated in the inversion. With 17 stations currently measuring 14CO2 across Europe using 2-week integrated sampling, the uncertainty reduction for monthly FFCO2 emissions in a country where the network is rather dense like Germany, is larger than 30 %. With the 43 14CO2 measurement stations planned in Europe, the uncertainty reduction for monthly FFCO2 emissions is increased for the UK, France, Italy, eastern Europe and the Balkans, depending on the configuration of prior uncertainty. Further increasing the number of stations or the sampling frequency improves the uncertainty reduction (up to 40 to 70 % in high emitting regions, but the performance of the inversion

  3. Operational limitations of light water reactors relating to fuel performance

    Cheng, H.S.

    1976-07-01

    General aspects of fuel performance for typical Boiling and Pressurized Water Reactors are presented. Emphasis is placed on fuel failures in order to make clear important operational limitations. A discussion of fuel element designs is first given to provide the background information for the subsequent discussion of several fuel failure modes that have been identified. Fuel failure experiences through December 31, 1974, are summarized. The operational limitations that are required to mitigate the effects of fuel failures are discussed

  4. Fuel performance, design and development

    Prasad, P.N.; Tripathi, Rahul Mani; Soni, Rakesh; Ravi, M.; Vijay Kumar, S.; Dwivedi, K.P.; Pandarinathan, P.R.; Neema, L.K.

    2006-01-01

    The normal fuel configurations for operating 220 MWe and 540 MWe PHWRs are natural uranium dioxide 19-element and 37- element fuel bundle types respectively. The fuel configuration for BWRs is 6 x 6 fuel. So far, about 330 thousand PHWR fuel bundles and 3500 number of BWR bundles have been irradiated in the 14 PHWRs and 2 BWRs. Improvements in fuel design, fabrication, quality control and operating practices are continuously carried out towards improving fuel utilization as well as reducing fuel failure rate. Efforts have been put to improve the fuel bundle utilization by increasing the fuel discharge burnup of the natural uranium bundles The overall fuel failure rate currently is less than 0.1 % . Presently the core discharge burnups in different reactors are around 7500 MWD/TeU. The paper gives the fuel performance experience over the years in the different power reactors and actions taken to improve fuel performance over the years. (author)

  5. Axial location of cladding failure during a slow transient overpower TREAT test

    Page, R.J.; Murphy, W.F.; Holland, J.W.

    1983-01-01

    The axial location of cladding failure following a transient overpower accident is of importance in fast reactor safety studies in that it is a determining factor in the relocation of fuel, and therefore in the possibility of inherent neutronic shutdown of the reactor. In-pile experimental data on the axial location of cladding failure of fuel in bundles of pins is sparse since, in general, the experimental fuel pin bundles are largely destroyed during the in-pile test. The post-test examination work has been completed for TREAT test J1. It was found that damage to the fuel elements during the irradiation was low enough for an accurate observation of the location of cladding failure to be made for each of the seven pins

  6. Time/motion observations and dose analysis of reactor loading, transportation, and dry unloading of an overweight truck spent fuel shipment

    Hostick, C.J.; Lavender, J.C.; Wakeman, B.H.

    1992-04-01

    This document presents observed activity durations and radiation dose analyses for an overweight truck shipment of pressurized water reactor (PWR) spent fuel from the Surry Power Station in Virginia to the Idaho National Engineering Laboratory. The shipment consisted of a TN-8L shipping cask carrying three 9-year-old PWR spent fuel assemblies. Handling times and dose analyses for at-reactor activities were completed by Virginia Electric and Power Company (Virginia Power) personnel. Observations of in-transit and unloading activities were made by Pacific Northwest Laboratory (PNL) personnel, who followed the shipment for approximately 2800 miles and observed cask unloading activities. In-transit dose estimates were calculated using dose rate maps provided by Virginia Power for a fully loaded TN-8L shipping cask. The dose analysis for the cask unloading operations is based on the observations of PNL personnel

  7. BWR fuel experience with zinc injection

    Levin, H.A.; Garcia, S.E.

    1995-01-01

    In 1982 a correlation between low primary recirculation system dose rates in BWR's and the presence of ionic zinc in reactor water was identified. The source of the zinc was primarily from Admiralty brass condensers. Plants with brass condensers are called ''natural zinc'' plants. Brass condensers were also a source of copper that was implicated in crude induced localized corrosion (CILC) fuel failures. In 1986 the first BWR intentionally injected zinc for the benefits of dose rate control. Although zinc alone was never implicated in fuel degradation of failures, a comprehensive fuel surveillance program was initiated to monitor fuel performance. Currently there are 14 plants that are injecting zinc. Six of these plants are also on hydrogen water chemistry. This paper describes the effect on both Zircaloy corrosion and the cruding characteristics as a result of these changes in water chemistry. Fuel rod corrosion was found to be independent of the specific water chemistry of the plants. The corrosion behavior was the same with the additions of zinc alone or zinc plus hydrogen and well within the operating experience for fuel without either of these additions. No change was observed in the amounts of crude deposited on the fuel rods, both for the adherent and loosely held deposits. One of the effects of the zinc addition was the trend to form more of the zinc rich iron spinel in the fuel deposits rather than the hematite deposits that are predominantly formed with non additive water chemistry

  8. Behavior and failure of uniformly hydrided Zircaloy-4 fuel claddings between 25 C and 480 C under various stress states, including RIA loading conditions

    Le Saux, M.; Carassou, S.; Averty, X.; Le Saux, M.; Besson, J.; Poussard, C.

    2010-01-01

    The anisotropic plastic behavior and the fracture of as-received and hydrided Cold-Worked Stress Relieved Zircaloy-4 cladding tubes are investigated under thermal-mechanical loading conditions representative of Pellet-Clad Mechanical Interaction during Reactivity Initiated Accidents in Pressurized Water Reactors. In order to study the combined effects of temperature, hydrogen content, loading direction and stress state, Axial Tensile, Hoop Tensile, Expansion Due to Compression and hoop Plane Strain Tensile tests are performed at room temperature, 350 C and 480 C on the material containing various hydrogen contents up to 1200 wt. ppm (hydrides are circumferential and homogeneously distributed). These tests are combined with digital image correlation and metallographic and fractographic observations at different scales. The flow stress of the material decreases with increasing temperature. The material is either strengthened or softened by hydrogen depending on temperature and hydrogen content. Plastic anisotropy depends on temperature but not on hydrogen content. The ductility of the material decreases with increasing hydrogen content at room temperature due to damage nucleation by hydride cracking. The plastic strain that leads to hydride fracture at room temperature decreases with increasing hydrogen content. The influence of stress triaxiality on hydride cracking is negligible in the studied range. The influence of hydrogen on material ductility is negligible at 350 C and 480 C since hydrides do not crack at these temperatures. The ductility of the material increases with increasing temperature. The evolution of material ductility is associated with a change in both the macroscopic fracture mode of the specimens and the microscopic failure mechanisms. (authors)

  9. RIA tests in CABRI with MOX fuel

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

    2000-01-01

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

  10. Legacy Vehicle Fuel System Testing with Intermediate Ethanol Blends

    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.

  11. Experience of iodine, caesium and noble gas release from AGR failures

    Chapman, C.J.; Harris, A.M.; Phillips, M.E.

    1985-01-01

    In the event of a fuel failure in an Advanced Gas Cooled Reactor (AGR), the quantity of fission products available for release to the environment is determined by the transport of fission products in the UO 2 fuel, by the possible retention of fission products in the fuel can interspace and by the deposition of fission products on gas circuit surfaces ('plate-out'). The fission products of principal radiological concern are radioactive caesium (Cs-137 and Cs-134) and iodine (principally I-131). Results are summarised of a number of experiments which were designed to study the release of these fission products from individual fuel failures in the prototype AGR at Windscale. Results are also presented of fission product release from failures in commercial AGRs. Comparisons of measured releases of caesium and iodine relative to the release of the noble gas fission products show that, for some fuel failures, there is a significant retention of caesium and iodine within the fuel can interspace. Under normal conditions circuit deposition reduces caesium and iodine gas concentrations by several orders of magnitude. Differing release behaviour of caesium and iodine from the failures is examined together with subsequent deposition within the sampling equipment. These observations are important factors which must be considered in developing an understanding of the mechanisms involved in circuit deposition. (author)

  12. Improving the reliability of fuel Enusa; Mejora de la fiabilidad del combustible en Enusa

    Choithramani, S.; Quecedo, M.

    2015-07-01

    ENUSA is committed to providing our customers with fuel designs that meet their needs for operational efficiency, power, energy, performance and reliability. ENUSAs current fuel designs, covering BWR and PWR technologies, incorporate highest performance with proven reliability features developed along nuclear power operation history. As of January 2015, ENUSA has manufactured more than 20.000 fuel assemblies (around half BWR and half PWR), with operating conditions reflecting varying reactor power densities, cycle lengths, operating strategies and water chemistry environments. This experience brings the knowledge to model our fuel behavior and acts as the principal instrument to identify and characterize the failure mechanisms of our fuel. Based on the information obtained from all this years of operation, ENUSA has progressively developed and implemented numerous mitigating actions identified upon the knowledge on failure mechanisms, which are the bases for the fuel reliability improvement program. Contemporaneously to this implementation, a positive trend on ENUSA fuel reliability has been observed. (Author)

  13. IAEA activities on nuclear fuel

    Basak, U.

    2011-01-01

    In this paper a brief description and the main objectives of IAEA Programme B on Nuclear fuel cycle are given. The following Coordinated Research Projects: 1) FUel performance at high burn-up and in ageing plant by management and optimisation of WAter Chemistry Technologies (FUWAC ); 2) Near Term and Promising Long Term Options for Deployment of Thorium Based Nuclear Energy; 3) Fuel Modelling (FUMEX-III) are shortly described. The data collected by the IAEA Expert Group of Fuel Failures in Water Cooled Reactors including information about fuel failure cause for PWR (1994-2006) and failure mechanisms for BWR fuel (1994-2006) are shown. The just published Fuel Failure Handbook as well as preparation of a Monograph on Zirconium including an overview of Zirconium for nuclear applications are presented. The current projects in Sub-programme B2 - Power Reactor Fuel Engineering are also listed

  14. Nanoscale observations of the operational failure for phase-change-type nonvolatile memory devices using Ge2Sb2Te5 chalcogenide thin films

    Yoon, Sung-Min; Choi, Kyu-Jeong; Lee, Nam-Yeal; Lee, Seung-Yun; Park, Young-Sam; Yu, Byoung-Gon

    2007-01-01

    In this study, a phase-change memory device was fabricated and the origin of device failure mode was examined using transmission electron microscopy (TEM) and energy dispersive X-ray spectroscopy (EDS). Ge 2 Sb 2 Te 5 (GST) was used as the active phase-change material in the memory device and the active pore size was designed to be 0.5 μm. After the programming signals of more than 2x10 6 cycles were repeatedly applied to the device, the high-resistance memory state (reset) could not be rewritten and the cell resistance was fixed at the low-resistance state (set). Based on TEM and EDS studies, Sb excess and Ge deficiency in the device operating region had a strong effect on device reliability, especially under endurance-demanding conditions. An abnormal segregation and oxidation of Ge also was observed in the region between the device operating and inactive peripheral regions. To guarantee an data endurability of more than 1x10 10 cycles of PRAM, it is very important to develop phase-change materials with more stable compositions and to reduce the current required for programming

  15. Normalizing the maximum permissible seal failure of the fuel cladding of VVER and the activity of the fission products in the coolant

    Luzanova, L.M.; Miglo, V.N.; Slavyagin, P.D.

    1993-01-01

    In most countries developing a nuclear power industry based on pressurized water reactors, one of the conditions for issuing a license under normal operating conditions for issuing a license stipulates that the fuel elements may not lose their hermetic seal either under normal operating conditions or during presumable disturbances of the conditions of normal use. At a conference on radiation safety the ALARA principle was taken to be fundamental, it being attempted to keep the activity of the coolant of the primary circuit, including the fission products emerging from unsealed fuel elements, to a level as low as reasonably possible. As many years of experience in the nuclear power industry have shown, nuclear power stations are in many cases operated with nonhermetic fuel elements in the core. Therefore, from the point of view of safety and economy, the best way to operate a power plant is to try to ensure maximum burnup of the fuel of the unsealed elements as they operate within the limits of safe activity of the fission products in the fuel circuits

  16. Full core operation in JRR-3 with LEU fuels

    Murayama, Y.; Issiki, M.

    1995-01-01

    The new JRR-3 a 20MWT swimming pool type research reactor, is made up of plate type LEU fuel elements with U-Al x fuel at 2.2 gU/cm 3 . Reconstruction work for the new JR-3 was a good success, and common operation started in November 1990, and 7 cycles (26 days operation/cycle) have passed. We have no experience in using such a high uranium density fuel element with aluminide fuel. So we plan to examine the condition of the irradiated fuel elements with three methods, that is, measurement of the value of FFD in operation, observation of external view of the fuels in refueling work and postirradiation examination after maximum burn-up will be established. In the results of the first two methods, the fuel elements of JRR-3 is burned up normally and have no evidence of failure. (author)

  17. Present status of fuel motion detection by radiation

    Sumita, Kenji; Mizuta, Hiroshi; Ishizuka, Makoto; Ara, Katsuyuki; Nakata, Hirokatsu.

    1978-05-01

    In reactor safety research, it is important to know transient fuel behavior under accidental conditions. Transient histories such as temperature and axial expansion of fuel and cladding and internal pressure of fuel rod are thus measured in experiments simulating accidents. If fuel motion could then be observed during and after fuel failure, this would greatly make for fuel behavior research. The present status is reviewed of fuel motion detections by radiations such as neutron, γ-ray and X-ray, including the principle and system. A neutron hodoscope among them is used already with practical results in in-reactor safety experiments of sodium-cooled fast breeder reactor. So, this is described in detail and its conceptual design as applied to the NSRR is presented. (auth.)

  18. Fission product phases in irradiated carbide fuels

    Ewart, F.T.; Sharpe, B.M.; Taylor, R.G.

    1975-09-01

    Oxide fuels have been widely adopted as 'first charge' fuels for demonstration fast reactors. However, because of the improved breeding characteristics, carbides are being investigated in a number of laboratories as possible advanced fuels. Irradiation experiments on uranium and mixed uranium-plutonium carbides have been widely reported but the instances where segregate phases have been found and subjected to electron probe analysis are relatively few. Several observations of such segregate phases have now been made over a period of time and these are collected together in this document. Some seven fuel pins have been examined. Two of the irradiations were in thermal materials testing reactors (MTR); the remainder were experimental assemblies of carbide gas bonded oxycarbide and sodium bonded oxycarbide in the Dounreay Fast Reactor (DFR). All fuel pins completed their irradiation without failure. (author)

  19. HTGR fuel performance basis

    Shamasundar, B.I.; Stansfield, O.M.; Jensen, D.D.

    1982-01-01

    The safety characteristics of the high-temperature gas-cooled reactor (HTGR) during normal and accident conditions are determined in part by HTGR fuel performance. During normal operation, less than 0.1% fuel failure occurs, primarily from defective particles. This low fuel failure fraction limits circulating activity to acceptable levels. During severe accidents, the radiological consequence is influenced by high-temperature fuel particle behavior. An empirical fuel failure model, supported by recent experimental data, is presented. The onset of significant fuel particle failure occurs at temperatures in excess of 1600 0 C, and complete fuel failure occurs at 2660 0 C. This indicates that the fuel is more retentive at higher temperatures than previously assumed. The more retentive nature of the fuel coupled with the high thermal capacitance of the core results in slow release of fission products from the core during severe accidents. The slow release of fission products over hundreds of hours allows for decay of short-lived isotopes. The slow and limited release of fission products under HTGR accident conditions results in very low off-site doses. The slow nature of the accident provides more time for operator action to mitigate the accident and for local and state authorities to respond. These features can be used to take advantage of close-in siting for process applications, flexibility in site selection, and emergency planning

  20. Respiratory Failure

    Respiratory failure happens when not enough oxygen passes from your lungs into your blood. Your body's organs, ... brain, need oxygen-rich blood to work well. Respiratory failure also can happen if your lungs can' ...

  1. A reassessment of the potential for an alpha-mode containment failure and a review of the current understanding of broader fuel-coolant interaction issues. Second steam explosion review group workshop

    Basu, S. [Nuclear Regulatory Commission, Washington, DC (United States); Ginsberg, T. [Brookhaven National Lab., Upton, NY (United States)

    1996-08-01

    This report summarizes the review and evaluation by experts of the current understanding of the molten fuel-coolant interaction (FCI) issues covering the complete spectrum of interactions, i.e., from mild quenching to very energetic interactions including those that could lead to the alpha-mode containment failure. Of the eleven experts polled, all but two concluded that the alpha-mode failure issue was resolved from a risk perspective, meaning that this mode of failure is of very low probability, that it is of little or no significance to the overall risk from a nuclear power plant, and that any further reduction in residual uncertainties is not likely to change the probability in an appreciable manner. To a lesser degree, discussions also took place on the broader FCI issues such as mild quenching of core melt during non-explosive FCI, and shock loading of lower head and ex-vessel support structures arising from explosive localized FCIs. These latter issues are relevant with regard to determining the efficacy of certain accident management strategies for operating reactors as well as for advanced light water reactors. The experts reviewed the status of understanding of the FCI phenomena in the context of these broader issues, identified residual uncertainties in the understanding, and recommended future research (both experimental and analytical) to reduce the uncertainties.

  2. A reassessment of the potential for an alpha-mode containment failure and a review of the current understanding of broader fuel-coolant interaction issues. Second steam explosion review group workshop

    Basu, S.; Ginsberg, T.

    1996-08-01

    This report summarizes the review and evaluation by experts of the current understanding of the molten fuel-coolant interaction (FCI) issues covering the complete spectrum of interactions, i.e., from mild quenching to very energetic interactions including those that could lead to the alpha-mode containment failure. Of the eleven experts polled, all but two concluded that the alpha-mode failure issue was resolved from a risk perspective, meaning that this mode of failure is of very low probability, that it is of little or no significance to the overall risk from a nuclear power plant, and that any further reduction in residual uncertainties is not likely to change the probability in an appreciable manner. To a lesser degree, discussions also took place on the broader FCI issues such as mild quenching of core melt during non-explosive FCI, and shock loading of lower head and ex-vessel support structures arising from explosive localized FCIs. These latter issues are relevant with regard to determining the efficacy of certain accident management strategies for operating reactors as well as for advanced light water reactors. The experts reviewed the status of understanding of the FCI phenomena in the context of these broader issues, identified residual uncertainties in the understanding, and recommended future research (both experimental and analytical) to reduce the uncertainties

  3. Heart Failure

    Heart failure is a condition in which the heart can't pump enough blood to meet the body's needs. Heart failure does not mean that your heart has stopped ... and shortness of breath Common causes of heart failure are coronary artery disease, high blood pressure and ...

  4. Analysis of material effect in molten fuel-coolant interaction, comparison of thermodynamic calculations and experimental observations

    Tyrpekl, Václav; Piluso, P.

    2012-01-01

    Roč. 46, AUGUST (2012), s. 197-203 ISSN 0306-4549 Institutional support: RVO:61388980 Keywords : Nuclear reactor severe accident * Fuel -Coolant Interaction * Material effect * Steam explosion Subject RIV: CA - Inorganic Chemistry Impact factor: 0.800, year: 2012

  5. Fuel rods

    Hattori, Shinji; Kajiwara, Koichi.

    1980-01-01

    Purpose: To ensure the safety for the fuel rod failures by adapting plenum springs to function when small forces such as during transportation of fuel rods is exerted and not to function the resilient force when a relatively great force is exerted. Constitution: Between an upper end plug and a plenum spring in a fuel rod, is disposed an insertion member to the lower portion of which is mounted a pin. This pin is kept upright and causes the plenum spring to function resiliently to the pellets against the loads due to accelerations and mechanical vibrations exerted during transportation of the fuel rods. While on the other hand, if a compression force of a relatively high level is exerted to the plenum spring during reactor operation, the pin of the insertion member is buckled and the insertion member is inserted to the inside of the plenum spring, whereby the pellets are allowed to expand freely and the failures in the fuel elements can be prevented. (Moriyama, K.)

  6. LMFBR operational and experimental local-fault experience, primarily with oxide fuel elements

    Warinner, D.K.

    1980-01-01

    Case-by-case reviews of selective world experience with severe local faults, particularly fuel failure and fuel degradation, are reviewed for two sodium-cooled thermal reactors, several LMFBRs, and LMFBR-fuels experiments. The review summarizes fuel-failure frequency and illustrates the results of the most damaging LMFBR local-fault experiences of the last 20 years beginning with BR-5 and including DFR, BOR-60, BR2's MFBS- and Mol-loops experiments, Fermi, KNK, Rapsodie, EBR-II, and TREAT-D2. Local-fault accommodation is demonstrated and a need to more thoroughly investigate delayed-neutron and gaseous-fission-product signals is highlighted in view of uranate formation, observed blockages, and slow fuel-element failure-propagation

  7. LMFBR operational and experimental in-core local-fault experience, primarily with oxide fuel elements

    Warinner, D.K.

    Case-by-case reviews of selective world experience with severe local faults, particularly fuel failure and fuel degradation, are reviewed for two sodium-cooled thermal reactors, several LMFBRs, and LMFBR-fuels experiments. The review summarizes fuel-failure frequency and illustrates the results of the most damaging LMFBR local-fault experiences of the last 20 years beginning with BR-5 and including DFR, BOR-60, BR2's MFBS-and Mol-loops experiments, Fermi, KNK, Rapsodie, EBR-II, and TREAT-D2. Local-fault accommodation is demonstrated and a need to more thoroughly investigate delayed-neutron and gaseous-fission-product signals is highlighted in view of uranate formation, observed blockages, and slow fuel-element failure-propagation

  8. Advances in HTGR fuel performance models

    Stansfield, O.M.; Goodin, D.T.; Hanson, D.L.; Turner, R.F.

    1985-01-01

    Advances in HTGR fuel performance models have improved the agreement between observed and predicted performance and contributed to an enhanced position of the HTGR with regard to investment risk and passive safety. Heavy metal contamination is the source of about 55% of the circulating activity in the HTGR during normal operation, and the remainder comes primarily from particles which failed because of defective or missing buffer coatings. These failed particles make up about 5 x 10 -4 fraction of the total core inventory. In addition to prediction of fuel performance during normal operation, the models are used to determine fuel failure and fission product release during core heat-up accident conditions. The mechanistic nature of the models, which incorporate all important failure modes, permits the prediction of performance from the relatively modest accident temperatures of a passively safe HTGR to the much more severe accident conditions of the larger 2240-MW/t HTGR. (author)

  9. High burnup (41 - 61 GWd/tU) BWR fuel behavior under reactivity initiated accident conditions

    Nakamura, Takehiko; Kusagaya, Kazuyuki; Yoshinaga, Makio; Uetsuka, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-12-01

    High burnup boiling water reactor (BWR) fuel was pulse irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity initiated accident (RIA) conditions. Temperature, deformation, failure, and fission gas release behavior under the simulated RIA condition was studied in the tests. Fuel failure due to pellet-cladding mechanical interaction (PCMI) did not occur in the tests with typical domestic BWR fuel at burnups up to 56 GWd/tU, because they had limited cladding embrittlement due to hydrogen absorption of about 100 ppm or less. However, the cladding failure occurred in tests with fuel at a burnup of 61 GWd/tU, in which the peak hydrogen content in the cladding was above 150 ppm. This type of failure was observed for the first time in BWR fuels. The cladding failure occurred at fuel enthalpies of 260 to 360 J/g (62 to 86 cal/g), which were higher than the PCMI failure thresholds decided by the Japanese Nuclear Safety Commission. From post-test examinations of the failed fuel, it was found that the crack in the BWR cladding progressed in a manner different from the one in PWR cladding failed in earlier tests, owing to its more randomly oriented hydride distribution. Because of these differences, the BWR fuel was judged to have failed at hydrogen contents lower than those of the PWR fuel. Comparison of the test results with code calculations revealed that the PCMI failure was caused by thermal expansion of pellets, rather than by the fission gas expansion in the pellets. The gas expansion, however, was found to cause large cladding hoop deformation later after the cladding temperature escalated. (author)

  10. Echo and natriuretic peptide guided therapy improves outcome and reduces worsening renal function in systolic heart failure: An observational study of 1137 outpatients.

    Simioniuc, Anca; Carluccio, Erberto; Ghio, Stefano; Rossi, Andrea; Biagioli, Paolo; Reboldi, Gianpaolo; Galeotti, Gian Giacomo; Lu, Fei; Zara, Cornelia; Whalley, Gillian; Temporelli, Pier Luigi; Dini, Frank Lloyd

    2016-12-01

    B-type natriuretic peptide (BNP) and echocardiography are potentially useful adjunct to guide management of patients with chronic heart failure (HF).Thus, the aim of this retrospective, multicenter study was to compare outcomes and renal function in outpatients with chronic HF with reduced ejection fraction (HFrEF) who underwent an echo and BNP guided or a clinically driven protocol for follow-up. In 1137 consecutive outpatients, management was guided according to echo-Doppler signs of elevated left ventricular filling pressure and BNP levels conforming to the protocol of the Network Labs Ultrasound (NEBULA) in HF Study Group in 570 (mean EF=30%), while management was clinically driven based on the institutional protocol of the HF Unit of the Cardiovascular and Thoracic Department in 567 (mean EF=33%). Propensity score, matching several confounding baseline variables, was used to match pairs based on treatment strategy. The median follow-up was 37.4months. After propensity matching, a lower incidence of death (HR 0.45, 95%CI: 0.30-0.67, p<0.0001), and death or worsening renal function (HR 0.49, 95%CI 0.36-0.67, p<0.0001) was apparent in echo-BNP-guided group compared to clinically-guided group. Worsening of renal function (≥0.3mg/dl increase in serum creatinine) was observed in 9.8% of echo-BNP-guided group and in 21.4% of clinical assessed group (p<0.0001). The daily dose of loop diuretics did not change in echo-BNP-guided group, while it increased in 65% of patients in clinically-guided group (p<0.0001). Echo and BNP guided management may improve the outcome and reduce worsening of renal function in outpatients with chronic HFrEF. Copyright © 2016 Elsevier Ireland Ltd. All rights reserved.

  11. Texas Disasters II: Utilizing NASA Earth Observations to Assist the Texas Forest Service in Mapping and Analyzing Fuel Loads and Phenology in Texas Grasslands

    Brooke, Michael; Williams, Meredith; Fenn, Teresa

    2016-01-01

    The risk of severe wildfires in Texas has been related to weather phenomena such as climate change and recent urban expansion into wild land areas. During recent years, Texas wild land areas have experienced sequences of wet and dry years that have contributed to increased wildfire risk and frequency. To prevent and contain wildfires, the Texas Forest Service (TFS) is tasked with evaluating and reducing potential fire risk to better manage and distribute resources. This task is made more difficult due to the vast and varied landscape of Texas. The TFS assesses fire risk by understanding vegetative fuel types and fuel loads. To better assist the TFS, NASA Earth observations, including Landsat and Moderate Resolution Imaging Specrtoradiometer (MODIS) data, were analyzed to produce maps of vegetation type and specific vegetation phenology as it related to potential wildfire fuel loads. Fuel maps from 2010-2011 and 2014-2015 fire seasons, created by the Texas Disasters I project, were used and provided alternating, complementary map indicators of wildfire risk in Texas. The TFS will utilize the end products and capabilities to evaluate and better understand wildfire risk across Texas.

  12. Fuel manufacturing and utilization

    2005-01-01

    The efficient utilisation of nuclear fuel requires manufacturing facilities capable of making advanced fuel types, with appropriate quality control. Once made, the use of such fuels requires a proper understanding of their behaviour in the reactor environment, so that safe operation for the design life can be achieved. The International Atomic Energy Agency supports Member States to improve in-pile fuel performance and management of materials; and to develop advanced fuel technologies for ensuring reliability and economic efficiency of the nuclear fuel cycle. It provides assistance to Member States to support fuel-manufacturing capability, including quality assurance techniques, optimization of manufacturing parameters and radiation protection. The IAEA supports the development fuel modelling expertise in Member States, covering both normal operation and postulated and severe accident conditions. It provides information and support for the operation of Nuclear Power Plant to ensure that the environment and water chemistry is appropriate for fuel operation. The IAEA supports fuel failure investigations, including equipment for failed fuel detection and for post-irradiation examination and inspection, as well as fuel repair, it provides information and support research into the basic properties of fuel materials, including UO 2 , MOX and zirconium alloys. It further offers guidance on the relationship with back-end requirement (interim storage, transport, reprocessing, disposal), fuel utilization and management, MOX fuels, alternative fuels and advanced fuel technology

  13. Fuel Coolant Interaction Results in the Fuel Pins Melting Facility (PMF)

    Urunashi, H.; Hirabayashi, T.; Mizuta, H.

    1976-01-01

    The experimental work related to FCI at PNC has been concentrated into the molten UO 2 dropping test. After the completion of molten UO 2 drop experiments, emphasis is directed toward the FCI phenomena of the initiating conditions of the accident under the more realistic geometry. The experiments are conducted within the Pin Melt Facility (PMF) in which UO 2 pellets clad in stainless steel are melted by direct electric heating under the stagnant or flowing sodium. The primary objectives of the PMF test are to: - obtain detail experimental results (heat-input, clad temperature, sodium temperature, etc.) on the FCI under TOP and LOF conditions; - observe the movement of the fuel before and after the pin failure by the X-ray cinematography; - observe the degree of coherence of the pin failures; - accumulate the experience of the FCI experiment which is applicable to the subassembly or more larger scale; - simulate the fuel behavior of the in-pile test (GETR, CABRI). The preliminary conclusions can be drawn from the foregoing observations are as follows: - Although the fuel motion and FCI of the closed test section appeared to be different from those of the open test section, the conclusion of the effect of the inside pressure on FCI needs more experimental data. - The best heating condition of the UO 2 pellet for the FCI study with PMF is established as 40 w/cm at the steady state and 1680 J/g of UO 2 during the additional transient state. The total energy deposition of the UO 2 pellet is thus estimated in the range of 2400 J/g of UO 2 -2600 J/g of UO 2 . The analytical model of the fuel pin failure and the subsequent FCI are suggested to count the following parameters: - The fuel pin failure due to the fuel vaporization due to the rapid energy deposition; - Molten fuel, clad and sodium interaction in the fuel pin after the pin failure; - The upward flow of molten fuel with molten clad or vapor sodium, as well as the slumping of molten fuel

  14. Failure analysis of thermally cycled columnar thermal barrier coatings produced by high-velocity-air fuel and axial-suspension-plasma spraying: A design perspective

    Ganvir, A.; Vaidhyanathan, V.; Markocsan, N.; Gupta, M.; Pala, Zdeněk; Lukáč, František

    2018-01-01

    Roč. 44, č. 3 (2018), s. 3161-3172 ISSN 0272-8842 Institutional support: RVO:61389021 Keywords : Columnar Thermal Barrier Coatings * Axial Suspension Plasma spraying * Thermal Cyclic Fatigue * High Velocity Air Fuel Spraying Subject RIV: JK - Corrosion ; Surface Treatment of Materials OBOR OECD: Coating and films Impact factor: 2.986, year: 2016 https://www.sciencedirect.com/science/article/pii/S0272884217325403

  15. Behavior of small-sized BWR fuel under reactivity initiated accident conditions

    Yanagisawa, Kazuaki; Fujishiro, Toshio; Horiki, Oichiro; Chen Dianshan; Takeuchi, Kiyoshi.

    1992-01-01

    The present work was performed on this small-sized BWR fuel, where Zr liner and rod prepressurization were taken as experimental parameters. Experiment was done under simulated reactivity initiated accident (RIA) conditions at Nuclear Safety Research Reactor (NSRR) belonged to Japan Atomic Energy Research Institute (JAERI). Major remarks obtained are as follows: (1) Three different types of the fuel rods consisted of (a) Zr lined/pressurized (0.65MPa), (b) Zr lined/non-pressurized and (c) non-Zr lined/pressurized (o.65MPa) were used, respectively. Failure thresholds of these were not less than that (260 cal/g·fuel) described in Japanese RIA Licensing Guideline. Small-sized BWR and conventional 8 x 8 BWR fuels were considered to be in almost the same level in failure threshold. Failure modes of the three were (a) cladding melt/brittle, (b) cladding melt/brittle and (c) rupture by large ballooning, respectively. (2) The magnitude of pressure pulse at fuel fragmentation was also studied by lined/pressurized and non-lined/pressurized fuels. Above the energy deposition of 370 cal/g·fuel, mechanical energy (or pressure) was found to be released from these fragmented fuels. No measurable difference was, however, observed between the tested fuels and NSRR standard (and conventional 8 x 8 BWR) fuels. (3) It is worthy of mentioning that Zr liner tended to prevent the cladding from large ballooning. Non-lined/pressurized fuel tended to cause wrinkle deformation at cladding. Hence, cladding external was notched much by the wrinkles. (4) Time to fuel failure measured from the tested BWR fuels (pressurization < 0.6MPA) was longer than that measured from PWR fuels (pressurization < 3.2MPa). The magnitude of the former was of the order of 3 ∼ 6s, while that of the latter was < 1s. (J.P.N.)

  16. Contraceptive failure

    Rasch, Vibeke

    2002-01-01

    Most studies focusing on contraceptive failure in relation to pregnancy have focused on contraceptive failure among women having induced abortions, thereby neglecting those women who, despite contraceptive failure, accept the pregnancy and intend to carry the fetus to term. To get a more complete...... picture of the problem of contraceptive failure, this study focuses on contraceptive failure among women with diverse pregnancy outcomes. In all, 3520 pregnant women attending Odense University Hospital were included: 373 had induced abortions, 435 had spontaneous abortions, 97 had ectopic pregnancies......, and 2614 received antenatal care. The variables studied comprise age, partner relationship, number of births, occupational and economical situation, and contraceptive use.Contraceptive failure, defined as contraceptive use (condom, diaphragm, IUD, oral contraception, or another modern method...

  17. Heart Failure

    McMurray, John; Ponikowski, Piotr

    2011-01-01

    Heart failure occurs in 3% to 4% of adults aged over 65 years, usually as a consequence of coronary artery disease or hypertension, and causes breathlessness, effort intolerance, fluid retention, and increased mortality. The 5-year mortality in people with systolic heart failure ranges from 25% to 75%, often owing to sudden death following ventricular arrhythmia. Risks of cardiovascular events are increased in people with left ventricular systolic dysfunction (LVSD) or heart failure.

  18. Longitudinal prospective observational type study about determinants of renal resistive index variations in chronic renal failure patients treated with conventional medical and dietetic therapy

    Simone Brardi

    2017-12-01

    Full Text Available Objective: This longitudinal prospective observational type study was conceived with the aim to examine the impact on renal resistive index (RRI of the variables that we can manipulate with therapeutic and or dietetic interventions in a chronic kidney disease population in order to known which of these variables was statistically related to changes in RRI and therefore could become the object of the greatest therapeutic effort. Material and methods: This study was undertaken between May 2016 to May 2017 in the outpatient nephrology and urology clinic of San Donato Hospital in Arezzo. The study population (84 patients: 47 males and 37 females was randomly selected among the chronic kidney patients (with various degrees of renal impairment affected by hypertension and or diabetes mellitus. After a comprehensive medical examination these patients were submitted to determination of serum creatinine, glycated hemoglobin, 24-hour urinary albumin excretion and finally renal Doppler ultrasonography. Then the patients were submitted to a full therapeutic and dietetic intervention to ameliorate the renal impairment by a wide range of actions and after on average a one-year interval were submitted again to a new medical examination and a second determination of serum creatinine, glycated hemoglobin, 24-hour urinary albumin excretion and a new renal Doppler ultrasonography too. Results: The comparison between basal and final data revealed a slight reduction in the mean of bilateral renal resistance indices (Delta RRI: -0.0182 ± 0.08, associated to a slight increase in the mean glomerular filtration rate (Delta GFR: 0.8738 ± 10.95 ml/min/1.73 m2, a reduction in mean body weight (Delta weight: -1.9548 ± 5.26 Kg and mean BMI (Delta BMI: -0.7643 ± 2.10 Kg/m2 as well as a reduction in the mean systolic blood pressure (Delta systolic blood pressure: -8.8333 ± 25.19 mmHg. Statistical analysis showed statistically significant correlations (p < 0.05 between

  19. Longitudinal prospective observational type study about determinants of renal resistive index variations in chronic renal failure patients treated with conventional medical and dietetic therapy.

    Brardi, Simone; Cevenini, Gabriele; Giovannelli, Vanni; Romano, Giuseppe

    2017-12-31

    This longitudinal prospective observational type study was conceived with the aim to examine the impact on renal resistive index (RRI) of the variables that we can manipulate with therapeutic and or dietetic interventions in a chronic kidney disease population in order to known which of these variables was statistically related to changes in RRI and therefore could become the object of the greatest therapeutic effort. This study was undertaken between May 2016 to May 2017 in the outpatient nephrology and urology clinic of San Donato Hospital in Arezzo. The study population (84 patients: 47 males and 37 females) was randomly selected among the chronic kidney patients (with various degrees of renal impairment) affected by hypertension and or diabetes mellitus. After a comprehensive medical examination these patients were submitted to determination of serum creatinine, glycated hemoglobin, 24-hour urinary albumin excretion and finally renal Doppler ultrasonography. Then the patients were submitted to a full therapeutic and dietetic intervention to ameliorate the renal impairment by a wide range of actions and after on average a one-year interval were submitted again to a new medical examination and a second determination of serum creatinine, glycated hemoglobin, 24-hour urinary albumin excretion and a new renal Doppler ultrasonography too. The comparison between basal and final data revealed a slight reduction in the mean of bilateral renal resistance indices (Delta RRI: -0.0182 ± 0.08), associated to a slight increase in the mean glomerular filtration rate (Delta GFR: 0.8738 ± 10.95 ml/min/1.73 m2), a reduction in mean body weight (Delta weight: -1.9548 ± 5.26 Kg) and mean BMI (Delta BMI: -0.7643 ± 2.10 Kg/m2) as well as a reduction in the mean systolic blood pressure (Delta systolic blood pressure: -8.8333 ± 25.19 mmHg). Statistical analysis showed statistically significant correlations (p chronic renal failure and as a valuable tool to drive the clinical

  20. Influence of Fuel-Matrix Interaction on the Deformation of U-Mo Dispersion Fuel

    Ryu, Ho Jin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, Chicago (United States)

    2014-05-15

    In order to predict the fuel plate failure leading to breakaway swelling in the meat, an understanding of the effects of the fuel-matrix interaction behavior on the deformation of fuel meat is necessary. However, the effects of IL formation on the development of breakaway swelling have not been studied thoroughly. A mechanism that explains large pore growth that leads to breakaway swelling has not been included in the existing fuel performance models. In this study, the effect of the fuel-matrix interaction on large interfacial porosity development at the IL-Al interface is analyzed using both mechanistic correlations and observations from the post-irradiation examination results of U-Mo Dispersion fuels. The effects of fuel-matrix interaction on the fuel performance of U-Mo/Al Dispersion fuel were investigated. Fuel-matrix interaction bears the causes for breakaway swelling that can lead to a fuel failure under a high-power irradiation condition. Fission gas atoms are released from U-Mo particles to the interaction layer via diffusion and recoil. The fission gases released from the U-Mo and produced in the ILs are further released to the IL-Al interface by diffusion in the IL and recoil. Large pore formation at the IL-Al interface is attributed to the active diffusion of fission gas atoms in the ILs and coalescence between the small bubbles there. A model calculation showed that IL growth increases the probability of forming a breakaway swelling condition. ILs are connected to each other and the Al matrix decreases as ILs grow. When more ILs are interconnected, breakaway swelling can occur when the effective stress from the fission gas pressure in the IL-Al interfacial pore becomes larger than the yield strength of the Al matrix.

  1. The analysis of failure data in the presence of critical and degraded failures

    Haugen, Knut; Hokstad, Per; Sandtorv, Helge

    1997-01-01

    Reported failures are often classified into severityclasses, e.g., as critical or degraded. The critical failures correspond to loss of function(s) and are those of main concern. The rate of critical failures is usually estimated by the number of observed critical failures divided by the exposure time, thus ignoring the observed degraded failures. In the present paper failure data are analyzed, applying an alternative estimate for the critical failure rate, also taking the number of observed degraded failures into account. The model includes two alternative failure mechanisms, one being of the shock type, immediately leading to a critical failure, another resulting in a gradual deterioration, leading to a degraded failure before the critical failure occurs. Failure data on safety valves from the OREDA (Offshore REliability DAta) data base are analyzed using this model. The estimate for the critical failure rate is obtained and compared with the standard estimate

  2. Resistance Upset Welding of ODS Steel Fuel Claddings—Evaluation of a Process Parameter Range Based on Metallurgical Observations

    Fabien Corpace

    2017-08-01

    Full Text Available Resistance upset welding is successfully applied to Oxide Dispersion Strengthened (ODS steel fuel cladding. Due to the strong correlation between the mechanical properties and the microstructure of the ODS steel, this study focuses on the consequences of the welding process on the metallurgical state of the PM2000 ODS steel. A range of process parameters is identified to achieve operative welding. Characterizations of the microstructure are correlated to measurements recorded during the welding process. The thinness of the clad is responsible for a thermal unbalance, leading to a higher temperature reached. Its deformation is important and may lead to a lack of joining between the faying surfaces located on the outer part of the join which can be avoided by increasing the dissipated energy or by limiting the clad stick-out. The deformation and the temperature reached trigger a recrystallization phenomenon in the welded area, usually combined with a modification of the yttrium dispersion, i.e., oxide dispersion, which can damage the long-life resistance of the fuel cladding. The process parameters are optimized to limit the deformation of the clad, preventing the compactness defect and the modification of the nanoscale oxide dispersion.

  3. Incidence of virological failure and major regimen change of initial combination antiretroviral therapy in the Latin America and the Caribbean: an observational cohort study

    Cesar, Carina; Jenkins, Cathy A.; Shepherd, Bryan E.; Padgett, Denis; Mejía, Fernando; Ribeiro, Sayonara Rocha; Cortes, Claudia P.; Pape, Jean W.; Madero, Juan Sierra; Fink, Valeria; Sued, Omar; McGowan, Catherine; Cahn, Pedro

    2015-01-01

    Background Access to combination antiretroviral therapy (cART) is expanding in Latin America and the Caribbean (LAC). There is little information in this region regarding incidence of and factors associated with regimen failure and regimen change. Methods Antiretroviral-naïve adults starting cART from 2000-2014 at sites in seven countries throughout LAC were included. Cumulative incidence of virologic failure and major regimen change were estimated with death considered a competing event. Findings 14,027 cART initiators (60% male, median age 37 years, median CD4 156 cells/mm3, median HIV-RNA 5·0 log10 copies/mL, and 28% with clinical AIDS) were followed for a median of 3·9 years. 1,719 patients presented virologic failure and 1,955 had a major regimen change. Excluding GHESKIO-Haiti (which did not regularly measure HIV-RNA), cumulative incidence of virologic failure was 7·8%, 19·2%, and 25·8% at one, three, and five years after cART initiation, respectively; cumulative incidence of major regimen change was 5·9%, 12·7%, and 18·2%. Incidence of major regimen change at GHESKIO-Haiti at five years was 10·7%. Virologic failure was associated with younger age (adjusted hazard ratio[aHR]=2·03 for 20 vs. 40 years; 95% confidence interval[CI] 1·68-2·44), infection through injection-drug use (IDU) (aHR=1·60; 95%CI 1·02-2·52), initiation in earlier calendar years (aHR=1·28 for 2002 vs. 2006; 95%CI 1·13-1·46), and starting with a boosted protease inhibitor (aHR=1·17 vs. non-nucleoside reverse transcriptase inhibitor; 95%CI 1·00-1·64). Interpretation Incidence of virologic failure was generally lower than in North America/Europe. Our results suggest the need to design strategies to reduce failure and major regimen change among younger patients and those with a history of IDU. Funding US National Institutes of Health: U01 AI069923. PMID:26520929

  4. The failure of earthquake failure models

    Gomberg, J.

    2001-01-01

    In this study I show that simple heuristic models and numerical calculations suggest that an entire class of commonly invoked models of earthquake failure processes cannot explain triggering of seismicity by transient or "dynamic" stress changes, such as stress changes associated with passing seismic waves. The models of this class have the common feature that the physical property characterizing failure increases at an accelerating rate when a fault is loaded (stressed) at a constant rate. Examples include models that invoke rate state friction or subcritical crack growth, in which the properties characterizing failure are slip or crack length, respectively. Failure occurs when the rate at which these grow accelerates to values exceeding some critical threshold. These accelerating failure models do not predict the finite durations of dynamically triggered earthquake sequences (e.g., at aftershock or remote distances). Some of the failure models belonging to this class have been used to explain static stress triggering of aftershocks. This may imply that the physical processes underlying dynamic triggering differs or that currently applied models of static triggering require modification. If the former is the case, we might appeal to physical mechanisms relying on oscillatory deformations such as compaction of saturated fault gouge leading to pore pressure increase, or cyclic fatigue. However, if dynamic and static triggering mechanisms differ, one still needs to ask why static triggering models that neglect these dynamic mechanisms appear to explain many observations. If the static and dynamic triggering mechanisms are the same, perhaps assumptions about accelerating failure and/or that triggering advances the failure times of a population of inevitable earthquakes are incorrect.

  5. Benefits of barrier fuel on fuel cycle economics

    Crowther, R.L.; Kunz, C.L.

    1988-01-01

    Barrier fuel rod cladding was developed to eliminate fuel rod failures from pellet/cladding stress/corrosion interaction and to eliminate the associated need to restrict the rate at which fuel rod power can be increased. The performance of barrier cladding has been demonstrated through extensive testing and through production application to many boiling water reactors (BWRs). Power reactor data have shown that barrier fuel rod cladding has a significant beneficial effect on plant capacity factor and plant operating costs and significantly increases fuel reliability. Independent of the fuel reliability benefit, it is less obvious that barrier fuel has a beneficial effect of fuel cycle costs, since barrier cladding is more costly to fabricate. Evaluations, measurements, and development activities, however, have shown that the fuel cycle cost benefits of barrier fuel are large. This paper is a summary of development activities that have shown that application of barrier fuel significantly reduces BWR fuel cycle costs

  6. Synthesis Report on the understanding of failed LMFBR fuel element performance

    Plitz, H.; Bagley, K.; Harbourne, B.

    1990-07-01

    In the coarse of LMFBR operation fuel element failures cannot entirely be avoided as experienced during the operation of PFR, PHENIX and KNK II, where 44 failed fuel elements have been registered between 1978 and 1989. In earlier irradiations, post irradiation examinations showed mixed oxide pin diameter increases up to pin pitch distance, urging to stress reactor safety questions on the potential of fuel pin failure propagation within pin bundles. The chemical interaction of sodium with mixed oxide fuel is regarded to be the key for the understanding of failed fuel behavior. Valuable results on the failed fuel pin behavior during operation were obtained from the SILOE sodium loop test. Based on the bulk of experience with the detection of fuel pin failures, with the continued operation and with the handling of failed pins respectively elements, one can state: 1. All fuel pin failures have been detected securely in time and have been located. 2. Small defects are developing slowly. 3. Even large defects at end-of-life pins resulted in limited fuel loss. 4. Clad failures behave benign in main aspects. 5. The chemical interaction of sodium with mixed oxide is an important factor in the behavior of failed fuel pins, especially at high burnup. 6. Despite different pin designs and different operation conditions, on the basis of 44 failed elements in PFR, PHENIX and KNK II no pin-to-pin propagation was observed and fuel release was rather low, often not detectable. 7. In no case hazard conditions affecting reactor safety have been experienced

  7. Final report of fuel dynamics Test E7

    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. Fuel assembly

    Abe, Hideaki; Sakai, Takao; Ishida, Tomio; Yokota, Norikatsu.

    1992-01-01

    The lower ends of a plurality of plate-like shape memory alloys are secured at the periphery of the upper inside of the handling head of a fuel assembly. As the shape memory alloy, a Cu-Zn alloy, a Ti-Pd alloy or a Fe-Ni alloy is used. When high temperature coolants flow out to the handling head, the shape memory alloy deforms by warping to the outer side more greatly toward the upper portion thereof with the temperature increase of the coolants. As the result, the shape of the flow channel of the coolants is changed so as to enlarge at the exit of the upper end of the fuel assembly. Then, the pressure loss of the coolants in the fuel assembly is decreased by the enlargement. Accordingly, the flow rate of the coolants in the fuel assembly is increased to lower the temperature of the coolants. Further, high temperature coolants and low temperature coolants are mixed sufficiently just above the fuel assembly. This can suppress the temperature fluctuation of the mixed coolants in the upper portion of the reactor core, thereby enabling to decrease a fatigue and failures of the structural components in the upper portion of the reactor core. (I.N.)

  9. MAXimising Involvement in MUltiMorbidity (MAXIMUM) in primary care: protocol for an observation and interview study of patients, GPs and other care providers to identify ways of reducing patient safety failures.

    Daker-White, Gavin; Hays, Rebecca; Esmail, Aneez; Minor, Brian; Barlow, Wendy; Brown, Benjamin; Blakeman, Thomas; Bower, Peter

    2014-08-18

    Increasing numbers of older people are living with multiple long-term health conditions but global healthcare systems and clinical guidelines have traditionally focused on the management of single conditions. Having two or more long-term conditions, or 'multimorbidity', is associated with a range of adverse consequences and poor outcomes and could put patients at increased risk of safety failures. Traditionally, most research into patient safety failures has explored hospital or inpatient settings. Much less is known about patient safety failures in primary care. Our core aims are to understand the mechanisms by which multimorbidity leads to safety failures, to explore the different ways in which patients and services respond (or fail to respond), and to identify opportunities for intervention. We plan to undertake an applied ethnographic study of patients with multimorbidity. Patients' interactions and environments, relevant to their healthcare, will be studied through observations, diary methods and semistructured interviews. A framework, based on previous studies, will be used to organise the collection and analysis of field notes, observations and other qualitative data. This framework includes the domains: access breakdowns, communication breakdowns, continuity of care errors, relationship breakdowns and technical errors. Ethical approval was received from the National Health Service Research Ethics Committee for Wales. An individual case study approach is likely to be most fruitful for exploring the mechanisms by which multimorbidity leads to safety failures. A longitudinal and multiperspective approach will allow for the constant comparison of patient, carer and healthcare worker expectations and experiences related to the provision, integration and management of complex care. This data will be used to explore ways of engaging patients and carers more in their own care using shared decision-making, patient empowerment or other relevant models. Published by

  10. Heart Failure

    ... Other diseases. Chronic diseases — such as diabetes, HIV, hyperthyroidism, hypothyroidism, or a buildup of iron (hemochromatosis) or ... transplantation or support with a ventricular assist device. Prevention The key to preventing heart failure is to ...

  11. Failed fuel detector

    Onodera, Koichi.

    1981-01-01

    Purpose: To improve the reliability of detecting the failure of a fuel rod by imparting a wire disconnection detecting function to a central electrode at the center of a failure mode thereto. Constitution: A wire disconnection detecting terminal is provided at the terminal opposite to the signal output terminal of a central electrode in a failed fuel detector used for detecting the failure of a fuel rod in an atomic power plant using liquid metal as a coolant, and a voltage monitor for monitoring the terminal voltage is connected to the terminal. The disconnection of the central electrode is detected by the failure of the output of the voltage monitor, and an alarm is thus generated. (Aizawa, K.)

  12. Dynamic environmental transmission electron microscopy observation of platinum electrode catalyst deactivation in a proton-exchange-membrane fuel cell.

    Yoshida, Kenta; Xudong, Zhang; Bright, Alexander N; Saitoh, Koh; Tanaka, Nobuo

    2013-02-15

    Spherical-aberration-corrected environmental transmission electron microscopy (AC-ETEM) was applied to study the catalytic activity of platinum/amorphous carbon electrode catalysts in proton-exchange-membrane fuel cells (PEMFCs). These electrode catalysts were characterized in different atmospheres, such as hydrogen and air, and a conventional high vacuum of 10(-5) Pa. A high-speed charge coupled device camera was used to capture real-time movies to dynamically study the diffusion and reconstruction of nanoparticles with an information transfer down to 0.1 nm, a time resolution below 0.2 s and an acceleration voltage of 300 kV. With such high spatial and time resolution, AC-ETEM permits the visualization of surface-atom behaviour that dominates the coalescence and surface-reconstruction processes of the nanoparticles. To contribute to the development of robust PEMFC platinum/amorphous carbon electrode catalysts, the change in the specific surface area of platinum particles was evaluated in hydrogen and air atmospheres. The deactivation of such catalysts during cycle operation is a serious problem that must be resolved for the practical use of PEMFCs in real vehicles. In this paper, the mechanism for the deactivation of platinum/amorphous carbon electrode catalysts is discussed using the decay rate of the specific surface area of platinum particles, measured first in a vacuum and then in hydrogen and air atmospheres for comparison.

  13. Intelligent Design and Intelligent Failure

    Jerman, Gregory

    2015-01-01

    Good Evening, my name is Greg Jerman and for nearly a quarter century I have been performing failure analysis on NASA's aerospace hardware. During that time I had the distinct privilege of keeping the Space Shuttle flying for two thirds of its history. I have analyzed a wide variety of failed hardware from simple electrical cables to cryogenic fuel tanks to high temperature turbine blades. During this time I have found that for all the time we spend intelligently designing things, we need to be equally intelligent about understanding why things fail. The NASA Flight Director for Apollo 13, Gene Kranz, is best known for the expression "Failure is not an option." However, NASA history is filled with failures both large and small, so it might be more accurate to say failure is inevitable. It is how we react and learn from our failures that makes the difference.

  14. Stochastic failure modelling of unidirectional composite ply failure

    Whiteside, M.B.; Pinho, S.T.; Robinson, P.

    2012-01-01

    Stochastic failure envelopes are generated through parallelised Monte Carlo Simulation of a physically based failure criteria for unidirectional carbon fibre/epoxy matrix composite plies. Two examples are presented to demonstrate the consequence on failure prediction of both statistical interaction of failure modes and uncertainty in global misalignment. Global variance-based Sobol sensitivity indices are computed to decompose the observed variance within the stochastic failure envelopes into contributions from physical input parameters. The paper highlights a selection of the potential advantages stochastic methodologies offer over the traditional deterministic approach.

  15. First TREAT transient overpower tests on U-Pu-Zr fuel: M5 and M6

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

    1987-01-01

    Transient Reactor Test Facility (TREAT) tests M5 and M6 were the first transient overpower (TOP) test of the margin to cladding breach and prefailure elongation of metallic U-Pu-Zr ternary fuel, the reference fuel of the integral fast reactor concept. Similar tests on U-5 wt% Fs fueled Experimental Breeder Reactor (EBR)-II driver pins were previously performed and reported. Results from these earlier tests indicated a margin to failure of ∼ 4 times nominal power and significant axial elongation prior to failure, a feature that was very pronounced at low burnups. While these two fuels types are similar in many respects, the ternary alloy exhibits a much more complex physical structure and is typically irradiated at much higher temperatures. Thus, a prime motivation for performing M5 and M6 was to compare the safety-related fuel performance characteristics of U-Fs and U-Pu-Zr. Tests M5 and M6 indicate that, under the TOP conditions used in the tests, ternary fuel displayed about the same margin to failure as U-Fs fuel. At low burnups, ternary fuel showed less prefailure axial elongation than observed in U-Fs pins, but elongations of 3 to 5% might turn out to be typical. Finally, fuel from the breached ternary pin in M6 showed, qualitatively, the same benignly dispersive behavior as U-Fs

  16. Coronary artery disease prevalence and outcome in patients hospitalized with acute heart failure: an observational report from seven Middle Eastern countries.

    Salam, Amar M; Sulaiman, Kadhim; Al-Zakwani, Ibrahim; Alsheikh-Ali, Alawi; Aljaraallah, Mohammed; Al Faleh, Husam; Elasfar, Abdelfatah; Panduranga, Prasanth; Singh, Rajvir; Abi Khalil, Charbel; Al Suwaidi, Jassim

    2016-12-01

    The purpose of this study was to report prevalence, clinical characteristics, precipitating factors, management and outcome of patients with coronary artery disease (CAD) among patients hospitalized with heart failure (HF) in seven Middle Eastern countries and compare them to non-CAD patients. Data were derived from Gulf CARE (Gulf aCute heArt failuRe rEgistry), a prospective multicenter study of 5005 consecutive patients hospitalized with acute HF during February-November 2012 in 7 Middle Eastern countries. The prevalence of CAD among Acute Heart Failure (AHF) patients was 60.2% and varied significantly among the 7 countries (Qatar 65.7%, UAE 66.6%, Kuwait 68.0%, Oman 65.9%, Saudi Arabia 62.5%, Bahrain 52.7% and Yemen 49.1%) with lower values in the lower income countries. CAD patients were older and more likely to have diabetes, hypertension, dyslipidemia and chronic kidney disease. Moreover, CAD patients were more likely to have history of cerebrovascular and peripheral vascular disease when compared to non-CAD patients. In-hospital mortality rates were comparable although CAD patients had more frequent re-hospitalization and worse long-term outcome. However, CAD was not an independent predictor of poor outcome. The prevalence of CAD amongst patients with HF in the Middle East is variable and may be related to healthcare sources. Regional and national studies are needed for assessing further the impact of various etiologies of HF and for developing appropriate strategies to combat this global concern.

  17. Study on light water reactor fuel behavior under reactivity initiated accident condition in TREAT

    Ohnishi, Nobuaki; Ishijima, Kiyomi; Ochiai, Masaaki; Tanzawa, Sadamitsu; Uemura, Mutsumi

    1981-05-01

    This report reviews the results of the fuel failure experiments performed in TREAT in the U.S.A. simulating Reactivity Initiated Accidents. One of the main purposes of the TREAT experiments is the study of the fuel failure behavior, and the other is the study of the molten fuel-water coolant interaction and the consequent hydrogen behavior. This report mainly shows the results of the TREAT experiments studying the fuel failure behavior in Light Water Reactor, and then it describes the fuel failure threshold and the fuel failure mechanism, considering the results of the photographic experiments of the fuel failure behavior with transparent capsules. (author)

  18. Effects of aerosol-vapor JP-8 jet fuel on the functional observational battery, and learning and memory in the rat.

    Baldwin, C M; Houston, F P; Podgornik, M N; Young, R S; Barnes, C A; Witten, M L

    2001-01-01

    To determine whether JP-8 jet fuel affects parameters of the Functional Observational Battery (FOB), visual discrimination, or spatial learning and memory, the authors exposed groups of male Fischer Brown Norway hybrid rats for 28 d to aerosol/vapor-delivered JP-8, or to JP-8 followed by 15 min of aerosolized substance P analogue, or to sham-confined fresh room air. Behavioral testing was accomplished with the U.S. Environmental Protection Agency's Functional Observational Battery. The authors used the Morris swim task to test visual and spatial learning and memory testing. The spatial test included examination of memory for the original target location following 15 d of JP-8 exposure, as well as a 3-d new target location learning paradigm implemented the day that followed the final day of exposure. Only JP-8 exposed animals had significant weight loss by the 2nd week of exposure compared with JP-8 with substance P and control rats; this finding compares with those of prior studies of JP-8 jet fuel. Rats exposed to JP-8 with or without substance P exhibited significantly greater rearing and less grooming behavior over time than did controls during Functional Observational Battery open-field testing. Exposed rats also swam significantly faster than controls during the new target location training and testing, thus supporting the increased activity noted during Functional Observational Battery testing. There were no significant differences between the exposed and control groups' performances during acquisition, retention, or learning of the new platform location in either the visual discrimination or spatial version of the Morris swim task. The data suggest that although visual discrimination and spatial learning and memory were not disrupted by JP-8 exposure, arousal indices and activity measures were distinctly different in these animals.

  19. Performance evaluation of UO2-Zr fuel in power ramp tests

    Knudsen, P.; Bagger, C.

    1977-01-01

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

  20. Where do fossil fuel carbon dioxide emissions from California go? An analysis based on radiocarbon observations and an atmospheric transport model

    Riley, W.J.; Hsueh, D.Y.; Randerson, J.T.; Fischer, M.L.; Hatch, J.G.; Pataki, D.E.; Wang, W.; Goulden, M.L.

    2008-05-01

    Characterizing flow patterns and mixing of fossil fuel-derived CO{sub 2} is important for effectively using atmospheric measurements to constrain emissions inventories. Here we used measurements and a model of atmospheric radiocarbon ({sup 14}C) to investigate the distribution and fluxes of atmospheric fossil fuel CO{sub 2} across the state of California. We sampled {sup 14}C in annual C{sub 3} grasses at 128 sites and used these measurements to test a regional model that simulated anthropogenic and ecosystem CO{sub 2} fluxes, transport in the atmosphere, and the resulting {sup 14}C of annual grasses ({Delta}{sub g}). Average measured {Delta}{sub g} in Los Angeles, San Francisco, the Central Valley, and the North Coast were 27.7 {+-} 20.0, 44.0 {+-} 10.9, 48.7 {+-} 1.9, and 59.9 {+-} 2.5{per_thousand}, respectively, during the 2004-2005 growing season. Model predictions reproduced regional patterns reasonably well, with estimates of 27.6 {+-} 2.4, 39.4 {+-} 3.9, 46.8 {+-} 3.0, and 59.3 {+-} 0.2{per_thousand} for these same regions and corresponding to fossil fuel CO{sub 2} mixing ratios (Cf) of 13.7, 6.1, 4.8, and 0.3 ppm. {Delta}{sub g} spatial heterogeneity in Los Angeles and San Francisco was higher in the measurements than in the predictions, probably from insufficient spatial resolution in the fossil fuel inventories (e.g., freeways are not explicitly included) and transport (e.g., within valleys). We used the model to predict monthly and annual transport patterns of fossil fuel-derived CO{sub 2} within and out of California. Fossil fuel CO{sub 2} emitted in Los Angeles and San Francisco was predicted to move into the Central Valley, raising Cf above that expected from local emissions alone. Annually, about 21, 39, 35, and 5% of fossil fuel emissions leave the California airspace to the north, east, south, and west, respectively, with large seasonal variations in the proportions. Positive correlations between westward fluxes and Santa Ana wind conditions were

  1. A statistical approach to nuclear fuel design and performance

    Cunning, Travis Andrew

    As CANDU fuel failures can have significant economic and operational consequences on the Canadian nuclear power industry, it is essential that factors impacting fuel performance are adequately understood. Current industrial practice relies on deterministic safety analysis and the highly conservative "limit of operating envelope" approach, where all parameters are assumed to be at their limits simultaneously. This results in a conservative prediction of event consequences with little consideration given to the high quality and precision of current manufacturing processes. This study employs a novel approach to the prediction of CANDU fuel reliability. Probability distributions are fitted to actual fuel manufacturing datasets provided by Cameco Fuel Manufacturing, Inc. They are used to form input for two industry-standard fuel performance codes: ELESTRES for the steady-state case and ELOCA for the transient case---a hypothesized 80% reactor outlet header break loss of coolant accident. Using a Monte Carlo technique for input generation, 105 independent trials are conducted and probability distributions are fitted to key model output quantities. Comparing model output against recognized industrial acceptance criteria, no fuel failures are predicted for either case. Output distributions are well removed from failure limit values, implying that margin exists in current fuel manufacturing and design. To validate the results and attempt to reduce the simulation burden of the methodology, two dimensional reduction methods are assessed. Using just 36 trials, both methods are able to produce output distributions that agree strongly with those obtained via the brute-force Monte Carlo method, often to a relative discrepancy of less than 0.3% when predicting the first statistical moment, and a relative discrepancy of less than 5% when predicting the second statistical moment. In terms of global sensitivity, pellet density proves to have the greatest impact on fuel performance

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

    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

  3. Computational modeling for hexcan failure under core distruptive accidental conditions

    Sawada, T.; Ninokata, H.; Shimizu, A. [Tokyo Institute of Technology (Japan)

    1995-09-01

    This paper describes the development of computational modeling for hexcan wall failures under core disruptive accident conditions of fast breeder reactors. A series of out-of-pile experiments named SIMBATH has been analyzed by using the SIMMER-II code. The SIMBATH experiments were performed at KfK in Germany. The experiments used a thermite mixture to simulate fuel. The test geometry of SIMBATH ranged from single pin to 37-pin bundles. In this study, phenomena of hexcan wall failure found in a SIMBATH test were analyzed by SIMMER-II. Although the original model of SIMMER-II did not calculate any hexcan failure, several simple modifications made it possible to reproduce the hexcan wall melt-through observed in the experiment. In this paper the modifications and their significance are discussed for further modeling improvements.

  4. Fuel reliability experience in Finland

    Kekkonen, L.

    2015-01-01

    Four nuclear reactors have operated in Finland now for 35-38 years. The two VVER-440 units at Loviisa Nuclear Power Plant are operated by Fortum and two BWR’s in Olkiluoto are operated by Teollisuuden Voima Oyj (TVO). The fuel reliability experience of the four reactors operating currently in Finland has been very good and the fuel failure rates have been very low. Systematic inspection of spent fuel assemblies, and especially all failed assemblies, is a good practice that is employed in Finland in order to improve fuel reliability and operational safety. Investigation of the root cause of fuel failures is important in developing ways to prevent similar failures in the future. The operational and fuel reliability experience at the Loviisa Nuclear Power Plant has been reported also earlier in the international seminars on WWER Fuel Performance, Modelling and Experimental Support. In this paper the information on fuel reliability experience at Loviisa NPP is updated and also a short summary of the fuel reliability experience at Olkiluoto NPP is given. Keywords: VVER-440, fuel reliability, operational experience, poolside inspections, fuel failure identification. (author)

  5. Crud deposition on fuel in WWER reactors

    Kysela, J.; Svarc, V.; Androva, K.; Ruzickova, M.

    2008-01-01

    Reliability of nuclear fuel and radiation fields surrounding primary systems are important aspects of overall nuclear reactor safety. Corrosion product (crud) deposition on fuel surfaces has implications for fuel performance through heat transfer and local chemistry modifications. Crud is currently one of the key industry issues and has been implicated in several recent cases of crud-related fuel failures and core plugging. Activated crud is deposited on out-of-core surfaces, mainly steam generators, resulting in high radiation fields and high doses of plant staff. Due to radiation build-up in primary circuit systems, decontamination of primary systems components and steam generators is used. Several issues involving decontamination were observed in some cases. After decontamination higher corrosion product release occurs followed by subsequent crud deposition on fuel surfaces. The paper summarizes experience with water chemistry and decontamination that can influence crud deposition on fuel surfaces. The following areas are discussed: 1) Experience with crud deposition, primary water chemistry and decontamination under operating conditions; 2) The behaviour of organic compounds in primary coolant and on fuel surfaces; 3) A proposed experimental programme to study crud deposition. (authors)

  6. Failure Modes

    Jakobsen, K. P.; Burcharth, H. F.; Ibsen, Lars Bo

    1999-01-01

    The present appendix contains the derivation of ten different limit state equations divided on three different failure modes. Five of the limit state equations can be used independently of the characteristics of the subsoil, whereas the remaining five can be used for either drained or undrained s...

  7. Metallic fuel development

    Walters, L.C.

    1987-01-01

    Metallic fuels are capable of achieving high burnup as a result of design modifications instituted in the late 1960's. The gap between the fuel slug and the cladding is fixed such that by the time the fuel swells to the cladding the fission gas bubbles interconnect and release the fission gas to an appropriately sized plenum volume. Interconnected porosity thus provides room for the fuel to deform from further swelling rather than stress the cladding. In addition, the interconnected porosity allows the fuel pin to be tolerant to transient events because as stresses are generated during a transient event the fuel flows rather than applying significant stress to the cladding. Until 1969 a number of metallic fuel alloys were under development in the US. At that time the metallic fuel development program in the US was discontinued in favor of ceramic fuels. However, development had proceeded to the point where it was clear that the zirconium addition to uranium-plutonium fuel would yield a ternary fuel with an adequately high solidus temperature and good compatibility with austenitic stainless steel cladding. Furthermore, several U-Pu-Zr fuel pins had achieved about 6 at.% bu by the late 1960's, without failure, and thus the prospect for high burnup was promising

  8. Behaviour of rock-like oxide fuels under reactivity-initiated accident conditions

    Kazuyuki, Kusagaya; Takehiko, Nakamura; Makio, Yoshinaga; Hiroshi, Akie; Toshiyuki, Yamashita; Hiroshi, Uetsuka

    2002-01-01

    Pulse irradiation tests of three types of un-irradiated rock-like oxide (ROX) fuel - yttria-stabilised zirconia (YSZ) single phase, YSZ and spinel (MgAl 2 O 4 ) homogeneous mixture and particle-dispersed YSZ/spinel - were conducted in the Nuclear Safety Research Reactor to investigate the fuel behaviour under reactivity-initiated accident conditions. The ROX fuels failed at fuel volumetric enthalpies above 10 GJ/m 3 , which was comparable to that of un-irradiated UO 2 fuel. The failure mode of the ROX fuels, however, was quite different from that of the UO 2 fuel. The ROX fuels failed with fuel pellet melting and a part of the molten fuel was released out to the surrounding coolant water. In spite of the release, no significant mechanical energy generation due to fuel/coolant thermal interaction was observed in the tested enthalpy range below∼12 GJ/m 3 . The YSZ type and homogenous YSZ/spinel type ROX fuels failed by cladding burst when their temperatures peaked, while the particle-dispersed YSZ/spinel type ROX fuel seemed to have failed by cladding local melting. (author)

  9. Stress intensity factor at the tip of cladding incipient crack in RIA-simulating experiments for high-burnup PWR fuels

    Udagawa, Yutaka; Suzuki, Motoe; Sugiyama, Tomoyuki; Fuketa, Toyoshi

    2009-01-01

    RIA-simulating experiments for high-burnup PWR fuels have been performed in the NSRR, and the stress intensity factor K 1 at the tip of cladding incipient crack has been evaluated in order to investigate its validity as a PCMI failure threshold under RIA conditions. An incipient crack depth was determined by observation of metallographs. The maximum hydride-rim thickness in the cladding of the test fuel rod was regarded as the incipient crack depth in each test case. Hoop stress in the cladding periphery during the pulse power transient was calculated by the RANNS code. K 1 was calculated based on crack depth and hoop stress. According to the RANNS calculation, PCMI failure cases can be divided into two groups: failure in the elastic phase and failure in the plastic phase. In the former case, elastic deformation was predominant around the incipient crack at failure time. K 1 is available only in this case. In the latter, plastic deformation was predominant around the incipient crack at failure time. Failure in the elastic phase never occurred when K 1 was less than 17 MPa m 1/2 . For failure in the plastic phase, the plastic hoop strain of the cladding periphery at failure time clearly showed a tendency to decrease with incipient crack depth. The combination of K 1 , for failure in the elastic phase, and plastic hoop strain at failure, for failure in the plastic phase, can be an effective index of PCMI failure under RIA conditions. (author)

  10. Nuclear piping criteria for Advanced Light-Water Reactors, Volume 1--Failure mechanisms and corrective actions

    Anon.

    1993-01-01

    This WRC Bulletin concentrates on the major failure mechanisms observed in nuclear power plant piping during the past three decades and on corrective actions taken to minimize or eliminate such failures. These corrective actions are applicable to both replacement piping and the next generation of light-water reactors. This WRC Bulletin was written with the objective of meeting a need for piping criteria in Advanced Light-Water Reactors, but there is application well beyond the LWR industry. This Volume, in particular, is equally applicable to current nuclear power plants, fossil-fueled power plants, and chemical plants including petrochemical. Implementation of the recommendations for mitigation of specific problems should minimize severe failures or cracking and provide substantial economic benefit. This volume uses a case history approach to high-light various failure mechanisms and the corrective actions used to resolve such failures. Particular attention is given to those mechanisms leading to severe piping failures, where severe denotes complete severance, large ''fishmouth'' failures, or long throughwall cracks releasing a minimum of 50 gpm. The major failure mechanisms causing severe failure are erosion-corrosion and vibrational fatigue. Stress corrosion cracking also has been a common problem in nuclear piping systems. In addition thermal fatigue due to mixing-tee and to thermal stratification also is discussed as is microbiologically-induced corrosion. Finally, water hammer, which represents the ultimate in internally-generated dynamic high-energy loads, is discussed

  11. Types of Heart Failure

    ... Introduction Types of Heart Failure Classes of Heart Failure Heart Failure in Children Advanced Heart Failure • Causes and ... and procedures related to heart disease and stroke. Heart Failure Questions to Ask Your Doctor Use these questions ...

  12. Classes of Heart Failure

    ... Introduction Types of Heart Failure Classes of Heart Failure Heart Failure in Children Advanced Heart Failure • Causes and ... and Advanced HF • Tools and Resources • Personal Stories Heart Failure Questions to Ask Your Doctor Use these questions ...

  13. Analysis of hexcan failures occurring during the simulation of severe accidents in liquid metal cooled reactors

    Peppler, W.; Will, H.

    1988-01-01

    Under the SIMBATH programme the physical phenomena of transient material movement and relocation during severe LMFBR accidents are investigated out-of-pile. In most of the SIMBATH bundle experiments a failure of the wrapper was observed. From the safety point of view this has implications on the issue of propagation. By openings into the inter-subassembly gaps pressure relief and material release are possible. From the development of failure, based on measurements made during the simulation tests, and from post-experiment investigations three types of failure mode have been identified: Melt-through of the wrapper wall by a jet of hot material from a failing pin. This happened very early during the test. Sodium boiling in the annular bypass prior to failure has not been detected. Melt-through in the simulated fuel region by severe ablation due to local crust instability combined with intense heat input from the flowing melt. Melt-through in the simulated breeding regions close to blockages. This failure mode was always observed together with sodium gross boiling in the annular channel, i.e. reduced cooling of the wrapper wall. No mechanical failure was detected as a result of the stress concentration in the corners of the hexcan walls. The influence of the internal overpressure is restricted mainly to final break-through after severe ablation and drives the material motions after wrapper failure; it does not control wrapper wall failure in these experiments. (orig.)

  14. Impacts of an ethanol-blended fuel release on groundwater and fate of produced methane: Simulation of field observations

    Rasa, Ehsan; Bekins, Barbara A.; Mackay, Douglas M.; de Sieyes, Nicholas R.; Wilson, John T.; Feris, Kevin P.; Wood, Isaac A.; Scow, Kate M.

    2013-08-01

    In a field experiment at Vandenberg Air Force Base (VAFB) designed to mimic the impact of a small-volume release of E10 (10% ethanol and 90% conventional gasoline), two plumes were created by injecting extracted groundwater spiked with benzene, toluene, and o-xylene, abbreviated BToX (no-ethanol lane) and BToX plus ethanol (with-ethanol lane) for 283 days. We developed a reactive transport model to understand processes controlling the fate of ethanol and BToX. The model was calibrated to the extensive field data set and accounted for concentrations of sulfate, iron, acetate, and methane along with iron-reducing bacteria, sulfate-reducing bacteria, fermentative bacteria, and methanogenic archaea. The benzene plume was about 4.5 times longer in the with-ethanol lane than in the no-ethanol lane. Matching this different behavior in the two lanes required inhibiting benzene degradation in the presence of ethanol. Inclusion of iron reduction with negligible growth of iron reducers was required to reproduce the observed constant degradation rate of benzene. Modeling suggested that vertical dispersion and diffusion of sulfate from an adjacent aquitard were important sources of sulfate in the aquifer. Matching of methane data required incorporating initial fermentation of ethanol to acetate, methane loss by outgassing, and methane oxidation coupled to sulfate and iron reduction. Simulation of microbial growth using dual Monod kinetics, and including inhibition by more favorable electron acceptors, generally resulted in reasonable yields for microbial growth of 0.01-0.05.

  15. Failure Analysis

    Iorio, A.F.; Crespi, J.C.

    1987-01-01

    After ten years of operation at the Atucha I Nuclear Power Station a gear belonging to a pressurized heavy water reactor refuelling machine, failed. The gear box was used to operate the inlet-outlet heavy-water valve of the machine. Visual examination of the gear device showed an absence of lubricant and that several gear teeth were broken at the root. Motion was transmitted with a speed-reducing device with controlled adjustable times in order to produce a proper fitness of the valve closure. The aim of this paper is to discuss the results of the gear failure analysis in order to recommend the proper solution to prevent further failures. (Author)

  16. Failure modes of composite sandwich beams

    Gdoutos E.; Daniel I.M.

    2008-01-01

    A thorough investigation of failure behavior of composite sandwich beams under three-and four-point bending was undertaken. The beams were made of unidirectional carbon/epoxy facings and a PVC closed-cell foam core. The constituent materials were fully characterized and in the case of the foam core, failure envelopes were developed for general two-dimensional states of stress. Various failure modes including facing wrinkling, indentation failure and core failure were observed and compared wit...

  17. Readmissions after Hospitalization for Heart Failure, Acute Myocardial Infarction, or Pneumonia among Young and Middle-Aged Adults: A Retrospective Observational Cohort Study

    Ranasinghe, Isuru; Wang, Yongfei; Dharmarajan, Kumar; Hsieh, Angela F.; Bernheim, Susannah M.; Krumholz, Harlan M.

    2014-01-01

    Background Patients aged ≥65 years are vulnerable to readmissions due to a transient period of generalized risk after hospitalization. However, whether young and middle-aged adults share a similar risk pattern is uncertain. We compared the rate, timing, and readmission diagnoses following hospitalization for heart failure (HF), acute myocardial infarction (AMI), and pneumonia among patients aged 18–64 years with patients aged ≥65 years. Methods and Findings We used an all-payer administrative dataset from California consisting of all hospitalizations for HF (n = 206,141), AMI (n = 107,256), and pneumonia (n = 199,620) from 2007–2009. The primary outcomes were unplanned 30-day readmission rate, timing of readmission, and readmission diagnoses. Our findings show that the readmission rate among patients aged 18–64 years exceeded the readmission rate in patients aged ≥65 years in the HF cohort (23.4% vs. 22.0%, preadmission risk in patients aged 18–64 years was similar to patients ≥65 years in the HF (HR 0.99; 95%CI 0.97–1.02) and pneumonia (HR 0.97; 95%CI 0.94–1.01) cohorts and was marginally lower in the AMI cohort (HR 0.92; 95%CI 0.87–0.96). For all cohorts, the timing of readmission was similar; readmission risks were highest between days 2 and 5 and declined thereafter across all age groups. Diagnoses other than the index admission diagnosis accounted for a substantial proportion of readmissions among age groups readmissions in the HF cohort and 37–45% of readmissions in the AMI cohort, while a non-pulmonary diagnosis represented 61–64% of patients in the pneumonia cohort. Conclusion When adjusted for differences in patient characteristics, young and middle-aged adults have 30-day readmission rates that are similar to elderly patients for HF, AMI, and pneumonia. A generalized risk after hospitalization is present regardless of age. Please see later in the article for the Editors' Summary PMID:25268126

  18. Fuel Exhaling Fuel Cell.

    Manzoor Bhat, Zahid; Thimmappa, Ravikumar; Devendrachari, Mruthyunjayachari Chattanahalli; Kottaichamy, Alagar Raja; Shafi, Shahid Pottachola; Varhade, Swapnil; Gautam, Manu; Thotiyl, Musthafa Ottakam

    2018-01-18

    State-of-the-art proton exchange membrane fuel cells (PEMFCs) anodically inhale H 2 fuel and cathodically expel water molecules. We show an unprecedented fuel cell concept exhibiting cathodic fuel exhalation capability of anodically inhaled fuel, driven by the neutralization energy on decoupling the direct acid-base chemistry. The fuel exhaling fuel cell delivered a peak power density of 70 mW/cm 2 at a peak current density of 160 mA/cm 2 with a cathodic H 2 output of ∼80 mL in 1 h. We illustrate that the energy benefits from the same fuel stream can at least be doubled by directing it through proposed neutralization electrochemical cell prior to PEMFC in a tandem configuration.

  19. MOX fuel irradiation behavior in steady state (irradiation test in HBWR)

    Kohno, S; Kamimura, K [Power Reactor and Nuclear Fuel Development Corp., Naka, Ibaraki (Japan)

    1997-08-01

    Two rigs of plutonium-uranium oxide (MOX) fuel rods have been irradiated in Halden boiling water reactor (HBWR) to investigate high burnup MOX fuel behavior for thermal reactor. The objective of irradiation tests is to investigate fuel behavior as influenced by pellet shape, pellet surface treatment, pellet-cladding gap size and MOX fuel powder preparations process. The two rigs have instrumentations for in-pile measurements of the fuel center-line temperature, plenum pressure, cladding elongation and fuel stack length change. The data, taken through in-operation instrumentation, have been analysed and compared with those from post-irradiation examination. The following observations are made: 1) PNC MOX fuels have achieved high burn-up as 59GWd/tMOX (67GWd/tM) at pellet peak without failure; 2) there was no significant difference in fission gas release fraction between PNC MOX fuels and UO{sub 2} fuels; 3) fission gas release from the co-converted fuel was lower than that from the mechanically blended fuel; 4) gap conductance was evaluated to decrease gradually with burn-up and to get stable in high burn-up region. 5) no evident difference of onset LHR for PCMI in experimental parameters (pellet shape and pellet-cladding gap size) was observed, but it decreased with burn-up. (author). 13 refs, 15 figs, 3 tabs.

  20. Code structure for U-Mo fuel performance analysis in high performance research reactor

    Jeong, Gwan Yoon; Cho, Tae Won; Lee, Chul Min; Sohn, Dong Seong [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Lee, Kyu Hong; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    A performance analysis modeling applicable to research reactor fuel is being developed with available models describing fuel performance phenomena observed from in-pile tests. We established the calculation algorithm and scheme to best predict fuel performance using radio-thermo-mechanically coupled system to consider fuel swelling, interaction layer growth, pore formation in the fuel meat, and creep fuel deformation and mass relocation, etc. In this paper, we present a general structure of the performance analysis code for typical research reactor fuel and advanced features such as a model to predict fuel failure induced by combination of breakaway swelling and pore growth in the fuel meat. Thermo-mechanical code dedicated to the modeling of U-Mo dispersion fuel plates is being under development in Korea to satisfy a demand for advanced performance analysis and safe assessment of the plates. The major physical phenomena during irradiation are considered in the code such that interaction layer formation by fuel-matrix interdiffusion, fission induced swelling of fuel particle, mass relocation by fission induced stress, and pore formation at the interface between the reaction product and Al matrix.

  1. Vitamin D and Heart Failure.

    Marshall Brinkley, D; Ali, Omair M; Zalawadiya, Sandip K; Wang, Thomas J

    2017-10-01

    Vitamin D is principally known for its role in calcium homeostasis, but preclinical studies implicate multiple pathways through which vitamin D may affect cardiovascular function and influence risk for heart failure. Many adults with cardiovascular disease have low vitamin D status, making it a potential therapeutic target. We review the rationale and potential role of vitamin D supplementation in the prevention and treatment of chronic heart failure. Substantial observational evidence has associated low vitamin D status with the risk of heart failure, ventricular remodeling, and clinical outcomes in heart failure, including mortality. However, trials assessing the influence of vitamin D supplementation on surrogate markers and clinical outcomes in heart failure have generally been small and inconclusive. There are insufficient data to recommend routine assessment or supplementation of vitamin D for the prevention or treatment of chronic heart failure. Prospective trials powered for clinical outcomes are warranted.

  2. U.S. report on fuel performance and technology

    Cook, T [Department of Energy, Washington, DC (United States). Office of Engineering and Technology Development

    1997-12-01

    The report reviews the following aspects of fuel performance and technology: increased demand on fuel performance;improved fuel failure rate; operating fuel cycles; capacity factor for US nuclear electric generating plants; potential reduction of SNF due to improved fuel burnup.

  3. Heart failure - tests

    CHF - tests; Congestive heart failure - tests; Cardiomyopathy - tests; HF - tests ... the best test to: Identify which type of heart failure (systolic, diastolic, valvular) Monitor your heart failure and ...

  4. X-ray cinematography on the nuclear fuel and cladding motion diagnostics

    Mizuta, Hiroshi; Uruwashi, Shinichi.

    1979-01-01

    X-ray cinematography has been used for monitoring fuel motion in the out-of-pile fuel pin joule melting experiments for nuclear, liquid metal cooled fast breeder reactor, safety studies related to fuel pin failure, initial fuel motion and thermal fuel-coolant interaction (FCI) of the hypothetical core distractive accident. In order to visually observe the nuclear fuel motion, the X-ray cinematography system consists of an X-ray source located about 5 cm from the test section and an image intensifier located at a corresponding position on the opposite side of the test section. The image from the image intensifier has been recorded both with a high speed camera and video recorder. (author)

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

    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

  6. LMFBR fuel analysis. Task A: Oxide fuel dynamics. Final report, October 1, 1976--September 30, 1977

    Dhir, V.K.; Doshi, J.; Frank, M.; Hauss, B.; Kastenberg, W.E.; Wong, K.

    1977-10-01

    The study presented deals with several areas of uncertainty in the analysis of the unprotected overpower transient for the Clinch River Breeder Reactor. These areas of uncertainty include the time, place, and mode of fuel pin failure; pre-failure fuel motion; fuel freezing, plugging, and plate-out following pin failure; and the potential for re-criticality. Internal molten fuel motion prior to pin failure was found to be sensitive to ramp rate and burnup. The strain-limit fuel failure criterion was found to be inappropriate for analysis based on existing data. The coupling of pre-transient- and transient-induced stresses tended to force the failure location towards the core midplane

  7. Study on the development of coating technology for UO{sub 2} nuclear fuel pellet and the microstructural observation of the coated layer

    Choi, Yong; Song, Moon Sup; Cho, In Sik; Kim Yu Sin; Lim Young Kyun [Sunmoon University, Asan (Korea)

    1998-04-01

    In order to enhance inherent safety of UO{sub 2} nuclear fuel pellet and develop future nuclear fuel technology, a coating method for the preparation multi-layers of pyrolytic carbon and silicon carbide on the fuel was developed. Inner pyrolytic carbon layer and outer silicon layer were prepared by thermal decomposition of propane in a fluidized bed type CVD unit and silane in ECR PECVD, respectively. Combustion reaction between two layers resulted in forming silicon carbide layer. The morphology depended on the initial carbon shape. Phase identification and microstructural analysis of the combustion product with XRD, AES, SEM and TEM showed that final products of inner layer and outer layer were pyrolytic carbon with isotropic structure and fine crystalline {beta}-SiC, respectively. This coating process is very useful for the fabrication of coated UO{sub 2} nuclear fuel pellet an future nuclear fuel fabrication technology. (author). 45 refs., 47 figs., 5 tabs.

  8. Heart failure - home monitoring

    ... this page: //medlineplus.gov/ency/patientinstructions/000113.htm Heart failure - home monitoring To use the sharing features on ... your high blood pressure Fast food tips Heart failure - discharge Heart failure - fluids and diuretics Heart failure - what to ...

  9. Fuel leak testing performance at NPP Jaslovske Bohunice

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

    1995-01-01

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

  10. Fuel leak testing performance at NPP Jaslovske Bohunice

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

    1996-12-31

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

  11. Potential for cladding thermal failure in LWRs during high temperature transients

    El Genk, M.S.

    1979-01-01

    The temperature increase in the fuel and the cladding during a PCM accident produces film boiling at the cladding surface which may induce zircaloy cladding failure, due to embrittlement, and fuel melting at the centerline of the fuel pellets. Molten fuel may extrude through radial cracks in the fuel and relocate in the fuel-cladding gap. Contact of extruded molten fuel with the cladding, which is at high temperature during film boiling, may induce cladding thermal failure due to melting. An assessment of central fuel melting and molten fuel extrusion into the fuel-cladding gap during a PCM accident is presented. The potential for thermal failure of the zircaloy cladding upon being contacted by molten fuel during such an accident is also analyzed and compared with the applicable experimental evidence

  12. Metal fuel safety performance

    Miles, K.J. Jr.; Tentner, A.M.

    1988-01-01

    The current development of breeder reactor systems has lead to the renewed interest in metal fuels as the driver material. Modeling efforts were begun to provide a mechanistic description of the metal fuel during anticipated and hypothetical transients within the context of the SAS4A accident analysis code system. Through validation exercises using experimental results of metal fuel TREAT tests, confidence is being developed on the nature and accuracy of the modeling and implementation. Prefailure characterization, transient pin response, margins to failure, axial in-pin fuel relocation prior to cladding breach, and molten fuel relocation after cladding breach are considered. Transient time scales ranging from milliseconds to many hours can be studied with all the reactivity feedbacks evaluated

  13. 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

    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)

  14. Conditioning of nuclear reactor fuel

    1975-01-01

    A method of conditioning the fuel of a nuclear reactor core to minimize failure of the fuel cladding comprising increasing the fuel rod power to a desired maximum power level at a rate below a critical rate which would cause cladding damage is given. Such conditioning allows subsequent freedom of power changes below and up to said maximum power level with minimized danger of cladding damage. (Auth.)

  15. Behavior of irradiated ATR/MOX fuel under reactivity initiated accident conditions (Joint research)

    Sasajima, Hideo; Fuketa, Toyoshi; Nakamura, Takehiko; Nakamura, Jinichi; Uetsuka, Hiroshi

    2000-03-01

    Pulse irradiation experiments with irradiated ATR/MOX fuel rods of 20 MWd/kgHM were conducted at the NSRR in JAERI to study the transient behavior of MOX fuel rod under reactivity initiated accident conditions. Four pulse irradiation experiments were performed with peak fuel enthalpy ranging from 335 J/g to 586 J/g, resulted in no failure of fuel rods. Deformation of the fuel rods due to PCMI occurred in the experiments with peak fuel enthalpy above 500 J/g. Significant fission gas release up to 20% was measured by rod puncture measurement. The generation of fine radial cracks in pellet periphery, micro-cracks and boundary separation over the entire region of pellet were observed. These microstructure changes might contribute to the swelling of fuel pellets during the pulse irradiation. This could cause the large radial deformation of fuel rod and high fission gas release when the pulse irradiation conducted at relatively high peak fuel enthalpy. In addition, fine grain structures around the plutonium spot and cauliflower structure in cavity of the plutonium spot were observed in the outer region of the fuel pellet. (author)

  16. Reliability assessment of the fueling machine of the CANDU reactor

    Al-Kusayer, T.A.

    1985-01-01

    Fueling of CANDU-reactors is carried out by two fueling machines, each serving one end of the reactor. The fueling machine becomes a part of the primary heat transport system during the refueling operations, and hence, some refueling machine malfunctions could result in a small scale-loss-of-coolant accident. Fueling machine failures and the failure sequences are discussed. The unavailability of the fueling machine is estimated by using fault tree analysis. The probability of mechanical failure of the fueling machine interface is estimated as 1.08 x 10 -5 . (orig.) [de

  17. Rethinking Heart Failure

    F?rstenwerth, Hauke

    2012-01-01

    An increasing body of clinical observations and experimental evidence suggests that cardiac dysfunction results from autonomic dysregulation of the contractile output of the heart. Excessive activation of the sympathetic nervous system and a decrease in parasympathetic tone are associated with increased mortality. Elevated levels of circulating catecholamines closely correlate with the severity and poor prognosis in heart failure. Sympathetic over-stimulation causes increased levels of catech...

  18. ABB high burnup fuel

    Andersson, S.; Helmersson, S.; Nilsson, S.; Jourdain, P.; Karlsson, L.; Limback, M.; Garde, A.M.

    1999-01-01

    Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both PWR and BWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter proven to meet the 6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10 x 10 fuel, where ABB is the only vendor to date with batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of PWR and BWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its utility customers. This paper provides an overview of recent fuel performance and reliability experience at ABB. Selected development and validation activities for PWR and BWR fuel are presented, for which the ABB test facilities in Windsor (TF-2 loop, mechanical test laboratory) and Vaesteras (FRIGG, BURE) are essential. (authors)

  19. Failure modes of composite sandwich beams

    Gdoutos E.

    2008-01-01

    Full Text Available A thorough investigation of failure behavior of composite sandwich beams under three-and four-point bending was undertaken. The beams were made of unidirectional carbon/epoxy facings and a PVC closed-cell foam core. The constituent materials were fully characterized and in the case of the foam core, failure envelopes were developed for general two-dimensional states of stress. Various failure modes including facing wrinkling, indentation failure and core failure were observed and compared with analytical predictions. The initiation, propagation and interaction of failure modes depend on the type of loading, constituent material properties and geometrical dimensions.

  20. Release of segregated nuclides from spent fuel

    Johnson, L.H.; Tait, J.C. [Atomic Energy Canada Ltd., Pinawa, MB (Canada). Whiteshell Laboratories

    1997-10-01

    The potential release of fission and activation products from spent nuclear fuel into groundwater after container failure in the Swedish deep repository is discussed. Data from studies of fission gas release from representative Swedish BWR fuel are used to estimate the average fission gas release for the spent fuel population. Information from a variety of leaching studies on LWR and CANDU fuel are then reviewed as a basis for estimating the fraction of the inventory of key radionuclides that could be released preferentially (the Instant Release Fraction of IRF) upon failure of the fuel cladding. The uncertainties associated with these estimates are discussed. 33 refs, 6 figs, 3 tabs.

  1. Rationale, Design, and Methodology of the APOLLON trial: A comPrehensive, ObservationaL registry of heart faiLure with midrange and preserved ejectiON fraction.

    Özlek, Bülent; Özlek, Eda; Çelik, Oğuzhan; Çil, Cem; Doğan, Volkan; Tekinalp, Mehmet; Zencirkıran Ağuş, Hicaz; Kahraman, Serkan; Ösken, Altuğ; Rencüzoğulları, İbrahim; Tanık, Veysel Ozan; Bekar, Lütfü; Çakır, Mustafa Ozan; Kaya, Bedri Caner; Tibilli, Hakan; Çelik, Yunus; Başaran, Özcan; Mert, Kadir Uğur; Sevinç, Samet; Demirci, Erkan; Dondurmacı, Engin; Biteker, Murat

    2018-05-01

    Although almost half of chronic heart failure (HF) patients have mid-range (HFmrEF) and preserved left-ventricular ejection fraction (HFpEF), no studies have been carried out with these patients in our country. This study aims to determine the demographic characteristics and current status of the clinical background of HFmrEF and HFpEF patients in a multicenter trial. A comPrehensive, ObservationaL registry of heart faiLure with mid range and preserved ejectiON fraction (APOLLON) trial will be an observational, multicenter, and noninterventional study conducted in Turkey. The study population will include 1065 patients from 12 sites in Turkey. All data will be collected at one point in time and the current clinical practice will be evaluated (ClinicalTrials.gov number NCT03026114). We will enroll all consecutive patients admitted to the cardiology clinics who were at least 18 years of age and had New York Heart Association class II, III, or IV HF, elevated brain natriuretic peptide levels within the last 30 days, and an left ventricular ejection fraction (LVEF) of at least 40%. Patients fulfilling the exclusion criteria will not be included in the study. Patients will be stratified into two categories according to LVEF: mid-range EF (HFmrEF, LVEF 40%-49%) and preserved EF (HFpEF, LVEF ≥50%). Regional quota sampling will be performed to ensure that the sample was representative of the Turkish population. Demographic, lifestyle, medical, and therapeutic data will be collected by this specific survey. The APOLLON trial will be the largest and most comprehensive study in Turkey evaluating HF patients with a LVEF ≥40% and will also be the first study to specifically analyze the recently designated HFmrEF category.

  2. Diagnosis of Equipment Failures by Pattern Recognition

    Pau, L. F.

    1974-01-01

    The main problems in relation to automatic fault finding and diagnosis in equipments or production systems are discussed: 1) compression of the syndrome and observation spaces for better discrimination between failure modes; 2) simultaneous display of the failure patterns and the failure instants...

  3. Fuel assemblies

    Mukai, Hideyuki

    1987-01-01

    Purpose: To prevent bending of fuel rods caused by the difference of irradiation growth between coupling fuel rods and standards fuel rods thereby maintain the fuel rod integrity. Constitution: The f value for a fuel can (the ratio of pole of zirconium crystals in the entire crystals along the axial direction of the fuel can) of a coupling fuel rod secured by upper and lower tie plates is made smaller than the f value for the fuel can of a standard fuel rod not secured by the upper and the lower tie plates. This can make the irradiation growth of the fuel can of the coupling fuel rod greater than the irradiation growth of the fuel can of the standard fuel rod and, accordingly, since the elongation of the standard fuel rod can always by made greater, bending of the standard fuel rod can be prevented. (Yoshihara, M.)

  4. Modeling of the PWR fuel mechanical behaviour and particularly study of the pellet-cladding interaction in a fuel rod

    Hourdequin, N.

    1995-05-01

    In Pressurized Water Reactor (PWR) power plants, fuel cladding constitutes the first containment barrier against radioactive contamination. Computer codes, developed with the help of a large experimental knowledge, try to predict cladding failures which must be limited in order to maintain a maximal safety level. Until now, fuel rod design calculus with unidimensional codes were adequate to prevent cladding failures in standard PWR's operating conditions. But now, the need of nuclear power plant availability increases. That leads to more constraining operating condition in which cladding failures are strongly influenced by the fuel rod mechanical behaviour, mainly at high power level. Then, the pellet-cladding interaction (PCI) becomes important, and is characterized by local effects which description expects a multidimensional modelization. This is the aim of the TOUTATIS 2D-3D code, that this thesis contributes to develop. This code allows to predict non-axisymmetric behaviour too, as rod buckling which has been observed in some irradiation experiments and identified with the help of TOUTATIS. By another way, PCI is influenced by under irradiation experiments and identified with the help of TOUTATIS which includes a densification model and a swelling model. The latter can only be used in standard operating conditions. However, the processing structure of this modulus provides the possibility to include any type of model corresponding with other operating conditions. In last, we show the result of these fuel volume variations on the cladding mechanical conditions. (author). 25 refs., 89 figs., 2 tabs., 12 photos., 5 appends

  5. Fuel safety research 2001

    Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-11-01

    The Fuel Safety Research Laboratory is in charge of research activity which covers almost research items related to fuel safety of water reactor in JAERI. Various types of experimental and analytical researches are being conducted by using some unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and the Reactor Fuel Examination Facility (RFEF) of JAERI. The research to confirm the safety of high burn-up fuel and MOX fuel under accident conditions is the most important item among them. The laboratory consists of following five research groups corresponding to each research fields; Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). Research group of fuel behavior analysis (FEMAXI group). Research group of radionuclides release and transport behavior from irradiated fuel under severe accident conditions (VEGA group). The research conducted in the year 2001 produced many important data and information. They are, for example, the fuel behavior data under BWR power oscillation conditions in the NSRR, the data on failure-bearing capability of hydrided cladding under LOCA conditions and the FP release data at very high temperature in steam which simulate the reactor core condition during severe accidents. This report summarizes the outline of research activities and major outcomes of the research executed in 2001 in the Fuel Safety Research Laboratory. (author)

  6. Postirradiation results and evaluation of helium-bonded uranium--plutonium carbide fuel elements irradiated in EBR-II. Interim report

    Latimer, T.W.; Barner, J.O.; Kerrisk, J.F.; Green, J.L.

    1976-02-01

    An evaluation was made of the performance of 74 helium-bonded uranium-plutonium carbide fuel elements that were irradiated in EBR-II at 38-96 kW/m to 2-12 at. percent burnup. Only 38 of these elements have completed postirradiation examination. The higher failure rate found in fuel elements which contained high-density (greater than 95 percent theoretical density) fuel than those which contained low-density (77-91 percent theoretical density) fuel was attributed to the limited ability of the high-density fuel to swell into the void space provided in the fuel element. Increasing cladding thickness and original fuel-cladding gap size were both found to influence the failure rates for elements containing low-density fuel. Lower cladding strain and higher fission-gas release were found in high-burnup fuel elements having smear densities of less than 81 percent. Fission-gas release was usually less than 5 percent for high-density fuel, but increased with burnup to a maximum of 37 percent in low-density fuel. Maximum carburization in elements attaining 5-10 at. percent burnup and clad in Types 304 or 316 stainless steel and Incoloy 800 ranged from 36-80 μm and 38-52 μm, respectively. Strontium and barium were the fission products most frequently found in contact with the cladding but no penetration of the cladding by uranium, plutonium, or fission products was observed

  7. Failed fuel detection device

    Doi, Akira.

    1994-01-01

    The device of the present invention concerns a failed fuel detection device for a nuclear reactor, such as an FBR type reactor, using electroconductive coolants. A sampling port is disposed at the upper portion of the fuel assembly so as to cover the assembly, so that coolants in the fuel assembly are sampled to improve a device for detecting fuel failure. That is, when coolants in the fuel assembly are sampled from the sampling port, the flow of electroconductive coolants in an sampling tube is detected by a flowmeter, to control an electromagnetic pump. The flow of electroconductive coolants is stopped against the waterhead pressure and dynamic pressure of the conductive coolants, and a predetermined amount of the coolants is pumped up to the sampling tank. Gas is supplied to the pumped up coolants so that fissile products are transferred from the coolants to a gas phase. Radiation in the gas in a gas recycling system is measured to detect presence of fuel failure. (I.S.)

  8. Temperature and Burnup Correlated FCCI in U-10Zr Metallic Fuel

    William J. Carmack

    2012-05-01

    Metallic fuels are proposed for use in advanced sodium cooled fast reactors. The experience basis for metallic fuels is extensive and includes development and qualification of fuels for the Experimental Breeder Reactor I, the Experimental Breeder Reactor II, FERMI-I, and the Fast Flux Test Facility (FFTF) reactors. Metallic fuels provide a number of advantages over other fuel types in terms of fabricability, performance, recyclability, and safety. Key to the performance of all nuclear fuel systems is the resistance to “breach” and subsequent release of fission products and fuel constituents to the primary coolant system of the nuclear power plant. In metallic fuel, the experience is that significant fuel-cladding chemical (FCCI) interaction occurs and becomes prevalent at high power-high temperature operation and ultimately leads to fuel pin breach and failure. Empirical relationships for metallic fuel pin failure have been developed from a large body of in-pile and out of pile research, development, and experimentation. It has been found that significant in-pile acceleration of the FCCI rate is experienced over similar condition out-of-pile experiments. The study of FCCI in metallic fuels has led to the quantification of in-pile failure rates to establish an empirical time and temperature dependent failure limit for fuel elements. Up until now the understanding of FCCI layer formation has been limited to data generated in EBR-II experiments. This dissertation provides new FCCI data extracted from the MFF-series of metallic fuel irradiations performed in the FFTF. These fuel assemblies contain valuable information on the formation of FCCI in metallic fuels at a variety of temperature and burnup conditions and in fuel with axial fuel height three times longer than EBR-II experiments. The longer fuel column in the FFTF and the fuel pins examined have significantly different flux, power, temperature, and FCCI profiles than that found in similar tests conducted in

  9. Economic consequences of QA and QC in fuel and fuel assembly production

    Strasser, A.A.

    1984-01-01

    The planning of quality control and quality assurance programs for fuel fabrication must balance the cost of the programs, their effectiveness, and the economic consequences of failure to meet the product specifications. The cost of fuel failures can be very high in comparison to the cost of quality control, and this provides considerable economic justification for increasing the level of quality control if its effectiveness in reducing failure potential can be demonstrated. Typical costs and examples are discussed. (orig.)

  10. Development of advanced neutron radiography for inspection on irradiated fuels and materials (2). Observation of hydride and oxide film on zircaloy cladding by using neutron radiography

    Yasuda, Ryou; Nakata, Masahito; Mastubayashi, Masahito; Harada, Katsuya

    2001-02-01

    Neutron radiography has been used as available diagnosis method of integrity on irradiated fuels, and has not been employed for estimating hydride and oxide film, which are influenced on integrity of Zircaloy cladding. Preliminary tests for PIE were carried out to assess possibility of neutron radiography as evaluation tool for hydrided and oxide film on the cladding. In these experiments, Zircaloy claddings with controlled amount of hydrogen absorption (200, 500, and 1000ppm) and thickness of oxide film were radiographed in center axis and in side directions of cladding tube by neutron imaging plate method. It is noted that thickness of oxide film was formed range from 7 μ m to 70 μ m at various temperatures (973, 1173, and 1323K) under steam atmosphere on the Zircaloy claddings. CT (Computed Tomography) restructure calculation was carried out to obtain cross section image of the claddings non-destructively. The Radiographs were qualitatively investigated about structure formation area and dependence of hydrogen absorption amount on PSL (Photo Simulated Luminescence) and CT values using by image analysis processor. At the results of imaging plate test, obvious difference was not found out between hydride formation (except for 1000ppm cladding) and standard claddings in side direction image. However, on the center axis direction image, outer circumference in the cladding cross-section that corresponded with hydride segregation area became blacker. In the case of oxide film formed cladding images, although oxide film could not find out on all speciments in the radiographs taken at the center axis and side directions, cross-section of claddings heat-processed at 973K showed appreciable blackness increasing with oxide film thickness on the radiographs. On the other hand, there is no effective difference between images of oxide film formed claddings processed at 1173K and 1323K and that of standard cladding. In CT image of 1000ppm hydrogen absorbed cladding, it is

  11. Fuel safety research 2000

    Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-03-01

    In April 1999, the Fuel Safety Research Laboratory was newly established as a part of reorganization of the Nuclear Safety Research Center, JAERI. The new laboratory was organized by combining three pre-existing laboratories, Reactivity Accident Laboratory, Fuel Reliability Laboratory, and a part of Severe Accident Research Laboratory. The Fuel Safety Research Laboratory becomes to be in charge of all fuel safety research in JAERI. Various experimental and analytical researches are conducted in the laboratory by using the unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and hot cells in JAERI. The laboratory consists of following five research groups corresponding to each research fields; (a) Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). (b) Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). (c) Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). (d) Research group of fuel behavior analysis (FEMAXI group). (e) Research group of FP release/transport behavior from irradiated fuel (VEGA group). The research activities in year 2000 produced many important data and information. They are, for example, failure of high burnup BWR fuel rod under RIA conditions, data on the behavior of hydrided Zircaloy cladding under LOCA conditions and FP release data from VEGA experiments at very high temperature/pressure condition. This report summarizes the outline of research activities and major outcomes of the research executed in 2000 in the Fuel Safety Research Laboratory. (author)

  12. Test system to simulate transient overpower LMFBR cladding failure

    Barrus, H.G.; Feigenbutz, L.V.

    1981-01-01

    One of the HEDL programs has the objective to experimentally characterize fuel pin cladding failure due to cladding rupture or ripping. A new test system has been developed which simulates a transient mechanically-loaded fuel pin failure. In this new system the mechanical load is prototypic of a fuel pellet rapidly expanding against the cladding due to various causes such as fuel thermal expansion, fuel melting, and fuel swelling. This new test system is called the Fuel Cladding Mechanical Interaction Mandrel Loading Test (FCMI/MLT). The FCMI/MLT test system and the method used to rupture cladding specimens very rapidly to simulate a transient event are described. Also described is the automatic data acquisition and control system which is required to control the startup, operation and shutdown of the very fast tests, and needed to acquire and store large quantities of data in a short time

  13. Safety assessment for the CANFLEX-NU fuel bundles with respect to the 37-element fuel bundles

    Suk, H. C.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-11-01

    The KAERI and AECL have jointly developed an advanced CANDU fuel, called CANFLEX-NU fuel bundle. CANFLEX 43-element bundle has some improved features of increased operating margin and enhanced safety compared to the existing 37-element bundle. Since CANFLEX fuel bundle is designed to be compatible with the CANDU-6 reactor design, the behaviour in the thermalhydraulic system will be nearly identical with 37-element bundle. But due to different element design and linear element power distribution between the two bundles, it is expected that CANFLEX fuel behaviour would be different from the behaviour of the 37-element fuel. Therefore, safety assessments on the design basis accidents which result if fuel failures are performed. For all accidents selected, it is observed that the loading of CANFLEX bundle in an existing CANDU-6 reactor would not worsen the reactor safety. It is also predicted that fission product release for CANFLEX fuel bundle generally is lower than that for 37-element bundle. 3 refs., 2 figs., 2 tabs. (Author)

  14. 14 CFR 25.981 - Fuel tank ignition prevention.

    2010-01-01

    ... AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System § 25.981 Fuel tank... system where catastrophic failure could occur due to ignition of fuel or vapors. This must be shown by... established, as necessary, to prevent development of ignition sources within the fuel tank system pursuant to...

  15. Worldwide experience with light water reactor fuel - a review

    Strasser, A.A.

    1986-01-01

    Continued attention to fuel performance has over the years improved fuel reliability and reduced fuel related failures. But further improvements can still be made by increased attention to reactor operating and maintenance methods, as well as to quality control during fuel fabrication. (author)

  16. Fact sheet on fuel manufacturing and utilization

    2006-01-01

    The Nuclear Fuel Cycle and Materials Section (NFCMS) supports Member States to improve in-pile fuel performance and management of materials; and to develop advanced fuel technologies for ensuring reliability and economic efficiency of the nuclear fuel cycle, provides assistance to Member States to support fuel-manufacturing capability, including quality assurance techniques, optimization of manufacturing parameters and radiation protection, supports the development fuel modeling expertise in Member States, covering both normal operation and postulated and severe accident conditions, provides information and support for the operation of Nuclear Power Plant to ensure that the environment and water chemistry is appropriate for fuel operation, supports fuel failure investigations, including equipment for failed fuel detection and for post-irradiation examination and inspection, as well as fuel repair, provides information and support research into the basic properties of fuel materials, including UO2, MOX, (Th, Pu)O2, (Th, U233)O2 fuels and zirconium alloy cladding and fuel assembly components and offers guidance on the relationship with back-end requirement (interim storage, transport, reprocessing, disposal), fuel utilization and management, MOX fuels, alternative fuels and advanced fuel technology and materials, economic and other aspects of nuclear fuel use (e.g. environmental impact). Recently NFCMS provided support to a Member State manufacturing Gadolinia doped fuel and provided in-mast sipping equipment to a Nuclear Power Plant to allow the determination of fuel failure. Member States interested in fuel performance and manufacture should contact the Technical Cooperation Department of the Agency and Member States interested in knowing more about the Agency's programme on source management should contact: C. Ganguly, Section Head, V. Inozemtsev, J. Killeen

  17. The summary of WWER-1000 fuel utilization in Ukraine

    Afanasyev, A [Ukrainian State Committee on Nuclear Power Utilization, Kiev (Ukraine)

    1997-12-01

    The report discusses the status of the fuel and fuel cycles of WWER-1000 reactors in Ukraine. The major reasons that caused the Ukrainian utilities to overcome the conservative design solutions in order to improve fuel utilization and extend fuel burnup are shown. At the same time the sufficient fuel reliability and fuel cycle flexibility are ensured. The burnup distribution in the unloaded fuel assemblies and average fuel rod failure rate are presented. The questions of reactor core operation safety and the economical problems of the front end of the fuel cycle are also considered. (author). 2 refs, 3 figs, 4 tabs.

  18. Post-pulse detail metallographic examinations of low-enriched uranium silicide plate-type miniature fuel

    Yanagisawa, Kazuaki

    1991-10-01

    Pulse irradiation at Nuclear Safety Research Reactor (NSRR) was performed using low-enriched (19.89 w% 235 U) unirradiated silicide plate-type miniature fuel which had a density of 4.8 gU/cm 3 . Experimental aims are to understand the dimensional stability and to clarify the failure threshold of the silicide plate-type miniature fuel under power transient conditions through post-pulse detail metallographic examinations. A silicide plate-type miniature fuel was loaded into an irradiation capsule and irradiated by a single pulse. Deposited energies given in the experiments were 62, 77, 116 and 154 cal/g·fuel, which lead to corresponding peak fuel plate temperatures, 201 ± 28degC, 187 ± 10degC, 418 ± 74degC and 871 ± 74degC, respectively. Below 400degC, reliability and dimensional stability of the silicide plate fuel was sustained, and the silicide plate fuel was intact. Up to 540degC, wall-through intergranular crackings occurred in the Al-3%Mg alloy cladding. With the increase of the temperature, the melting of the aluminum cladding followed by recrystallization, the denudation of fuel core and the plate-through intergranular cracking were observed. With the increase of the temperature beyond 400degC, the bowing of fuel plate became significant. Above the temperature of 640degC molten aluminum partially reacted with the fuel core, partially flowed downward under the influence of surface tension and gravity, and partially formed agglomerations. Judging from these experimental observations, the fuel-plate above 400degC tends to reduce its dimensional stability. Despite of the apparent silicide fuel-plate failure, neither generation of pressure pulse nor that of mechanical energy occurred at all. (J.P.N.)

  19. Heart failure - medicines

    CHF - medicines; Congestive heart failure - medicines; Cardiomyopathy - medicines; HF - medicines ... You will need to take most of your heart failure medicines every day. Some medicines are taken ...

  20. Post-irradiation examinations of uranium-plutonium mixed nitride fuel irradiated in JMTR (89F-3A capsule)

    Iwai, Takashi; Nakajima, Kunihisa; Kikuchi, Hironobu; Arai, Yasuo; Kimura, Yasuhiko; Nagashima, Hisao; Sekita, Noriaki

    2000-03-01

    Two helium-bonded fuel pins filled with uranium-plutonium mixed nitride pellets were encapsulated in 89F-3A and irradiated in JMTR up to 5.5% FIMA at a maximum linear power of 73 kW/m. The capsule cooled for ∼5 months was transported to Reactor Fuel Examination Facility and subjected to non-destructive and destructive post irradiation examinations. Any failure was not observed in the irradiated fuel pins. Very low fission gas release rate of about 2 ∼ 3% was observed, while the diametric increase of fuel pin was limited to ∼0.4% at the position of maximum reading. The inner surface of cladding tube did not show any signs of chemical interaction with fuel pellet. (author)

  1. Microstructural characteristics of HIP-bonded monolithic nuclear fuels with a diffusion barrier

    Jue, Jan-Fong; Keiser, Dennis D.; Breckenridge, Cynthia R.; Moore, Glenn A.; Meyer, Mitchell K.

    2014-05-01

    Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative is developing an advanced monolithic fuel to convert US high-performance research reactors to low-enriched uranium. Hot-isostatic-press (HIP) bonding was the single process down-selected to bond monolithic U-Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U-Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between the fuel meat, the cladding, and the diffusion barrier, as well as between the U-10Mo fuel meat and the Al-6061 cladding, were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are: line. Some of these attributes might be critical to the irradiation performance of monolithic U-10Mo nuclear fuel. There are several issues or concerns that warrant more detailed study, such as precipitation along the cladding-to-cladding bond line, chemical banding, uncovered fuel-zone edge, and the interaction layer between the U-Mo fuel meat and zirconium. Future post-irradiation examination results will focus, among other things, on identifying in-reactor failure mechanisms and, eventually, directing further fresh fuel characterization efforts.

  2. Development of MOX fuel database

    Ikusawa, Yoshihisa; Ozawa, Takayuki

    2007-03-01

    We developed MOX Fuel Database, which included valuable data from several irradiation tests in FUGEN and Halden reactor, for help of LWR MOX use. This database includes the data of fabrication and irradiation, and the results of post-irradiation examinations for seven fuel assemblies, i.e. P06, P2R, E03, E06, E07, E08 and E09, irradiated in FUGEN. The highest pellet peak burn-up reached ∼48GWd/t in MOX fuels, of which the maximum plutonium content was ∼6 wt%, irradiated in E09 fuel assembly without any failure. Also the data from the instrumented MOX fuels irradiated in HBWR to study the irradiation behavior of BWR MOX fuels under the steady state condition (IFA-514/565 and IFA-529), under the load-follow operation condition (IFA-554/555) and under the transit condition (IFA-591) are included in this database. The highest assembly burn-up reached ∼56 GWd/t in IFA-565 steady state irradiation test, and the maximum linear power of MOX fuel rods was 58.3-68.4 kW/m without any failure in IFA-591 ramp test. In addition, valuable instrument data, i.e. cladding elongation, fuel stack elongation, fuel center temperature and rod inner pressure were obtained from IFA-554/555 load-follow test. (author)

  3. Alternative Fuels

    Alternative fuels include gaseous fuels such as hydrogen, natural gas, and propane; alcohols such as ethanol, methanol, and butanol; vegetable and waste-derived oils; and electricity. Overview of alternative fuels is here.

  4. Recent metal fuel safety tests in TREAT

    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

  5. Failure Modes of thin supported Membranes

    Hendriksen, Peter Vang; Høgsberg, J.R.; Kjeldsen, Ane Mette

    2007-01-01

    Four different failure modes relevant to tubular supported membranes (thin dense films on a thick porous support) were analyzed. The failure modes were: 1) Structural collapse due to external pressure 2) burst of locally unsupported areas, 3) formation of surface cracks in the membrane due to TEC......-mismatches, and finally 4) delamination between membrane and support due to expansion of the membrane on use. Design criteria to minimize risk of failure by the four different modes are discussed. The theoretical analysis of the two last failure modes is compared to failures observed on actual components....

  6. Nuclear fuel rods

    Wada, Toyoji.

    1979-01-01

    Purpose: To remove failures caused from combination of fuel-cladding interactions, hydrogen absorptions, stress corrosions or the likes by setting the quantity ratio of uranium or uranium and plutonium relative to oxygen to a specific range in fuel pellets and forming a specific size of a through hole at the center of the pellets. Constitution: In a fuel rods of a structure wherein fuel pellets prepared by compacting and sintering uranium dioxide, or oxide mixture consisting of oxides of plutonium and uranium are sealed with a zirconium metal can, the ratio of uranium or uranium and plutonium to oxygen is specified as 1 : 2.01 - 1 : 2.05 in the can and a passing hole of a size in the range of 15 - 30% of the outer diameter of the fuel pellet is formed at the center of the pellet. This increases the oxygen partial pressure in the fuel rod, oxidizes and forms a protection layer on the inner surface of the can to control the hydrogen absorption and stress corrosion. Locallized stress due to fuel cladding interaction (PCMI) can also be moderated. (Horiuchi, T.)

  7. Fuel assembly

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

    1998-01-01

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

  8. Fuel assembly

    Yamazaki, Hajime.

    1995-01-01

    In a fuel assembly having fuel rods of different length, fuel pellets of mixed oxides of uranium and plutonium are loaded to a short fuel rod. The volume ratio of a pellet-loaded portion to a plenum portion of the short fuel rod is made greater than the volume ratio of a fuel rod to which uranium fuel pellets are loaded. In addition, the volume of the plenum portion of the short fuel rod is set greater depending on the plutonium content in the loaded fuel pellets. MOX fuel pellets are loaded on the short fuel rods having a greater degree of freedom relevant to the setting for the volume of the plenum portion compared with that of a long rod fuel, and the volume of the plenum portion is ensured greater depending on the plutonium content. Even if a large amount of FP gas and He gas are discharged from the MOX fuels compared with that from the uranium fuels, the internal pressure of the MOX fuel rod during operation is maintained substantially identical with that of the uranium fuel rod, so that a risk of generating excess stresses applied to the fuel cladding tubes and rupture of fuels are greatly reduced. (N.H.)

  9. On the observation of a huge lattice contraction and crystal habit modifications in LiMn2O4 prepared by a fuel assisted solution combustion

    Ragavendran, K.; Sherwood, D.; Vasudevan, D.; Emmanuel, Bosco

    2009-01-01

    Two batches of poly-crystalline lithium manganate were prepared by a fuel assisted solution combustion method. LiMn 2 O 4 (S) was prepared using starch as the fuel and LiMn 2 O 4 (P) was prepared using poly vinyl alcohol (PVA) as the fuel. XRD studies indicated a significant and consistent shift in the 2θ values of all the hkl peaks to higher values in LiMn 2 O 4 (P) compared to LiMn 2 O 4 (S) indicating a lattice contraction in the former. TG/DTA studies indicated a higher formation temperature (∼25 deg. C higher) for LiMn 2 O 4 (P). The higher formation temperature most likely promotes the oxidation of some Mn 3+ to Mn 4+ with a lower ionic radius causing a lattice contraction. This hypothesis is confirmed through XPS studies which indicated the presence of a higher fraction of Mn 4+ in LiMn 2 O 4 (P) than that present in LiMn 2 O 4 (S). A crystal shape algorithm was used to generate the crystal habits of lithium manganate from their XRD data leading to an understanding on the exposed hkl planes in these materials. From the atomic arrangement on the exposed hkl planes it is predicted that LiMn 2 O 4 (P) would be less prone to manganese dissolution and hence would possess a higher cycle life when compared to LiMn 2 O 4 (S).

  10. On the observation of a huge lattice contraction and crystal habit modifications in LiMn 2O 4 prepared by a fuel assisted solution combustion

    Ragavendran, K.; Sherwood, D.; Vasudevan, D.; Emmanuel, Bosco

    2009-08-01

    Two batches of poly-crystalline lithium manganate were prepared by a fuel assisted solution combustion method. LiMn 2O 4(S) was prepared using starch as the fuel and LiMn 2O 4(P) was prepared using poly vinyl alcohol (PVA) as the fuel. XRD studies indicated a significant and consistent shift in the 2 θ values of all the hkl peaks to higher values in LiMn 2O 4(P) compared to LiMn 2O 4(S) indicating a lattice contraction in the former. TG/DTA studies indicated a higher formation temperature (∼25 °C higher) for LiMn 2O 4(P). The higher formation temperature most likely promotes the oxidation of some Mn 3+ to Mn 4+ with a lower ionic radius causing a lattice contraction. This hypothesis is confirmed through XPS studies which indicated the presence of a higher fraction of Mn 4+ in LiMn 2O 4(P) than that present in LiMn 2O 4(S). A crystal shape algorithm was used to generate the crystal habits of lithium manganate from their XRD data leading to an understanding on the exposed hkl planes in these materials. From the atomic arrangement on the exposed hkl planes it is predicted that LiMn 2O 4(P) would be less prone to manganese dissolution and hence would possess a higher cycle life when compared to LiMn 2O 4(S).

  11. Detection and location of leaking TRIGA fuel elements

    Bouchey, G.D.; Gage, S.J.

    1970-01-01

    Several TRIGA facilities have experienced difficulty resulting from cladding failures of aluminum clad TRIGA fuel elements. Recently, at the University of Texas at Austin reactor facility, fission product releases were observed during 250 kW operation and were attributed to a leaking fuel element. A rather extensive testing program has been undertaken to locate the faulty element. The used sniffer device is described, which provides a quick, easily constructed, and extremely sensitive means of locating leaking fuel elements. The difficulty at The University of Texas was compounded by extremely low levels and the sporadic nature of the releases. However, in the more typical situation, in which a faulty element consistently releases relatively large quantities of fission gas, such a device should locate the leak with little difficulty

  12. PWR and WWER fuel performance. A comparison of major characteristics

    Weidinger, H.

    2006-01-01

    PWR and WWER fuel technologies have the same basic performance targets: most effective use of the energy stored in the fuel and highest possible reliability. Both fuel technologies use basically the same strategies to reach these targets: 1) Optimized reload strategies; 2) Maximal use of structural material with low neutron cross sections; 3) Decrease the fuel failure frequency towards a 'zero failure' performance by understanding and eliminating the root causes of those defects. The key driving force of the technology of both, PWR and WWER fuel is high burn-up. Presently a range of 45 - 50 MWD/kgU have been reached commercially for PWR and WWER fuel. The main technical limitations to reach high burn-up are typically different for PWR and WWER fuel: for PWR fuel it is the corrosion and hydrogen uptake of the Zr-based materials; for WWER fuel it is the mechanical and dimensional stability of the FA (and the whole core). Corrosion and hydrogen uptake of Zr-materials is a 'non-problem' for WWER fuel. Other performance criteria that are important for high burn-up are the creep and growth behaviour of the Zr materials and the fission gas release in the fuel rod. There exists a good and broad data base to model and design both fuel types. FA and fuel rod vibration appears to be a generic problem for both fuel types but with more evidence for PWR fuel performance reliability. Grid-to-rod fretting is still a major issue in the fuel failure statistics of PWR fuel. Fuel rod cladding defects by debris fretting is no longer a key problem for PWR fuel, while it still appears to be a significant root cause for WWER fuel failures. 'Zero defect' fuel performance is achievable with a high probability, as statistics for US PWR and WWER-1000 fuel has shown

  13. Advanced Heart Failure

    ... Artery Disease Venous Thromboembolism Aortic Aneurysm More Advanced Heart Failure Updated:May 9,2017 When heart failure (HF) ... Making This content was last reviewed May 2017. Heart Failure • Home • About Heart Failure • Causes and Risks for ...

  14. Experimental assessment of fuel-cladding interactions

    Wood, Elizabeth Sooby [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-06-29

    A range of fuel concepts designed to better tolerate accident scenarios and reactor transients are currently undergoing fundamental development at national laboratories as well as university and industrial partners. Pellet-clad mechanical and chemical interaction can be expected to affect fuel failure rates experienced during steady state operation, as well as dramatically impact the response of the fuel form under loss of coolant and other accident scenarios. The importance of this aspect of fuel design prompted research initiated by AFC in FY14 to begin exploratory efforts to characterize this phenomenon for candidate fuelcladding systems of immediate interest. Continued efforts in FY15 and FY17 aimed to better understand and simulate initial pellet-clad interaction with little-to-no pressure on the pellet-clad interface. Reported here are the results from 1000 h heat treatments at 400, 500, and 600°C of diffusion couples pairing UN with a FeCrAl alloy, SiC, and Zr-based cladding candidate sealed in evacuated quartz ampoules. No gross reactions were observed, though trace elemental contaminants were identified.

  15. Irradiation performance of HTGR recycle fissile fuel

    Homan, F.J.; Long, E.L. Jr.

    1976-08-01

    The irradiation performance of candidate HTGR recycle fissile fuel under accelerated testing conditions is reviewed. Failure modes for coated-particle fuels are described, and the performance of candidate recycle fissile fuels is discussed in terms of these failure modes. The bases on which UO 2 and (Th,U)O 2 were rejected as candidate recycle fissile fuels are outlined, along with the bases on which the weak-acid resin (WAR)-derived fissile fuel was selected as the reference recycle kernel. Comparisons are made relative to the irradiation behavior of WAR-derived fuels of varying stoichiometry and conclusions are drawn about the optimum stoichiometry and the range of acceptable values. Plans for future testing in support of specification development, confirmation of the results of accelerated testing by real-time experiments, and improvement in fuel performance and reliability are described

  16. Nuclear fuels

    Gangwani, Saloni; Chakrabortty, Sumita

    2011-01-01

    Nuclear fuel is a material that can be consumed to derive nuclear energy, by analogy to chemical fuel that is burned for energy. Nuclear fuels are the most dense sources of energy available. Nuclear fuel in a nuclear fuel cycle can refer to the fuel itself, or to physical objects (for example bundles composed of fuel rods) composed of the fuel material, mixed with structural, neutron moderating, or neutron reflecting materials. Long-lived radioactive waste from the back end of the fuel cycle is especially relevant when designing a complete waste management plan for SNF. When looking at long-term radioactive decay, the actinides in the SNF have a significant influence due to their characteristically long half-lives. Depending on what a nuclear reactor is fueled with, the actinide composition in the SNF will be different. The following paper will also include the uses. advancements, advantages, disadvantages, various processes and behavior of nuclear fuels

  17. Fuel and nuclear fuel cycle

    Prunier, C.

    1998-01-01

    The nuclear fuel is studied in detail, the best choice and why in relation with the type of reactor, the properties of the fuel cans, the choice of fuel materials. An important part is granted to the fuel assembly of PWR type reactor and the performances of nuclear fuels are tackled. The different subjects for research and development are discussed and this article ends with the particular situation of mixed oxide fuels ( materials, behavior, efficiency). (N.C.)

  18. Advanced PWR fuel design concepts

    Andersor, C.K.; Harris, R.P.; Crump, M.W.; Fuhrman, N.

    1987-01-01

    For nearly 15 years, Combustion Engineering has provided pressurized water reactor fuel with the features most suppliers are now introducing in their advanced fuel designs. Zircaloy grids, removable upper end fittings, large fission gas plenum, high burnup, integral burnable poisons and sophisticated analytical methods are all features of C-E standard fuel which have been well proven by reactor performance. C-E's next generation fuel for pressurized water reactors features 24-month operating cycles, optimal lattice burnable poisons, increased resistance to common industry fuel rod failure mechanisms, and hardware and methodology for operating margin improvements. Application of these various improvements offer continued improvement in fuel cycle economics, plant operation and maintenance. (author)

  19. Fracture of Zircaloy cladding by interactions with uranium dioxide pellets in LWR fuel rods. Technical report 10

    Smith, E.; Ranjan, G.V.; Cipolla, R.C.

    1976-11-01

    Power reactor fuel rod failures can be caused by uranium dioxide fuel pellet-Zircaloy cladding interactions. The report summarizes the current position attained in a detailed theoretical study of Zircaloy cladding fracture caused by the growth of stress corrosion cracks which form near fuel pellet cracks as a consequence of a power increase after a sufficiently high burn-up. It is shown that stress corrosion crack growth in irradiated Zircaloy must be able to proceed at very low stress intensifications if uniform friction effects are operative at the fuel-cladding interface, when the interfacial friction coefficient is less than unity, when a symmetric distribution of fuel cracks exists, and when symmetric interfacial slippage occurs (i.e., ''uniform'' conditions). Otherwise, the observed fuel rod failures must be due to departures from ''uniform'' conditions, and a very high interfacial friction coefficient and particularly fuel-cladding bonding, are means of providing sufficient stess intensification at a cladding crack tip to explain the occurrence of cladding fractures. The results of the investigation focus attention on the necessity for reliable experimental data on the stress corrosion crack growth behavior of irradiated Zircaloy, and for further investigations on the correlation between local fuel-cladding bonding and stress corrosion cracking

  20. Immune mediated liver failure

    Wang, Xiaojing; Ning, Qin

    2014-01-01

    Liver failure is a clinical syndrome of various etiologies, manifesting as jaundice, encephalopathy, coagulopathy and circulatory dysfunction, which result in subsequent multiorgan failure. Clinically, liver failure is classified into four categories: acute, subacute, acute-on-chronic and chronic liver failure. Massive hepatocyte death is considered to be the core event in the development of liver failure, which occurs when the extent of hepatocyte death is beyond the liver regenerative capac...

  1. Chronic heart failure

    Hopper, Ingrid; Easton, Kellie

    2017-01-01

    1. The common symptoms and signs of chronic heart failure are dyspnoea, ankle swelling, raised jugular venous pressure and basal crepitations. Other conditions may be confused with chronic heart failure, including dependent oedema or oedema due to renal or hepatic disease. Shortness of breath may be due to respiratory disease or severe anaemia. Heart failure secondary to lung disease (cor pulmonale) should be distinguished from congestive cardiac failure. Heart failure may also present with l...

  2. Improvements in quality of as-manufactured fuels for high-temperature gas-cooled reactors

    Minato, Kazuo; Kikuchi, Hironobu; Tobita, Tsutomu; Fukuda, Kousaku; Kaneko, Mitsunobu; Suzuki, Nobuyuki; Yoshimuta, Shigeharu; Tomimoto, Hiroshi.

    1997-01-01

    The mechanisms of coating failure of the fuel particles for the high-temperature gas-cooled reactors during coating and compaction processes of the fuel fabrication were studied to determine a way to reduce the defective particle fraction of the as-manufactured fuels. Through the observation of the defective particles, it was found that the coating failure during the coating process was mainly caused by the strong mechanical shocks to the particles given by violent particle fluidization in the coater and by unloading and loading of the particles. The coating failure during the compaction process was probably related to the direct contact with neighboring particles in the fuel compacts. The coating process was improved by optimizing the mode of the particle fluidization and by developing the process without unloading and loading of the particles at intermediate coating process. The compaction process was improved by optimizing the combination of the pressing temperature and the pressing speed of the overcoated particles. Through these modifications of the fabrication process, the quality of the as-manufactured fuel compacts was improved outstandingly. (author)

  3. The multi-class binomial failure rate model for the treatment of common-cause failures

    Hauptmanns, U.

    1995-01-01

    The impact of common cause failures (CCF) on PSA results for NPPs is in sharp contrast with the limited quality which can be achieved in their assessment. This is due to the dearth of observations and cannot be remedied in the short run. Therefore the methods employed for calculating failure rates should be devised such as to make the best use of the few available observations on CCF. The Multi-Class Binomial Failure Rate (MCBFR) Model achieves this by assigning observed failures to different classes according to their technical characteristics and applying the BFR formalism to each of these. The results are hence determined by a superposition of BFR type expressions for each class, each of them with its own coupling factor. The model thus obtained flexibly reproduces the dependence of CCF rates on failure multiplicity insinuated by the observed failure multiplicities. This is demonstrated by evaluating CCFs observed for combined impulse pilot valves in German NPPs. (orig.) [de

  4. Behavior of LWR fuel elements under accident conditions

    Albrecht, H.; Bocek, M.; Erbacher, F.; Fiege, A.; Fischer, M.; Hagen, S.; Hofmann, P.; Holleck, H.; Karb, E.; Leistikow, S.; Melang, S.; Ondracek, G.; Thuemmler, F.; Wiehr, K.

    1977-01-01

    In the frame of the German reactor safety research program, the Kernforschungszentrum Karlsruhe is carrying out a comprehensive program on the behavior of LWR fuel elements under a variety of power cooling mismatch conditions in particular during loss-of-coolant accidents. The major objectives are to establish a detailed quantitative understanding of fuel rod failures mechanisms and their thresholds, to evaluate the safety margins of power reactor cores under accident conditions and to investigate the feedback of fuel rod failures on the efficiency of emergency core cooling systems. This detailed quantitative understanding is achieved through extensive basic and integral experiments and is incorporated in a fuel behavior code. On the basis of these results the design of power reactor fuel elements and of safety devices can be further improved. The results of investigations on the inelastic deformation (ballooning) behavior of Zircaloy 4 cladding at LOCA temperatures in oxidizing atmosphere are presented. Depending upon strain rate and temperature superplastic deformation behavior was observed. In the equation of state of Zry 4 the strain rate sensitivity index depends strongly upon strain and in the superplastic region upon sample anisotropy. Oxidation kinetics experiments with Zry-tubes at 900-1300 0 C showed that the Baker-Just correlation describes the reality quite conservative. Therefore a reduction of the amount of Zry oxidation can be assumed in the course of a LOCA. The external oxidation of Zry-cladding by steam as well as internal oxidation by the oxygen in oxide fuel and fission products (Cs, I, Te) have an influence on the strain and rupture behavior of Zry-cladding at LOCA temperatures. In out-of-pile and inpile experiments the mechanical and thermal behavior of fuel rods during the blowdown, the heatup and the reflood phases of a LOCA are investigated under representative and controlled thermohydraulic conditions. The task of the inpile experiments is

  5. Fuel assembly

    Sakuyama, Tadashi; Mukai, Hideyuki.

    1988-01-01

    Purpose: To prevent the bending of a fuel rod caused by the difference in the elongation between a joined fuel rod and a standard fuel rod thereby maintain the fuel rod integrity. Constitution: A joined fuel rod is in a thread engagement at its lower end plug thereof with a lower plate, while passed through at its upper end plug into an upper tie plate and secured with a nut. Further, a standard fuel rod is engaged at its upper end plug and lower end plug with the upper tie plate and the lower tie plate respectively. Expansion springs are mounted to the upper end plugs of these bonded fuel rods and the standard fuel rods for preventing this lifting. Each of the fuel rods comprises a plurality of sintered pellets of nuclear fuel materials laminated in a zircaloy fuel can. The content of the alloy ingredient in the fuel can of the bonded fuel rod is made greater than that of the alloy ingredient of the standard fuel rod. this can increase the elongation for the bonded fuel rod, and the spring of the standard fuel rod is tightly bonded to prevent the bending. (Yoshino, Y.)

  6. Study of Maxwell–Wagner (M–W) relaxation behavior and hysteresis observed in bismuth titanate layered structure obtained by solution combustion synthesis using dextrose as fuel

    Subohi, Oroosa; Shastri, Lokesh; Kumar, G.S.; Malik, M.M.; Kurchania, Rajnish

    2014-01-01

    Graphical abstract: X-ray diffraction studies show that phase formation and crystallinity was reached only after calcinations at 800 °C. Dielectric constant versus temperature curve shows ferroelectric to paraelectric transition temperature (T c ) to be 650 °C. Complex impedance curves show deviation from Debye behavior. The material shows a thin PE Loop with low remnant polarization due to high conductivity in the as prepared sample. - Highlights: • Bi 4 Ti 3 O 12 is synthesized using solution combustion technique with dextrose as fuel. • Dextrose has high reducing capacity (+24) and generates more no. of moles of gases. • Impedance studies show that the sample follows Maxwell–Wagner relaxation behavior. • Shows lower remnant polarization due to higher c-axis ratio. - Abstract: Structural, dielectric and ferroelectric properties of bismuth titanate (Bi 4 Ti 3 O 12 ) obtained by solution combustion technique using dextrose as fuel is studied extensively in this paper. Dextrose is used as fuel as it has high reducing valancy and generates more number of moles of gases during the reaction. X-ray diffraction studies show that phase formation and crystallinity was reached only after calcinations at 800 °C. Dielectric constant versus temperature curve shows ferroelectric to paraelectric transition temperature (T c ) to be 650 °C. The dielectric loss is very less (tan δ < 1) at lower temperatures but increases around T c due to structural changes in the sample. Complex impedance curves show deviation from Debye behavior. The material shows a thin PE Loop with low remnant polarization due to high conductivity in the as prepared sample

  7. Application of Cherenkov light observation to reactor measurements (3). Evaluation of spent fuel elements of LWRs with Cherenkov light estimation system

    Yamamoto, Keiichi; Takeuchi, Tomoaki; Tsuchiya, Kunihiko; Hayashi, Takayasu; Kosuge, Fumiaki

    2016-11-01

    Development of the reactor measurement system has been carried out to obtain the real-time in-core nuclear and thermal information, where the quantitative measurement of brightness of Cherenkov light was investigated. The system would be applied as a monitoring system in severe accidents and for the advanced operation management technology in existing LWRs. This report summarized the modification of Cherenkov light estimation system described JAEA-Testing 2015-001 and the result of the burn-up evaluation by Cherenkov light image emitted from spent fuel elements of LWRs with the modified system. (author)

  8. Analysis of WWER-440 fuel performance under normal operating conditions

    Gunduz, Oe; Koese, S; Akbas, T [Atomenerjisi Komisyonu, Ankara (Turkey); Colak, Ue [Ankara Nuclear Research and Training Center (Turkey)

    1994-12-31

    FRAPCON-2 code originally developed for LWR fuel behaviour simulation is used to analyse the WWER-440 fuel rod behaviour at normal operational conditions. The code is capable of utilizing different models for mechanical analysis and gas release calculations. Heat transfer calculations are accomplished through a collocation technique by the method of weighted residuals. Temperature and burnup element properties are evaluated using MATPRO package. As the material properties of Zr-1%Nb used as cladding in WWER-440s are not provided in the code, Zircaloy-4 is used as a substitute for Zr-1%Nb. Mac-Donald-Weisman model is used for gas release calculation. FRACAS-1 and FRACAS-2 models are used in the mechanical calculations. It is assumed that the reactor was operated for 920 days (three consecutive cycles), the burnup being 42000 Mwd/t U. Results of the fuel rod behaviour analysis are given for three axial nodes: bottom node, central node and top node. The variations of the following characteristic fuel rod parameters are studied through the prescribed power history: unmoved gap thickness, gap heat transfer coefficient, fuel axial elongation, cladding axial elongation, fuel centerline temperature and ZrO-thickness at cladding surface. The value of each parameter is calculated as a function of the effective power days for the three nodes by using FRACAS-1 and FRACAS-2 codes for comparison.The results show that calculations with deformable pellet approximation with FRACAS-II model could provide better information for the behaviour of a typical fuel rod. Calculations indicate that fuel rod failure is not observed during the operation. All fuel rod parameters investigated are found to be within the safety limits. It is concluded, however, that for better assessment of reactor safety these calculations should be extended for transient conditions such as LOCA. 1 tab., 10 figs., 4 refs.

  9. Fuel processing

    Allardice, R.H.

    1990-01-01

    The technical and economic viability of the fast breeder reactor as an electricity generating system depends not only upon the reactor performance but also on a capability to recycle plutonium efficiently, reliably and economically through the reactor and fuel cycle facilities. Thus the fuel cycle is an integral and essential part of the system. Fuel cycle research and development has focused on demonstrating that the challenging technical requirements of processing plutonium fuel could be met and that the sometimes conflicting requirements of the fuel developer, fuel fabricator and fuel reprocessor could be reconciled. Pilot plant operation and development and design studies have established both the technical and economic feasibility of the fuel cycle but scope for further improvement exists through process intensification and flowsheet optimization. These objectives and the increasing processing demands made by the continuing improvement to fuel design and irradiation performance provide an incentive for continuing fuel cycle development work. (author)

  10. B-type natriuretic peptide measurement in primary care; magnitude of associations with cardiovascular risk factors and their therapies. Observations from the STOP-HF (St. Vincent's Screening TO Prevent Heart Failure) study.

    Conlon, Carmel M

    2012-02-01

    BACKGROUND: An effective prevention strategy for heart failure in primary care requires a reliable screening tool for asymptomatic ventricular dysfunction. Preliminary data indicate that B-type natriuretic peptide (BNP) may be suitable for this task. However, for the most effective use of this peptide, the interrelationships between associated risk factors and their therapies on BNP, and in particular their magnitude of effect, needs to be established in a large primary care population. Therefore, the objective of the study was to establish the extent of the association between BNP, cardiovascular risk factors and their therapies. METHODS: BNP measurement and clinical review was preformed on 1122 primary care patients with cardiovascular risk factors. Multivariate analyses identified significant associates of BNP concentrations which were further explored to establish the magnitude of their association. RESULTS: Associates of BNP were age (1.36-fold increase in BNP\\/decade), female (1.28), beta-blockers (1.90), myocardial infarction (1.36), arrhythmia (1.98), diastolic blood pressure; all p<0.01. A novel method was devised that plotted median BNP per sliding decade of age for the various combinations of these principal associates. CONCLUSIONS: The data presented underline the importance of considering several clinical and therapeutic factors when interpreting BNP concentrations. Most of these variables were associated with increased concentrations, which may in part explain the observed false-positive rates for detecting ventricular dysfunction using this peptide. Furthermore, the design of studies or protocols using BNP as an endpoint or a clinical tool should take particular account of these associations. This analysis provides the foundation for age, risk factor and therapy adjusted reference ranges for BNP in this setting.

  11. Heart failure - surgeries and devices

    ... surgery; HF - surgery; Intra-aortic balloon pumps - heart failure; IABP - heart failure; Catheter based assist devices - heart failure ... problem may cause heart failure or make heart failure worse. Heart valve surgery may be needed to repair or ...

  12. Teton Dam failure

    Snorteland, N. [United States Dept. of the Interior, Washington, DC (United States). Bureau of Reclamation

    2009-07-01

    This case summary discussed an internal erosion failure that occurred at the embankment foundation of Teton Dam. The project was designed as a run-of-the-river power generation facility and to provide irrigation, flood protection, and power generation to the lower Teton region of southern Idaho. The dam site was located next to the eastern Snake River plain, a volcanic filled depression. The foundation's cutoff trench was excavated into the bedrock along the length of the dam. The dam was designed as a zoned earthfill with a height of 305 feet. A trench made of low plasticity windblown silt was designed to connect the embankment core to the rock foundation. Seeps were noted in 1976, and a leak was observed near the toe of the dam. A wet spot appeared on the downstream face of the dam at elevation 5200. A sinkhole then developed. The embankment crest collapsed, and the dam breached. Peak outflow was estimated at 1,000,000 cfs. The failure was attributed to a lack of communication between designers, a failure to understand geologic information about the region, and an insufficient review of designs and specifications by designers and field personnel. No monitoring instrumentation was installed in the embankment. Approximately 300 square miles were inundated, and 25,000 people were displaced. Eleven people were killed. A review group noted that the rock surface was not adequately sealed, and that the dam failed as a result of inadequate protection of the impervious core material from internal erosion. 42 figs.

  13. Failed fuel detection device

    Sudo, Takayuki.

    1983-01-01

    Purpose: To enable early and sure detection of failed fuels by automatically changing the alarm set value depending on the operation states of a nuclear reactor. Constitution: Gaseous fission products released into coolants are transferred further into cover gases and then introduced through a pipeway to a failed fuel detector. The cover gases introduced from the pipeway to the pipeway or chamber within the detection device are detected by a radiation detector for the radiation dose of the gaseous fission products contained therein. The detected value is converted and amplified as a signal and inputted to a comparator. While on the other hand, a signal corresponding to the reactor power is converted by an alarm setter into a set value and inputted to the comparator. In such a structure, early and sure detection can be made for the fuel failures. (Yoshino, Y.)

  14. Studies in Phebus reactor of fuel behaviour upon LOCA conditions

    Manin, A.; Del Negro, R.; Reocreux, M.

    1980-09-01

    The fuel behaviour upon LOCA conditions is studied in an in-pile loop, in Phebus reactor. This paper presents: a short description of Phebus reactor; the current program (adjusting the thermohydraulic conditions in order to get cladding failure); the program developments (consequences involved by cladding failure); the fuel test conditions determination [fr

  15. Renal failure in patients with multiple myeloma.

    Almueilo, Samir H

    2015-01-01

    Renal dysfunction is encountered in 20-25% of patients with multiple myeloma (MM) at the time of diagnosis. There is often a precipitating event. Several biochemical and clinical correlations with renal failure in MM have been reported. Renal failure in MM is associated with worse outcome of the disease. We retrospectively analyzed the medical records of 64 patients with MM admitted to our institution during the period January 1992 to December 2012. Abnormal renal function was observed in 24 (37.5%) patients and 17 (26.6%) of them had renal failure; 14 of the 17 (82.4%) of patients with renal failure had Stage III MM. Urine Bence- Jones protein was positive in ten (58.8%) patients with renal failure versus ten (21.3%) patients without renal failure (P = 0.004). Potential precipitating factors of renal failure were determined in nine patients. Renal function normalized in 11 patients with simple measures, while six patients required hemodialysis; one remained dialysis dependent till time of death. Early mortality occurred in five (29.4%) patients with renal failure as compared with two (4.3%) patients in the group without renal failure (P = 0.005). In conclusion, renal failure is associated with a higher tumor burden and Bence-Jones proteinuria in patients with MM. It is reversible in the majority of patients; however, early mortality tends to be higher in patients with persistent renal failure.

  16. Corrosion induced failure analysis of subsea pipelines

    Yang, Yongsheng; Khan, Faisal; Thodi, Premkumar; Abbassi, Rouzbeh

    2017-01-01

    Pipeline corrosion is one of the main causes of subsea pipeline failure. It is necessary to monitor and analyze pipeline condition to effectively predict likely failure. This paper presents an approach to analyze the observed abnormal events to assess the condition of subsea pipelines. First, it focuses on establishing a systematic corrosion failure model by Bow-Tie (BT) analysis, and subsequently the BT model is mapped into a Bayesian Network (BN) model. The BN model facilitates the modelling of interdependency of identified corrosion causes, as well as the updating of failure probabilities depending on the arrival of new information. Furthermore, an Object-Oriented Bayesian Network (OOBN) has been developed to better structure the network and to provide an efficient updating algorithm. Based on this OOBN model, probability updating and probability adaptation are performed at regular intervals to estimate the failure probabilities due to corrosion and potential consequences. This results in an interval-based condition assessment of subsea pipeline subjected to corrosion. The estimated failure probabilities would help prioritize action to prevent and control failures. Practical application of the developed model is demonstrated using a case study. - Highlights: • A Bow-Tie (BT) based corrosion failure model linking causation with the potential losses. • A novel Object-Oriented Bayesian Network (OOBN) based corrosion failure risk model. • Probability of failure updating and adaptation with respect to time using OOBN model. • Application of the proposed model to develop and test strategies to minimize failure risk.

  17. Failure in imperfect anisotropic materials

    Legarth, Brian Nyvang

    2005-01-01

    The fundamental cause of crack growth, namely nucleation and growth of voids, is investigated numerically for a two phase imperfect anisotropic material. A unit cell approach is adopted from which the overall stress strain is evaluated. Failure is observed as a sudden stress drop and depending...

  18. Stress Analysis of Single Spacer Grid Support considering Fuel Rod

    Yoo, Y. G.; Jung, D. H.; Kim, J. H. [Chungnam National University, Daejeon (Korea, Republic of); Park, J. K.; Jeon, K. L. [Korea Nuclear Fuel, Daejeon (Korea, Republic of)

    2010-10-15

    Pressurized water reactor (PWR) nuclear fuel assembly is mainly composed of a top-end piece, a bottom-end piece, lots of fuel rods, and several spacer grids. Among them, the main function of spacer grid is protecting fuel rods from Fluid Induced Vibration (FIV). The cross section of spacer grid assembled by laser welding in upper and lower point. When the fuel rod inserted in spacer gird, spring and dimple and around of welded area got a stresses. The main hypothesis of this analysis is the boundary area of HAZ and base metal can get a lot of damage than other area by FIV. So, design factors of spacer grid mainly considered to preventing the fatigue failure in HAZ and spring and dimple of spacer grid. From previous researching, the environment in reactor verified. Pressure and temperature of light water observed 15MPa and 320 .deg. C, and vibration of the fuel rod observed within 0 {approx} 50Hz. In this study, mechanical properties of zirconium alloy that extracted from the test and the spacer grid model which used in the PWR were applied in stress analyzing. General-purpose finite element analysis program was used ANSYS Workbench 12.0.1 version. 3-D CAD program CATIA was used to create spacer grid model

  19. Sulphur in liquid fuels 2002

    Guthrie, J. [Environment Canada, Gatineau, PQ (Canada). Fuels Div., Oil, Gas and Energy Branch ; Sabourin, R. [Carleton Univ., Ottawa, ON (Canada)

    2003-08-01

    Environment Canada has developed new regulations for sulphur content in fuels in an effort to align with requirements recently passed by the U.S. Environmental Protection Agency. This report summarizes data regarding sulphur content in liquid fuels for the year 2002. The requirements of the Sulphur in Gasoline Regulation came into effect in 2002, limiting the average sulphur content of gasoline to 150 mg/kg. In January 2005, a 30 mg/kg average limit will come into effect. Also, in July 2002, the Sulphur in Diesel Fuel Regulation stipulated a maximum limit of 500 mg/kg for on-road diesel fuel. The new regulation continues this limit until mid-2006 at which time a 15 mg/kg limit will come into effect for on-road diesel fuel. Nationally, the average sulphur content in gasoline in 2002 was 246 mg/kg, which was 14.3 per cent lower than in 2001. The data covers the period from January 1 to December 31, 2002 and was obtained from petroleum refineries and importing companies that are required to submit quarterly information to the regional office of Environment Canada. Failure to comply results in penalties. The report includes data for aviation turbo fuel, motor gasoline, aviation gasoline, kerosene oil, low-sulphur diesel fuel, diesel fuel, light fuel oil, and heavy fuel oil. 16 tabs., 17 figs., 7 appendices.

  20. A pellet-clad interaction failure criterion

    Howl, D.A.; Coucill, D.N.; Marechal, A.J.C.

    1983-01-01

    A Pellet-Clad Interaction (PCI) failure criterion, enabling the number of fuel rod failures in a reactor core to be determined for a variety of normal and fault conditions, is required for safety analysis. The criterion currently being used for the safety analysis of the Pressurized Water Reactor planned for Sizewell in the UK is defined and justified in this paper. The criterion is based upon a threshold clad stress which diminishes with increasing fast neutron dose. This concept is consistent with the mechanism of clad failure being stress corrosion cracking (SCC); providing excess corrodant is always present, the dominant parameter determining the propagation of SCC defects is stress. In applying the criterion, the SLEUTH-SEER 77 fuel performance computer code is used to calculate the peak clad stress, allowing for concentrations due to pellet hourglassing and the effect of radial cracks in the fuel. The method has been validated by analysis of PCI failures in various in-reactor experiments, particularly in the well-characterised power ramp tests in the Steam Generating Heavy Water Reactor (SGHWR) at Winfrith. It is also in accord with out-of-reactor tests with iodine and irradiated Zircaloy clad, such as those carried out at Kjeller in Norway. (author)

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

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

    2013-01-01

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

  2. The outcome of the working visits by the experts to the fuel provider enterprises

    Mironov, Y.; Molchanov, V.

    2015-01-01

    The working visits by the international experts to the fabrication plants of nuclear fuels and the zirconium alloy component facility took part in the framework of the „Zero failure level“ project. The purpose of these working visits was to determine whether there are any cause-effect relationships between the production of nuclear fuel and the fuel failures as well as to identify the trends and reasons for such failures. As an outcome of the analysis of the production processes at the fabrication plants of the nuclear fuel and the zirconium alloys component facility, the system root causes affecting the failures of the nuclear fuel were not identified

  3. Stainless steel clad for light water reactor fuels. Final report

    Rivera, J.E.; Meyer, J.E.

    1980-07-01

    Proper reactor operation and design guidelines are necessary to assure fuel integrity. The occurrence of fuel rod failures for operation in compliance with existing guidelines suggests the need for more adequate or applicable operation/design criteria. The intent of this study is to develop such criteria for light water reactor fuel rods with stainless steel clad and to indicate the nature of uncertainties in its development. The performance areas investigated herein are: long term creepdown and fuel swelling effects on clad dimensional changes and on proximity to clad failure; and short term clad failure possibilities during up-power ramps

  4. Transport device of spent fuel

    Watanabe, Takashi.

    1976-01-01

    Object: To provide a transport device of spent fuel particularly used in a fast breeder, which can enhance accessibility to travelling mechanism portions and exchangeability thereof to facilitate maintenance in the event of failure. Structure: On a travelling floor, which has a function to shield radioactive rays, extending in a direction of transporting spent fuel and being formed with a break passing through in a direction wall thickness, a travelling body is moved along the break. The travelling body has a support rod member mounted thereon, and the support rod member is moved within the break, the support rod member having a fuel support pocket suspended therefrom. (Furukawa, Y.)

  5. Fuel pin transient behavior technology applied to safety analyses. Presentation to AEC Regulatory Staff 4th Regulatory Briefing on safety technology, Washington, D.C., November 19--20, 1974

    1974-11-01

    Information is presented concerning LMFBR fuel pin performance requirements and evaluation; fuels behavior codes with safety interfaces; performance evaluations; ex-reactor materials and simulation tests; models for fuel pin failure; and summary of continuing fuels technology tasks. (DCC)

  6. Fuel assembly

    Fujibayashi, Toru.

    1970-01-01

    Herein disclosed is a fuel assembly in which a fuel rod bundle is easily detachable by rotating a fuel rod fastener rotatably mounted to the upper surface of an upper tie-plate supporting a fuel bundle therebelow. A locking portion at the leading end of each fuel rod protrudes through the upper tie-plate and is engaged with or separated from the tie-plate by the rotation of the fastener. The removal of a desired fuel rod can therefore be remotely accomplished without the necessity of handling pawls, locking washers and nuts. (Owens, K.J.)

  7. Nuclear fuel

    D Hondt, P.

    1998-01-01

    The research and development programme on nuclear fuel at the Belgian Nuclear Research Centre SCK/CEN is described. The objective of this programme is to enhance the quantitative prediction of the operational limits of nuclear fuel and to assess the behaviour of fuel under incidental and accidental conditions. Progress is described in different domains including the modelling of fission gas release in LWR fuel, thermal conductivity, basic physical phenomena, post-irradiation examination for fuel performance assessment, and conceptual studies of incidental and accidental fuel experiments

  8. Lubricity characteristics of marine distillate fuels

    Crutchley, Ian [Innospec Fuel Specialties, Ellesmere Port (United Kingdom); Green, Michael [Intertek Lintec ShipCare Services, Darlington (United Kingdom)

    2012-08-15

    This article from Innospec Fuel Specialties, Ellesmere Port, UK, and Intertek Lintec ShipCare Services, Darlington, UK, examines the lubricity characteristics of marine distillate fuels available today in relation to the requirements and limits imposed in ISO8217:2010. It will estimate expected failure rates and also asses the perceived relationship between lubricity, sulphur content and viscosity. (orig.)

  9. Proposed model for fuel-coolant mixing during a core-melt accident

    Corradini, M.L.

    1983-01-01

    If complete failure of normal and emergency coolant flow occurs in a light water reactor, fission product decay heat would eventually cause melting of the reactor fuel and cladding. The core melt may then slump into the lower plenum and later into the reactor cavity and contact residual liquid water. A model is proposed to describe the fuel-coolant mixing process upon contact. The model is compared to intermediate scale experiments being conducted at Sandia. The modelling of this mixing process will aid in understanding three important processes: (1) fuel debris sizes upon quenching in water, (2) the hydrogen source term during fuel quench, and (3) the rate of steam production. Additional observations of Sandia data indicate that the steam explosion is affected by this mixing process

  10. Acute liver failure

    Larsen, Fin Stolze; Bjerring, Peter Nissen

    2011-01-01

    Acute liver failure (ALF) results in a multitude of serious complications that often lead to multi-organ failure. This brief review focuses on the pathophysiological processes in ALF and how to manage these.......Acute liver failure (ALF) results in a multitude of serious complications that often lead to multi-organ failure. This brief review focuses on the pathophysiological processes in ALF and how to manage these....

  11. Dodewaard fuel supply agreement - a model for the future

    Raven, L.F.; Hubers, C.

    1980-01-01

    An Agreement between the Utility GKN and the Fuel Supplier BNFL has eliminated any Utility imposed penalty clauses for fuel failure due to operational conditions and, consequently, there are no restrictions imposed by the Fuel Supplier on the reactor operational manoeuvres. The result is that the Utility can now decide if the risk of fuel clad failure during a reactor power ramp outweighs the financial loss due to slower ramp rates. It is the Utility and not the Fuel Supplier who is in the best position to make this judgment provided adequate operational experience and computer codes are available to quantify the risk. The paper discusses the reactor operational experience, including the fuel failure rate and the confirmation of PCI failure by post irradiation examination. It establishes the practicality of the Agreement for the Dodewaard reactor and suggests such arrangements could be beneficial to other Utilities. (author)

  12. Zero tolerance for failure. An AREVA initiative to improve reliability

    Lippert, Hans-Joachim; Gentet, Guy; Mollard, Pierre; Garner, Norman

    2010-01-01

    Significant improvements in fuel reliability have been realized over the past 2 decades, but total elimination of failures has remained elusive. Driving reliability to higher levels requires a philosophy that does not accept that even infrequent and isolated failures are inevitable - it was on this foundation that Areva's Zero Tolerance for Failure (ZTF) initiative was established. This is not in itself either a program or project, but a fundamental shift in the way of thinking about work according to the following four principles: - Failures are avoidable, - Zero failures are our goal, - We will respond rapidly to any failure, - We succeed when we fix failures in a way that precludes recurrence. The shift to a ZTF philosophy is a broad change in corporate culture that expands the concept of failure far beyond cases where fuel rod cladding integrity is breached. While this paper specifically illustrates the ways in which ZTF has shaped the company's response to enhancing fuel rod reliability, ZTF extends to any failures of fuel products to deliver expected levels of performance, manufacturing processes to meet specifications and high first-pass acceptance criteria, and beyond to error-free performance of engineering analyses and cycle design and licensing services. Application of ZTF to enhancing fuel reliability deploys efforts in the areas of manufacturing, human factors, design, R and D, processes and product strategy. In order to achieve the necessary improvements, a number of important actions have been initiated across regions and facilities. In addition to these global scale projects and measures, each region contributes by adopting measures which are relevant to its particular activities and market needs. (orig.)

  13. Failures and Defects in the Building Process

    Jørgensen, Kirsten

    2009-01-01

    Function failures, defects, mistakes and poor communication are major problems for the construction sector. As the empirical element in the research, a large construction site was observed from the very start to the very end and all failures and defects of a certain size were recorded and analysed...

  14. In Support of Failure

    Carr, Allison

    2013-01-01

    In this essay, I propose a concerted effort to begin devising a theory and pedagogy of failure. I review the discourse of failure in Western culture as well as in composition pedagogy, ultimately suggesting that failure is not simply a judgement or indication of rank but is a relational, affect-bearing concept with tremendous relevance to…

  15. Database for the OECD-IAEA Paks Fuel Project

    Szabo, Emese; Hozer, Zoltan; Gyori, Csaba; Hegyi, Gyoergy

    2010-01-01

    On 10 April 2003 severe damage of fuel assemblies took place during an incident at Unit 2 of Paks Nuclear Power Plant in Hungary. The assemblies were being cleaned in a special tank below the water level of the spent fuel storage pool in order to remove crud buildup. That afternoon, the chemical cleaning of assemblies was completed and the fuel rods were being cooled by circulation of storage pool water. The first sign of fuel failure was the detection of some fission gases released from the cleaning tank during that evening. The cleaning tank cover locks were released after midnight and this operation was followed by a sudden increase in activity concentrations. The visual inspection revealed that all 30 fuel assemblies were severely damaged. The first evaluation of the event showed that the severe fuel damage happened due to inadequate coolant circulation within the cleaning tank. The damaged fuel assemblies will be removed from the cleaning tank in 2005 and will be stored in special canisters in the spent fuel storage pool of the Paks NPP. Following several discussions between expert from different countries and international organisations the OECD-IAEA Paks Fuel Project was proposed. The project is envisaged in two phases. - Phase 1 is to cover organization of visual inspection of material, preparation of database, performance of analyses and preparatory work for fuel examination. - Phase 2 is to cover the fuel transport and the hot cell examination. The first meeting of the project was held in Budapest on 30-31 January 2006. Phase 1 of the Paks Fuel Project will focus on the numerical simulation of the most important aspects of the incident. This activity will help in the reconstruction of the accidental scenario. The first step of Phase 1 was the collection of a database necessary for the code calculations. The main objective of database collection was to provide input data for calculations. For this reason the collection was focused on such data that are

  16. Component failure data handbook

    Gentillon, C.D.

    1991-04-01

    This report presents generic component failure rates that are used in reliability and risk studies of commercial nuclear power plants. The rates are computed using plant-specific data from published probabilistic risk assessments supplemented by selected other sources. Each data source is described. For rates with four or more separate estimates among the sources, plots show the data that are combined. The method for combining data from different sources is presented. The resulting aggregated rates are listed with upper bounds that reflect the variability observed in each rate across the nuclear power plant industry. Thus, the rates are generic. Both per hour and per demand rates are included. They may be used for screening in risk assessments or for forming distributions to be updated with plant-specific data

  17. Iridium: failures & successes

    Christensen, CarissaBryce; Beard, Suzette

    2001-03-01

    This paper will provide an overview of the Iridium business venture in terms of the challenges faced, the successes achieved, and the causes of the ultimate failure of the venture — bankruptcy and system de-orbit. The paper will address technical, business, and policy issues. The intent of the paper is to provide a balanced and accurate overview of the Iridium experience, to aid future decision-making by policy makers, the business community, and technical experts. Key topics will include the history of the program, the objectives and decision-making of Motorola, the market research and analysis conducted, partnering strategies and their impact, consumer equipment availability, and technical issues — target performance, performance achieved, technical accomplishments, and expected and unexpected technical challenges. The paper will use as sources trade media and business articles on the Iridium program, technical papers and conference presentations, Wall Street analyst's reports, and, where possible, interviews with participants and close observers.

  18. CONSEQUENCES OF FAILURE OF GAS NETWORK INFRASTRUCTURE

    Marek URBANIK

    2016-06-01

    Full Text Available Ecology today is becoming increasingly important. Increasing air pollution and greenhouse gas emissions make the search for such fuels which will not have such a negative effect on the environment as the fuel use currently - mainly coal. At present it seems that the substitute fuel can be gaseous fuels (propane-butane, methane. Their combustion is less harmful to the environment and their transport is relatively not very complicated. As it turns out, the use of gas is increasing in industry, automotive, heating systems (power plants that operate in the so-called cogeneration. The increase in demand carries continuous development of gas infrastructure, which in turn may increase probability of failure. As a conclusion of this article, taking into account all the construction disasters induced by the gas, the number of such failures is relatively small. It should be remembered, that the disaster caused by gas explosion may cause very large material and human losses. Not without significance is the impact of gas leakage, eg. of the pipeline on the environment. An example is the methane which is a greenhouse gas, less persistent in the air, but much more active than CO2. The article presents selected disasters associated with natural gas or propane-butane and the impact of these gases on the environment because these fuels are most commonly used in most sectors of the economy.

  19. Field failure mechanisms for photovoltaic modules

    Dumas, L. N.; Shumka, A.

    1981-01-01

    Beginning in 1976, Department of Energy field centers have installed and monitored a number of field tests and application experiments using current state-of-the-art photovoltaic modules. On-site observations of module physical and electrical degradation, together with in-depth laboratory analysis of failed modules, permits an overall assessment of the nature and causes of early field failures. Data on failure rates are presented, and key failure mechanisms are analyzed with respect to origin, effect, and prospects for correction. It is concluded that all failure modes identified to date are avoidable or controllable through sound design and production practices.

  20. Failure Waves in Cylindrical Glass Bars

    Cazamias, James U.; Bless, Stephan J.; Marder, Michael P.

    1997-07-01

    Failure waves, a propagating front separating virgin and comminuted material, have been receiving a fair amount of attention the last couple of years. While most scientists have been looking at failure waves in plate impact geometries, we have conducted a series of experiments on Pyrex bars. In this paper, we present two types of photographic data from a series of tests. A streak camera was used to determine velocities of the failure front as a function of impact stress. A polaroid camera and a flash lamp provide detailed pictures of the actual event. Attempts were made to observe failure waves in amorphous quartz and acrylic.

  1. End-of-life destructive examination of light water breeder reactor fuel rods (LWBR Development Program)

    Richardson, K.D.

    1987-10-01

    Destructive examination of 12 representative Light Water Breeder Reactor fuel rods was performed following successful operation in the Shippingport Atomic Power Station for 29,047 effective full power hours, about five years. Light Water Breeder Reactor fuel rods were unique in that the thorium oxide and uranium-233 oxide fuel was contained within Zircaloy-4 cladding. Destructive examinations included analysis of released fission gas; chemical analysis of the fuel to determine depletion, iodine, and cesium levels; chemical analysis of the cladding to determine hydrogen, iodine, and cesium levels; metallographic examination of the cladding, fuel, and other rod components to determine microstructural features and cladding corrosion features; and tensile testing of the irradiated cladding to determine mechanical strength. The examinations confirmed that Light Water Breeder Reactor fuel rod performance was excellent. No evidence of fuel rod failure was observed, and the fuel operating temperature was low (below 2580 0 F at which an increased percentage of fission gas is released). 21 refs., 80 figs., 20 tabs

  2. Preliminary test results for post irradiation examination on the HTTR fuel

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

    2007-01-01

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

  3. Irradiated-Microsphere Gamma Analyzer (IMGA): an integrated system for HTGR coated particle fuel performance assessment

    Kania, M.J.; Valentine, K.H.

    1980-02-01

    The Irradiated-Microsphere Gamma Analyzer (IMGA) System, designed and built at ORNL, provides the capability of making statistically accurate failure fraction measurements on irradiated HTGR coated particle fuel. The IMGA records the gamma-ray energy spectra from fuel particles and performs quantitative analyses on these spectra; then, using chemical and physical properties of the gamma emitters it makes a failed-nonfailed decision concerning the ability of the coatings to retain fission products. Actual retention characteristics for the coatings are determined by measuring activity ratios for certain gamma emitters such as 137 Cs/ 95 Zr and 144 Ce/ 95 Zr for metallic fission product retention and 134 Cs/ 137 Cs for an indirect measure of gaseous fission product retention. Data from IMGA (which can be put in the form of n failures observed in N examinations) can be accurately described by the binomial probability distribution model. Using this model, a mathematical relationship between IMGA data (n,N), failure fraction, and confidence level was developed. To determine failure fractions of less than or equal to 1% at confidence levels near 95%, this model dictates that from several hundred to several thousand particles must be examined. The automated particle handler of the IMGA system provides this capability. As a demonstration of failure fraction determination, fuel rod C-3-1 from the OF-2 irradiation capsule was analyzed and failure fraction statistics were applied. Results showed that at the 1% failure fraction level, with a 95% confidence level, the fissile particle batch could not meet requirements; however, the fertile particle exceeded these requirements for the given irradiation temperature and burnup

  4. Determination and microscopic study of incipient defects in irradiated power reactor fuel rods. Final report

    Pasupathi, V.; Perrin, J.S.; Roberts, E.

    1978-05-01

    This report presents the results of nondestructive and destructive examinations carried out on the Point Beach-1 (PWR) and Dresden-3 (BWR) candidate fuel rods selected for the study of pellet-clad interaction (PCI) induced incipient defects. In addition, the report includes results of examination of sections from Oskarshamn-1 (BWR) fuel rods. Eddy current examination of Point Beach-1 rods showed indications of possible incipient defects in the fuel rods. The profilometry and the gamma scan data also indicated that the source of the eddy current indications may be incipient defects. No failed rods or rods with incipient failure were found in the sample from Point Beach-1. Despite the lack of success in finding incipient defects and filed rods, the mechanism for fuel rod failures in Point Beach-1 is postulated to be PCI-related, with high startup rates and fuel handling being the key elements. Nine out of the 10 candidate fuel rods from Dresden-3 (BWR) were failed, and all the failed rods had leaked water so that the initial mechanism was observed. Examination of clad inner surfaces of the specimens from failed and unfailed rods showed fuel deposits of widely varying appearance. The deposits were found to contain uranium, cesium, and tellurium. Transmission electron microscopy of clad specimens showed evidence of microscopic strain. Metallographic examination of fuel pellets from the peak transient power location showed extensive grain boundary separation and axial movement of the fuel indicative of rapid release of fission products. Examination of Oskarshamn clad specimens did not show any stress corrosion crack (SCC) type defects. The defects found in the examinations appear to be related to secondary hydriding. The clad inner surface of the Oskarshamn specimens also showed uranium-rich deposits of varying features

  5. Fuel management

    Schwarz, E.R.

    1975-01-01

    Description of the operation of power plants and the respective procurement of fuel to fulfil the needs of the grid. The operation of the plants shall be optimised with respect to the fuel cost. (orig./RW) [de

  6. Fuel gases

    Anon.

    1996-01-01

    This paper gives a brief presentation of the context, perspectives of production, specificities, and the conditions required for the development of NGV (Natural Gas for Vehicle) and LPG-f (Liquefied Petroleum Gas fuel) alternative fuels. After an historical presentation of 80 years of LPG evolution in vehicle fuels, a first part describes the economical and environmental advantages of gaseous alternative fuels (cleaner combustion, longer engines life, reduced noise pollution, greater natural gas reserves, lower political-economical petroleum dependence..). The second part gives a comparative cost and environmental evaluation between the available alternative fuels: bio-fuels, electric power and fuel gases, taking into account the processes and constraints involved in the production of these fuels. (J.S.)

  7. Fuel cycles

    Hawley, N.J.

    1983-05-01

    AECL publications, from the open literature, on fuels and fuel cycles used in CANDU reactors are listed in this bibliography. The accompanying index is by subject. The bibliography will be brought up to date periodically

  8. VVER fuel. Results of post irradiation examination

    Smirnov, V.P.; Markov, D.V.; Smirnov, A.V.; Polenok, V.S.; Perepelkin, S.O.; Ivashchenko, A.A.

    2005-01-01

    The present paper presents the main results of post-irradiation examination of more than 40 different fuel assemblies (FA) operated in the cores of VVER-1000 and VVER-440-type power reactors in a wide range of fuel burnup. The condition of fuel assembly components from the viewpoint of deformation, corrosion resistance and mechanical properties is described here. A serviceability of the FA design as a whole and interaction between individual FA components under vibration condition and mechanical load received primary emphasis. The reasons of FA damage fuel element failure in a wide range of fuel burnup are also analyzed. A possibility and ways of fuel burnup increase have been proved experimentally for the case of high-level serviceability maintenance of fuel elements to provide for advanced fuel cycles. (author)

  9. The chemistry of water reactor fuel

    Potter, P.E.

    1990-01-01

    In this paper, the authors discuss features of the changes in chemical constitution which occur in fuel and fuel rods for water reactors during operation and in fault conditions. The fuel for water reactors consists of pellets of urania (UO 2 ) clad in Zircaloy. An essential step in the prediction of the fate of all the radionuclides in a fault or accident is to possess a detailed knowledge of their chemical behavior at all stages of the development of such incidents. In this paper, the authors consider: the chemical constitution of fuel during operation at temperatures corresponding to rather low ratings, together with a quite detailed discussion of the chemistry within the fuel-clad gap; the behavior of fuel subjected to higher temperatures and ratings than those of contemporary fuel; and the changes in constitution on failure of fuel rods in fault or accident conditions

  10. Nuclear fuels

    2008-01-01

    The nuclear fuel is one of the key component of a nuclear reactor. Inside it, the fission reactions of heavy atoms, uranium and plutonium, take place. It is located in the core of the reactor, but also in the core of the whole nuclear system. Its design and properties influence the behaviour, the efficiency and the safety of the reactor. Even if it represents a weak share of the generated electricity cost, its proper use represents an important economic stake. Important improvements remain to be made to increase its residence time inside the reactor, to supply more energy, and to improve its robustness. Beyond the economical and safety considerations, strategical questions have to find an answer, like the use of plutonium, the management of resources and the management of nuclear wastes and real technological challenges have to be taken up. This monograph summarizes the existing knowledge about the nuclear fuel, its behaviour inside the reactor, its limits of use, and its R and D tracks. It illustrates also the researches in progress and presents some key results obtained recently. Content: 1 - Introduction; 2 - The fuel of water-cooled reactors: aspect, fabrication, behaviour of UO 2 and MOX fuels inside the reactor, behaviour in loss of tightness situation, microscopic morphology of fuel ceramics and evolution under irradiation - migration and localisation of fission products in UOX and MOX matrices, modeling of fuels behaviour - modeling of defects and fission products in the UO 2 ceramics by ab initio calculations, cladding and assembly materials, pellet-cladding interaction, advanced UO 2 and MOX ceramics, mechanical behaviour of the fuel assembly, fuel during a loss of coolant accident, fuel during a reactivity accident, fuel during a serious accident, fuel management inside reactor cores, fuel cycle materials balance, long-term behaviour of the spent fuel, fuel of boiling water reactors; 3 - the fuel of liquid metal fast reactors: fast neutrons radiation

  11. Does rim microstructure formation degrade the fuel rod performance?

    Baron, D.; Spino, J.

    2002-01-01

    High burnup extension of LWR fuel is progressing to reduce the total process flow and eventually the costs of the nuclear fuel cycle. A particular fuel restructuring at high burnups, commonly observed at the periphery of LWR fuel pellets (rim structure), but also in FBR fuels to some extent and in the Plutonium rich clusters of the MOX Fuels, was considered a priori as a limitation for burnup extension. Since more than ten years this rim effect have been deeply investigated. Its causes and consequences are however not yet totally elucidated. The three steps actually identified of this phenomenon are first a progressive disappearing of the intra-granular Xenon, the outset of numerous 0.5 to 1 m pores and finally a grain subdivision around the pores. Penalty of the porosity increase on the thermal conductivity is obvious. One expect the fission gases to remain trapped in the rim porosity up to a 75 MWd/kgUO 2 local burnup. Above this threshold, 15 to 20 % of the fission gases seem to be quickly released. Microindentation tests conducted at ITU have shown the rim structure to resist fracture extension under punching. It is still open whether this implies certain ductility and viscosity of the material, or if it corresponds to stress relaxation by microcracking. Whatever the case be, it is suggested that the rim material would be able to decrease the interaction stresses and to equalise the cladding strains during a power ramp. Moreover, in the RIA tests, it was concluded so far that the grain de-cohesion caused by gas expansion at the grain boundaries was responsible for the cladding strain and failure. However, not the rim zone was affected by grain de-cohesion but the region adjacent to it. Therefore, in front of the question whether the rim structure degrades the fuel rod behaviour, we continue to argue on its benefit for fuel burnup extension. (author)

  12. Fuel pellet

    Hayashi, K.

    1980-01-01

    Fuel pellet for insertion into a cladding tube in order to form a fuel element or a fuel rod. The fuel pellet has got a belt-like projection around its essentially cylindrical lateral circumferential surface. The upper and lower edges in vertical direction of this belt-like projection are wave-shaped. The projection is made of the same material as the bulk pellet. Both are made in one piece. (orig.) [de

  13. Fuel chemistry and pellet-clad interaction related to high burnup fuel. Proceedings of the technical committee

    2000-10-01

    The purpose of the meeting was to review new developments in clad failures. Major findings regarding the causes of clad failures are presented in this publication, with the main topics being fuel chemistry and fission product behaviour, swelling and pellet-cladding mechanical interaction, cladding failure mechanism at high burnup, thermal properties and fuel behaviour in off-normal conditions. This publication contains 17 individual presentations delivered at the meeting; each of them was indexed separately

  14. Fossil Fuels.

    Crank, Ron

    This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fossil fuels. Some topics covered are historic facts, development of fuels, history of oil production, current and future trends of the oil industry, refining fossil fuels, and environmental problems. Material in each unit may…

  15. Fuel element

    1974-01-01

    A new fuel can with a loose bottom and head is described. The fuel bar is attached to the loose bottom and head with two grid poles keeping the distance between bottom and head. A bow-shaped handle is attached to the head so that the fuel bar can be lifted from the can

  16. Prediction of power-ramp defects in CANDU fuel

    Gillespie, P.; Wadsworth, S.; Daniels, T.

    2010-01-01

    Power ramps result in fuel pellet expansion and can lead to fuel sheath failures by fission product induced stress corrosion cracking (SCC). Historically, empirical models fit to experimental test data were used to predict the onset of power-ramp failures in CANDU fuel. In 1988, a power-ramped fuel defect event at PNGS-1 led to the refinement of these empirical models. This defect event has recently been re-analyzed and the empirical model updated. The empirical model is supported by a physically based model which can be used to extrapolate to fuel conditions (density, burnup) outside of the 1988 data set. (author)

  17. LMFBR fuel analysis. Task A: oxide fuel dynamics. Final report, October 1977--September 1978

    Dhir, V.K.; Frank, M.; Kastenberg, W.E.; McKone, T.E.

    1979-03-01

    Three aspects of LMFBR safety are discussed. The first concerns the potential reactivity effects of whole core fuel motion prior to pin failure in low ramp rate transient overpower accidents. The second concerns the effects of flow blockages following pin failure on the coolability of a core following an unprotected overpower transient. The third aspect concerns the safety related implications of using thorium based fuels in LMFBR's

  18. Fuel pin behavior under slow ramp-type transient-overpower conditions in the CABRI-FAST experiments

    Fukano, Yoshitaka; Onoda, Yuichi; Sato, Ikken; Charpenel, Jean

    2009-01-01

    In the CABRI-FAST experimental program, four in-pile tests were performed with slow power-ramp-type transient-overpower conditions (called hereafter as 'slow TOP') to study transient fuel pin behavior under inadvertent control rod withdrawal events in liquid metal cooled fast breeder reactors. Annular-pellet fuel pins were used in three tests, while a solid-pellet fuel pin was used in the other test. All of these pins were pre-irradiated in Phenix. The slow TOP test with a solid-pellet fuel pin was realized as a comparatory test against an existing test (E12) in the CABRI-2 program. In the CABRI-FAST test (BCF1), a power ramp rate of 3% Po/s was applied, while in the CABRI-2 test, 1% Po/s was adopted. Moreover, overpower condition was maintained for a few seconds beyond the observed pin failure in the BCF1 test. In spite of the different power ramp rates, evaluated fuel thermal conditions at the observed failure time are quite similar. The continued overpower condition in the BCF1 test resulted in gradual degradation of the pin structure providing information effective for evaluation of various accident scenarios. Three slow TOP tests with the annular fuel in the CABRI-FAST program resulted in no pin failure showing high failure threshold. Based on post-test examination data and a theoretical evaluation, it was concluded that intra-pin free spaces, such as central hole, macroscopic cracks and fuel-cladding gap effectively mitigated fuel cladding mechanical interaction. It was also clarified that cavity pressurization became effective only in case of very large amount of fuel melting. Furthermore, such cavity pressurization was effectively mitigated by a molten-fuel squirting into the upper blanket region pushing the blanket pellets upward. These CABRI FAST slow TOP tests, in combination with the existing CABRI and TREAT tests, provided an extended slow TOP test database with various fuel and transient conditions. (author)

  19. 14 CFR 23.967 - Fuel tank installation.

    2010-01-01

    ... the engine compartment may act as the wall of an integral tank. (d) Each fuel tank must be isolated... loads without permanent deformation or failure under the conditions of §§ 23.365 and 23.843 of this part. A bladder-type fuel cell, if used, must have a retaining shell at least equivalent to a metal fuel...

  20. Spent fuel storage pool

    Murakami, Naoshi.

    1996-01-01

    Fences are disposed to a fuel exchange floor surrounding the upper surface of a fuel pool for preventing overflow of pool water. The fences comprise a plurality of flat boards arranged in parallel with each other in the longitudinal direction while being vertically inclined, and slits are disposed between the boards for looking down the pool. Further, the fences comprise wide boards and are constituted so as to be laid horizontally on the fuel exchange floor in a normal state and uprisen by means of the signals from an earthquake sensing device. Even if pool water is overflow from the fuel pool by the vibrations occurred upon earthquake and flown out to the floor of the fuel exchange floor, the overflow from the fuel exchange floor is prevented by the fences. An operator who monitors the fuel pool can observe the inside of the fuel pool through the slits formed to the fences during normal operation. The fences act as resistance against overflowing water upon occurrence of an earthquake thereby capable of reducing the overflowing amount of water due to the vibrations of pool water. The effect of preventing overflowing water can be enhanced. (N.H.)