Nondestructive examination of Oconee 1 fuel assemblies after four cycles of irradiation

International Nuclear Information System (INIS)

Pyecha, T.D.; Mayer, J.T.; Guthrie, B.A. III; Riordan, J.E.

1980-12-01

Five B and W Mark B (15 x 15) pressurized water reactor fuel assemblies were nondestructively examined after four cycles of irradiation in the Oconee 1 reactor. Four of the five assemblies examined had a burnup of 40,000 MWd/mtU; the fifth assembly had a burnup of 36,800 MWd/mtU. This effort is part of a Department of Energy program to improve uranium utilization by extending the burnup of light water reactor fuel. The examinations were conducted in the Oconee 1 and 2 spent fuel storage pool. Data obtained included fuel assembly and fuel rod dimensions, water channel spacings, spacer grid and holddown spring forces, fuel column stack and axial gap lengths, and crud samples. The results indicate that the assemblies performed well through four cycles of operation; all of the data were within design limits

Fuel and fuel pin behaviour in a high burnup fast breeder fuel subassembly: Results of destructive post-irradiation examinations of the KNK II/1 fuel subassembly NY-205

International Nuclear Information System (INIS)

Patzer, G.

1991-05-01

The report gives a summarizing overview of the design characteristics, of the irradiation history and of the results of the destructive post-irradiation examinations of the fuel pins of the high-burnup fuel subassembly NY-205 of the KNK II first core. This element was operated for about 10 years and reached a maximum local burnup of 175 MWd/kg(HM) and a maximum neutron dose of 67 dpa-NRT. The main design data of this subassembly agree with those of the SNR 300 Mark-Ia, and it reached more than twice of the burnup and a similar neutron dose as foreseen for the SNR 300 fuel subassemblies [de

Design studies for the Mark-III core of experimental multi-purpose VHTR

International Nuclear Information System (INIS)

Yasuno, Takehiko; Miyamoto, Yoshiaki; Mitake, Susumu; Shindo, Ryuiti; Arai, Taketoshi

1979-08-01

The Mark-III core in the first conceptual design made in 1975 is a fundamental core for VHTR. Subsequently, further design studies were made fuel loading scheme and control rod withdrawal sequence for the core to increase its safety margin (shutdown margin, etc.) and operational margin (minimum Reynolds number, maximum fuel temperature, etc.). It was shown that the Mark-III should exhibit the performance expected of VHTR, unless changes are made in the preconditions for its nuclear, thermal-hydraulic design. Also, the needs as below were indicated: (1) reasonable core design criteria and guidelines, (2) fuel-loading-scheme requirements in fuel management, fuel misloading and reactor operation, (3) confirmation on precision of the core design method and its further refinement. (author)

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

International Nuclear Information System (INIS)

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)

Flow sheet development for the dissolution of unirradiated Mark 42 fuel tubes in F-Canyon, Part II

International Nuclear Information System (INIS)

Murray, A.M.

1999-01-01

Two dissolution flow sheets were tested for the desorption of unirradiated Mark 42 fuel tubes. Both the aluminum (from the can, cladding, and fuel core) and the plutonium oxide (PuO 2 ) are dissolved simultaneously, i.e., a co-dissolution flow sheet. In the first series of tests, 0.15 and 0.20 molar (M) potassium fluoride (KF) solutions were used and the dissolution extended over several days. In the other series of tests, solutions with higher concentrations of fluoride (0.25 to 0.30 M) were used. Calcium fluoride (CaF 2 ) was used in those tests as the fluoride source

26 CFR 1.1296-1 - Mark to market election for marketable stock.

Science.gov (United States)

2010-04-01

... 26 Internal Revenue 11 2010-04-01 2010-04-01 true Mark to market election for marketable stock. 1....1296-1 Mark to market election for marketable stock. (a) Definitions—(1) Eligible RIC. An eligible RIC... included in gross income by the company pursuant to such mark to market election with respect to such stock...

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

International Nuclear Information System (INIS)

1979-11-01

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

Neutron spectra in two beam ports of a TRIGA Mark III reactor with HEU fuel

International Nuclear Information System (INIS)

Vega C, H. R.; Hernandez D, V. M.; Paredes G, L.; Aguilar, F.

2012-10-01

Before to change the HEU for Leu fuel of the ININ's TRIGA Mark III nuclear reactor the neutron spectra were measured in two beam ports using 5 and 10 W. Measurements were carried out in a tangential and a radial beam port using a Bonner sphere spectrometer. It was found that neutron spectra are different in the beam ports, in radial beam port the amplitude of thermal and fast neutrons are approximately the same while, in the tangential beam port thermal neutron peak is dominant. In the radial beam port the fluence-to-ambient dose equivalent factors are 131±11 and 124±10 p Sv-cm 2 for 5 and 10 W respectively while in the tangential beam port the fluence-to-ambient dose equivalent factor is 55±4 p Sv-cm 2 for 10 W. (Author)

Spent nuclear fuel management. Moving toward a century of spent fuel management: A view from the halfway mark

International Nuclear Information System (INIS)

Shephard, L.

2004-01-01

Full text: A half-century ago, President Eisenhower in his 1953 'Atoms for Peace' speech, offered nuclear technology to other nations as part of a broad nuclear arms control initiative. In the years that followed, the nuclear power generation capabilities of many nations has helped economic development and contributed to the prosperity of the modern world. The growth of nuclear power, while providing many benefits, has also contributed to an increasing global challenge over safe and secure spent fuel management. Over 40 countries have invested in nuclear energy, developing over 400 nuclear power reactors. Nuclear power supplies approximately 16% of the global electricity needs. With the finite resources and challenges of fossil fuels, nuclear power will undoubtedly become more prevalent in the future, both in the U.S. and abroad. We must address this inevitability with new paradigms for managing a global nuclear future. Over the past fifty years, the world has come to better understand the strong interplay between all elements of the nuclear fuel cycle, global economics, and global security. In the modern world, the nuclear fuel cycle can no longer be managed as a simple sequence of technological, economic and political challenges. Rather it must be seen, and managed, as a system of strongly interrelated challenges. Spent fuel management, as one element of the nuclear fuel system, cannot be relegated to the back-end of the fuel cycle as only a disposal or storage issue. There exists a wealth of success and experience with spent fuel management over the past fifty years. We must forge this experience with a global systems perspective, to reshape the governing of all aspects of the nuclear fuel cycle, including spent fuel management. This session will examine the collective experience of spent fuel management enterprises, seeking to shape the development of new management paradigms for the next fifty years. (author)

Regulation of glucose homeostasis by KSR1 and MARK2.

Directory of Open Access Journals (Sweden)

Paula J Klutho

Full Text Available Protein scaffolds control the intensity and duration of signaling and dictate the specificity of signaling through MAP kinase pathways. KSR1 is a molecular scaffold of the Raf/MEK/ERK MAP kinase cascade that regulates the intensity and duration of ERK activation. Relative to wild-type mice, ksr1⁻/⁻ mice are modestly glucose intolerant, but show a normal response to exogenous insulin. However, ksr1⁻/⁻ mice also demonstrate a three-fold increase in serum insulin levels in response to a glucose challenge, suggesting a role for KSR1 in insulin secretion. The kinase MARK2 is closely related to C-TAK1, a known regulator of KSR1. Mice lacking MARK2 have an increased rate of glucose disposal in response to exogenous insulin, increased glucose tolerance, and are resistant to diet-induced obesity. mark2⁻/⁻ksr1⁻/⁻ (DKO mice were compared to wild type, mark2⁻/⁻, and ksr1⁻/⁻ mice for their ability to regulate glucose homeostasis. Here we show that disruption of KSR1 in mark2⁻/⁻ mice reverses the increased sensitivity to exogenous insulin resulting from MARK2 deletion. DKO mice respond to exogenous insulin similarly to wild type and ksr1⁻/⁻ mice. These data suggest a model whereby MARK2 negatively regulates insulin sensitivity in peripheral tissue through inhibition of KSR1. Consistent with this model, we found that MARK2 binds and phosphorylates KSR1 on Ser392. Phosphorylation of Ser392 is a critical regulator of KSR1 stability, subcellular location, and ERK activation. These data reveal an unexpected role for the molecular scaffold KSR1 in insulin-regulated glucose metabolism.

Calculation of Savannah River K Reactor Mark-22 assembly LOCA/ECS power limits

International Nuclear Information System (INIS)

Fischer, S.R.; Farman, R.F.; Birdsell, S.A.

1992-01-01

This paper summarizes the results of TRAC-PF1/MOD3 calculations of Mark-22 fuel assembly of loss-of-coolant accident/emergency cooling system (LOCA/ECS) power limits for the Savannah River Site (SRS) K Reactor. This effort was part of a larger effort undertaken by the Los Alamos National Laboratory for the US Department of Energy to perform confirmatory power limits calculations for the SRS K Reactor. A method using a detailed three-dimensional (3D) TRAC model of the Mark-22 fuel assembly was developed to compute LOCA/ECS power limits. Assembly power was limited to ensure that no point on the fuel assembly walls would exceed the local saturation temperature. The detailed TRAC model for the Mark-22 assembly consisted of three concentric 3D vessel components which simulated the two targets, two fuel tubes, and three main flow channels of the fuel assembly. The model included 100% eccentricity between the assembly annuli and a 20% power tilt. Eccentricity in the radial alignment of the assembly annuli arises because axial spacer ribs that run the length of the fuel and targets are used. Wall-shear, interfacial-shear, and wall heat-transfer correlations were developed and implemented in TRAC-PF1/MOD3 specifically for modeling flow and heat transfer in the narrow ribbed annuli encountered in the Mark-22 fuel assembly design. We established the validity of these new constitutive models using separate-effects benchmarks. TRAC system calculations of K Reactor indicated that the limiting ECS-phase accident is a double-ended guillonite break in a process water line at the pump discharge (i.e., a PDLOCA). The fuel assembly with the minimum cooling potential is identified from this system calculation. Detailed assembly calculations then were performed using appropriate boundary conditions obtained from this limiting system LOCA. Coolant flow rates and pressure boundary conditions were obtained from this system calculation and applied to the detailed assembly model

Criticality and shielding calculations for containers in dry of spent fuel of TRIGA Mark III reactor of ININ

International Nuclear Information System (INIS)

Barranco R, F.

2015-01-01

In this thesis criticality and shielding calculations to evaluate the design of a container of dry storage of spent nuclear fuel generated in research reactors were made. The design of such container was originally proposed by Argentina and Brazil, and the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico. Additionally, it is proposed to modify the design of this container to store spent fuel 120 that are currently in the pool of TRIGA Mark III reactor, the Nuclear Center of Mexico and calculations and analyzes are made to verify that the settlement of these fuel elements is subcritical limits and dose rates to workers and the general public are not exceeded. These calculations are part of the design criteria for security protection systems in dry storage system (Dss for its acronym in English) proposed by the Nuclear Regulatory Commission (NRC) of the United States. To carry out these calculations simulation codes of Monte Carlo particle transport as MCNPX and MCNP5 were used. The initial design (design 1) 78 intended to store spent fuel with a maximum of 115. The ININ has 120 fuel elements and spent 3 control rods (currently stored in the reactor pool). This leads to the construction of two containers of the original design, but for economic reasons was decided to modify (design 2) to store in a single container. Criticality calculations are performed to 78, 115 and fresh fuel elements 124 within the container, to the two arrangements described in Chapter 4, modeling the three-dimensional geometry assuming normal operating conditions and accident. These calculations are focused to demonstrate that the container will remain subcritical, that is, that the effective multiplication factor is less than 1, in particular not greater than 0.95 (as per specified by the NRC). Spent fuel 78 and 124 within the container, both gamma radiation to neutron shielding calculations for only two cases were simulated. First actinides and fission products generated

Safety analysis and optimization of the core fuel reloading for the Moroccan TRIGA Mark-II reactor

International Nuclear Information System (INIS)

Nacir, B.; Boulaich, Y.; Chakir, E.; El Bardouni, T.; El Bakkari, B.; El Younoussi, C.

2014-01-01

Highlights: • Additional fresh fuel elements must be added to the reactor core. • TRIGA reactor could safely operate around 2 MW power with 12% fuel elements. • Thermal–hydraulic parameters were calculated and the safety margins are respected. • The 12% fuel elements will have no influence on the safety of the reactor. - Abstract: The Moroccan TRIGA MARK II reactor core is loaded with 8.5% in weight of uranium standard fuel elements. Additional fresh fuel elements must periodically be added to the core in order to remedy the observed low power and to return to the initial reactivity excess at the End Of Cycle. 12%-uranium fuel elements are available to relatively improve the short fuel lifetime associated with standard TRIGA elements. These elements have the same dimensions as standards elements, but with different uranium weight. The objective in this study is to demonstrate that the Moroccan TRIGA reactor could safely operate, around 2 MW power, with new configurations containing these 12% fuel elements. For this purpose, different safety related thermal–hydraulic parameters have been calculated in order to ensure that the safety margins are largely respected. Therefore, the PARET model for this TRIGA reactor that was previously developed and combined with the MCNP transport code in order to calculate the 3-D temperature distribution in the core and all the most important parameters like the axial distribution of DNBR (Departure from Nucleate Boiling Ratio) across the hottest channel. The most important conclusion is that the 12% fuel elements utilization will have no influence on the safety of the reactor while working around 2 MW power especially for configurations based on insertions in C and D-rings

Nonlinear analyses of spent-fuel racks for consolidated fuel loading

International Nuclear Information System (INIS)

Kabir, A.F.; Godha, P.C.; Malik, L.E.; Bolourchi, S.

1987-01-01

Storage racks for spent-fuel assemblies in nuclear power plants are designed to withstand various combinations of loads generated by gravity, seismic, thermal, and accidental fuel drops. Due to the need for storing increased amounts of spent fuel in the existing fuel pools, many nuclear power utilities are evaluating existing fuel racks to safely carry the additional loads. The current study presents the seismic analyses of existing fuel racks of Northeast Utility Company's Millstone Unit Number 1 (BWR Mark I) nuclear plant to accommodate a 2:1 fuel consolidation. This objective requires rigorous nonlinear analyses to establish the full available capacities of the racks and thereby avoid expensive modifications or minimize any needed upgrades

TRIGA Mark II Ljubljana - spent fuel transportation

International Nuclear Information System (INIS)

Ravnik, M.; Dimic, V.

2008-01-01

The most important activity in 1999 was shipment of the spent fuel elements back to the United States for final disposal. This activity started already in 1998 with some governmental support. In July 1999 all spent fuel elements (219 pieces) from the TRIGA research reactor in Ljubljana were shipped back to the United Stated by the ship from the port Koper in Slovenia. At the same time shipment of the spent fuel from the research reactor in Pitesti, Romania, and the research reactor in Rome, Italy, was conducted. During the loading the radiation exposure to the workers was rather low. The loading and shipment of the spent nuclear fuel went very smoothly and according the accepted time table. During the last two years the TRIGA research reactor in Ljubljana has been in operation about 1100 hours per year and without any undesired shut-down. (authors)

MARK/Par1 Kinase Is Activated Downstream of NMDA Receptors through a PKA-Dependent Mechanism.

Directory of Open Access Journals (Sweden)

Laura P Bernard

Full Text Available The Par1 kinases, also known as microtubule affinity-regulating kinases (MARKs, are important for the establishment of cell polarity from worms to mammals. Dysregulation of these kinases has been implicated in autism, Alzheimer's disease and cancer. Despite their important function in health and disease, it has been unclear how the activity of MARK/Par1 is regulated by signals from cell surface receptors. Here we show that MARK/Par1 is activated downstream of NMDA receptors in primary hippocampal neurons. Further, we show that this activation is dependent on protein kinase A (PKA, through the phosphorylation of Ser431 of Par4/LKB1, the major upstream kinase of MARK/Par1. Together, our data reveal a novel mechanism by which MARK/Par1 is activated at the neuronal synapse.

46 CFR 167.55-1 - Draft marks and draft indicating systems.

Science.gov (United States)

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false Draft marks and draft indicating systems. 167.55-1... NAUTICAL SCHOOL SHIPS Special Markings Required § 167.55-1 Draft marks and draft indicating systems. (a... aft to the location of the draft marks, due to a raked stem or cut away skeg, the draft must be...

Technical specifications manual for the MARK-1 pulsed ionizing radiation detection system. Volume 1

Energy Technology Data Exchange (ETDEWEB)

Lawrence, R.S.; Harker, Y.D.; Jones, J.L.; Hoggan, J.M.

1993-03-01

The MARK-1 detection system was developed by the Idaho National Engineering Laboratory for the US Department of Energy Office of Arms Control and Nonproliferation. The completely portable system was designed for the detection and analysis of intense photon emissions from pulsed ionizing radiation sources. This manual presents the technical design specifications for the MARK-1 detection system and was written primarily to assist the support or service technician in the service, calibration, and repair of the system. The manual presents the general detection system theory, the MARK-1 component design specifications, the acquisition and control software, the data processing sequence, and the system calibration procedure. A second manual entitled: Volume 2: Operations Manual for the MARK-1 Pulsed Ionizing Radiation Detection System (USDOE Report WINCO-1108, September 1992) provides a general operational description of the MARK-1 detection system. The Operations Manual was written primarily to assist the field operator in system operations and analysis of the data.

Experience with advanced driver fuels in EBR-II

International Nuclear Information System (INIS)

Lahm, C.E.; Koenig, J.F.; Pahl, R.G.; Porter, D.L.; Crawford, D.C.

1992-01-01

The Experimental Breeder Reactor II (EBR-II) is a complete nuclear power plant, incorporating a pool-type liquid-metal reactor (LMR) with a fuel-power thermal output of 62.5 MW and an electrical output of 20 MW. Initial criticality was in 1961, utilizing a metallic driver fuel design called the Mark-I. The fuel design has evolved over the last 30 yr, and significant progress has been made on improving performance. The first major innovations were incorporated into the Mark-II design, and burnup then increased dramatically. This design performed successfully, and fuel element lifetime was limited by subassembly hardware performance rather than the fuel element itself. Transient performance of the fuel was also acceptable and demonstrated the ability of EBR-II to survive severe upsets such as a loss of flow without scram. In the mid 1980s, with renewed interest in metallic fuels and Argonne's integral fast reactor (IFR) concept, the Mark-II design was used as the basis for new designs, the Mark-III and Mark-IV. In 1987, the Mark-III design began qualification testing to become a driver fuel for EBR-II. This was followed in 1989 by the Mark-IIIA and Mark-IV designs. The next fuel design, the Mark-V, is being planned to demonstrate the utilization of recycled fuel. The fuel cycle facility attached to EBR-II is being refurbished to produce pyroprocessed recycled fuel as part of the demonstration of the IFR

Marks on the petroleum fiscality

International Nuclear Information System (INIS)

2007-02-01

This document offers some marks on the petroleum fiscality in France: the taxes as the 'accises' and the 'TVA', the part of the taxes in the sale price at the service station, the comparison with other countries of Europe, the tax revenues and the Government budget. It provides also marks on the fuels prices formation (margins), the world petroleum markets (supply and demand) and the part of the petroleum companies on the petroleum market. (A.L.B.)

NotaMark industrial laser marking system: a new security marking technology

Science.gov (United States)

Moreau, Vincent G.

2004-06-01

Up until now, the only variable alphanumeric data which could be added to banknotes was the number, applied by means of impact typographical numbering boxes. As an additional process or an alternative to this mechanical method, a non-contact laser marking process can be used offering high quality and greater levels of flexibility. For this purpose KBA-GIORI propose an exclusive laser marking solution called NotaMark. The laser marking process NotaMark is the ideal solution for applying variable data and personalizing banknotes (or any other security documents) with a very high resolution, for extremely large production volumes. A completely integrated solution has been developed comprised of laser light sources, marking head units, and covers and extraction systems. NotaMark allows the marking of variable data by removing locally and selectively, specific printed materials leaving the substrate itself untouched. A wide range of materials has already been tested extensively. NotaMark is a new security feature which is easy to identify and difficult to counterfeit, and which complies with the standard mechanical and chemical resistance tests in the security printing industry as well as with other major soiling tests. The laser marking process opens up a whole new range of design possibilities and can be used to create a primary security feature such as numbering, or to enhance the value of existing features.

Sensor system for fuel transport vehicle

Science.gov (United States)

Earl, Dennis Duncan; McIntyre, Timothy J.; West, David L.

2016-03-22

An exemplary sensor system for a fuel transport vehicle can comprise a fuel marker sensor positioned between a fuel storage chamber of the vehicle and an access valve for the fuel storage chamber of the vehicle. The fuel marker sensor can be configured to measure one or more characteristics of one or more fuel markers present in the fuel adjacent the sensor, such as when the marked fuel is unloaded at a retail station. The one or more characteristics can comprise concentration and/or identity of the one or more fuel markers in the fuel. Based on the measured characteristics of the one or more fuel markers, the sensor system can identify the fuel and/or can determine whether the fuel has been adulterated after the marked fuel was last measured, such as when the marked fuel was loaded into the vehicle.

Assessment of TRAC-PF1/MOD3 Mark-22 assembly model using SRL ''A'' tank single-assembly flow experiments

International Nuclear Information System (INIS)

Fischer, S.R.; Lam, K.; Lin, J.C.

1991-01-01

This paper summarizes the results of an assessment of our TRAC-PF1/MOD3 Mark-22 prototype fuel assembly model against single-assembly data obtained from the ''A'' Tank single-assembly tests that were performed at the Savannah River Laboratory. We felt the data characterize prototypic assembly behavior over a range of air-water flow conditions of interest for loss-of-coolant accident (LOCA) calculations. This study was part of a benchmarking effort performed to evaluate and validate a multiple-assembly, full-plant model that is being developed by Los Alamos National Laboratory to study various aspects of the Savannah River plant operating conditions, including LOCA transients, using TRAC-PF1/MOD3 Version 1.10. The results of this benchmarking effort demonstrate that TRAC-PF1/MOD3 is capable pf calculating plenum conditions and assembly flows during conditions thought to be typical of the Emergency Cooling System (ECS) phase of a LOCA. 10 refs., 12 fig

1L Mark-IV Target Design Review

Energy Technology Data Exchange (ETDEWEB)

Koehler, Paul E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-11-16

This presentation includes General Design Considerations; Current (Mark-III) Lower Tier; Mark-III Upper Tier; Performance Metrics; General Improvements for Material Science; General Improvements for Nuclear Science; Improving FOM for Nuclear Science; General Design Considerations Summary; Design Optimization Studies; Expected Mark-IV Performance: Material Science; Expected Mark-IV Performance: Nuclear Science (Disk); Mark IV Enables Much Wider Range of Nuclear-Science FOM Gains than Mark III; Mark-IV Performance Summary; Rod or Disk? Center or Real FOV?; and Project Cost and Schedule.

Fuel assembly for a nuclear reactor

International Nuclear Information System (INIS)

Gjertsen, R.K.; Tower, S.N.; Huckestein, E.A.

1982-01-01

A fuel assembly for a nuclear reactor comprises a 5x5 array of guide tubes in a generally 20x20 array of fuel elements, the guide tubes being arranged to accommodate either control rods or water displacer rods. The fuel assembly has top and bottom Inconel (Registered Trade Mark) grids and intermediate Zircaloy grids in engagement with the guide tubes and supporting the fuel elements and guide tubes while allowing flow of reactor coolant through the assembly. (author)

H3K56me1 marks a spot for PCNA

DEFF Research Database (Denmark)

Lee, Sung-Po; Jasencakova, Zusana; Groth, Anja

2012-01-01

In the current issue of Molecular Cell, Yu et al. (2012) establish H3K56 monomethylation (H3K56me1) as a new mammalian chromatin mark, imposed by the G9a methyltransferase and recognized by the replication clamp PCNA....

Full-Scale Mark II CRT Program data report, 1

International Nuclear Information System (INIS)

Namatame, Ken; Kukita, Yutaka; Yamamoto, Nobuo; Shiba, Masayoshi

1979-12-01

The Full-Scale Mark II CRT (Containment Response Test) Program was initiated in April 1976 to provide a full-scale data basis for the evaluation of the pressure suppression pool hydrodynamic loads associated with a hypothetical LOCA in a BWR Mark II Containment. The test facility, completed in March 1979, is 1/18 in volume of a typical 1100 MWe Mark II, and has a wetwell which is a full-scale replica of one 20 0 -sector of that of the reference Mark II. The present report documents experimental data from TEST 0002, a medium size (100 mm) water blowdown test, performed by Hitachi Ltd. for JAERI as the second of the four shakedown tests. Test data is provided for the vessel depressurization, the pressure and temperature responses in the test containment, and especially for the chugging phenomena associated with low flux steam condensation in the pool. (author)

Evaluation of fuel-temperature feedback mechanisms in TRAC-PF1/MOD2/NESTLE

International Nuclear Information System (INIS)

Knepper, Paula L.; Feltus, Madeline; Hochreiter, L.E.; Ivanov, Kostadin

1999-01-01

Coupled spatial kinetics and thermal-hydraulics system codes provide a means to model transient nuclear reactor behavior more accurately. Transients marked by strong perturbations, both with thermal-hydraulics and neutronics, such as a control-rod ejection or a main steam-line break, are especially of interest. It is now feasible to model complex reactor behavior with a coupled thermal-hydraulics and spatial kinetics code that provides a means to forecast safety margins. Recently, the Transient Reactor Analysis Code (TRAC)-PF1/MOD2, Version 5.4.25, was coupled with the NESTLE code. This coupled code (TRAC-PF1/MOD2/NESTLE) is used to examine effective fuel-temperature models. The Electric Power Research Institute (EPRI) rod-ejection benchmark was analyzed to evaluate the influence of effective fuel temperature. The rod-ejection transient tests only the fuel-rod, heat-conduction coupling. The coolant thermal-hydraulic coupling is not tested because of the speed of the transient. The neutronics solution changes extremely rapidly, whereas the convective heat transfer at the fuel surface requires more time to influence the coolant temperature of the system. The need to model the response of the system coolant temperature is not crucial in this analysis. The influence of the effective fuel temperature is the key component of this study. Various models were examined using the coupled code to calculate effective fuel temperatures. The influence of different, effective fuel-temperature models on the coupled-code results is studied. Three effective fuel-temperature models are examined: (l) volume average effective fuel temperature, (2) the effective fuel-temperature model suggested by the Office of Economic Cooperation and Development (OECD) rod-ejection benchmark, and (3) the NESTLE effective fuel-temperature model. A discussion is provided describing the effective fuel-temperature models examined in TRAC-PF1/MOD2/NESTLE and the influence of effective fuel temperature in

Preliminary investigations of a mixed standard-flip core for a TRIGA Mark II

International Nuclear Information System (INIS)

Ringle, John C.; Johnson, A.G.; Anderson, T.V.

1974-01-01

Several years ago it became apparent that due to our rapidly- increasing use rate, we would need a substantial amount of new fuel by late 1974 or early 1975. After investigations and discussions with GA, we decided that FLIP fuel would best meet our requirements for maximum fuel economy and high peak pulsing power. A proposal was submitted to the AEC for fuel assistance, and late in 1973 we were awarded a grant of $61,875. This will allow us to buy 3 FLIP-fueled-follower control rods, 1 instrumented FLIP fuel element, and 26 standard FLIP elements, giving us then a mixed core of approximately one-third FLIP and two-thirds standard elements. License amendments to accommodate this change are rather straightforward; modifications to the Technical Specifications will be somewhat more involved. The largest revisions which we envision are to our Safety Analysis Report. Although a few reactors have operated with a full FLIP core, and a few others have converted to mixed standard-FLIP cores, none of these has a standard Mark II core configuration. Those who have already converted to a mixed core have data and calculations which may be helpful to us, but the extent to which we can use these remains to be seen. The present status of our investigations into the analysis of a mixed standard-FLIP core will be presented. Any problems in calculational methods, finding appropriate data, modifications to Technical Specifications, etc., will be identified, and suggestions and help in these areas will be welcomed. (author)

Preliminary investigations of a mixed standard-flip core for a TRIGA Mark II

Energy Technology Data Exchange (ETDEWEB)

Ringle, John C; Johnson, A G; Anderson, T V [Oregon State University (United States)

1974-07-01

Several years ago it became apparent that due to our rapidly- increasing use rate, we would need a substantial amount of new fuel by late 1974 or early 1975. After investigations and discussions with GA, we decided that FLIP fuel would best meet our requirements for maximum fuel economy and high peak pulsing power. A proposal was submitted to the AEC for fuel assistance, and late in 1973 we were awarded a grant of $61,875. This will allow us to buy 3 FLIP-fueled-follower control rods, 1 instrumented FLIP fuel element, and 26 standard FLIP elements, giving us then a mixed core of approximately one-third FLIP and two-thirds standard elements. License amendments to accommodate this change are rather straightforward; modifications to the Technical Specifications will be somewhat more involved. The largest revisions which we envision are to our Safety Analysis Report. Although a few reactors have operated with a full FLIP core, and a few others have converted to mixed standard-FLIP cores, none of these has a standard Mark II core configuration. Those who have already converted to a mixed core have data and calculations which may be helpful to us, but the extent to which we can use these remains to be seen. The present status of our investigations into the analysis of a mixed standard-FLIP core will be presented. Any problems in calculational methods, finding appropriate data, modifications to Technical Specifications, etc., will be identified, and suggestions and help in these areas will be welcomed. (author)

Technical specifications manual for the MARK-1 pulsed ionizing radiation detection system

International Nuclear Information System (INIS)

Lawrence, R.S.; Harker, Y.D.; Jones, J.L.; Hoggan, J.M.

1993-03-01

The MARK-1 detection system was developed by the Idaho National Engineering Laboratory for the US Department of Energy Office of Arms Control and Nonproliferation. The completely portable system was designed for the detection and analysis of intense photon emissions from pulsed ionizing radiation sources. This manual presents the technical design specifications for the MARK-1 detection system and was written primarily to assist the support or service technician in the service, calibration, and repair of the system. The manual presents the general detection system theory, the MARK-1 component design specifications, the acquisition and control software, the data processing sequence, and the system calibration procedure. A second manual entitled: Volume 2: Operations Manual for the MARK-1 Pulsed Ionizing Radiation Detection System (USDOE Report WINCO-1108, September 1992) provides a general operational description of the MARK-1 detection system. The Operations Manual was written primarily to assist the field operator in system operations and analysis of the data

The new area monitoring system and the fuel database of the TRIGA Mark II reactor in Vienna

International Nuclear Information System (INIS)

Villa, M.; Boeck, H.; Hofbauer, M.; Schwarz, V.

2004-01-01

The 250 kW TRIGA Mark-II reactor operates since March 1962 at the Atominstitut, Vienna, Austria. Its main tasks are nuclear education and training in the fields of neutron- and solid state physics, nuclear technology, reactor safety, radiochemistry, radiation protection and dosimetry, and low temperature physics and fusion research. Academic research is carried out by students in the above mentioned fields coordinated and supervised by about 70 staff members with the aim of a masters- or PhD degree in one of the above mentioned areas. After 25 years of successful operation, it was necessary to exchange the old area monitoring system with a new digital one. The purpose of the new system is the permanent control of the reactor hall, the primary and secondary cooling system and the monitoring of the ventilation system. The paper describes the development and implementation of the new area monitoring system. The second topic in this paper describes the development of the new fuel database. Since March 7th, 1962, the TRIGA Mark II reactor Vienna operates with an average of 263 MWh per year, which corresponds to a uranium burn-up of 13.7 g per year. Presently we have 81 TRIGA fuel elements in the core, 55 of them are old aluminium clad elements from the initial criticality while the rest are stainless steel clad elements which had been added later to compensate the uranium consumption. Because 67 % of the elements are older than 40 years, it was necessary to put the history of every element in a database, to get an easy access to all the relevant data for every element in our facility. (author)

75 FR 14669 - Regulation of Fuels and Fuel Additives: Changes to Renewable Fuel Standard Program

Science.gov (United States)

2010-03-26

... RINs from producers of the renewable fuel. The obligated parties do not need lead time for construction... fuels and new limits on renewable biomass feedstocks. This rulemaking marks the first time that... advanced biofuel and multiple cellulosic-based fuels with their 60% threshold. Additional fuel pathways...

Safety analysis calculations for a mixed and full FLIP core in a TRIGA Mark II

International Nuclear Information System (INIS)

Ringle, John C.; Hornyik, K.; Robinson, A.H.; Anderson, T.V.; Johnson, A.G.

1976-01-01

The Oregon State TRIGA Reactor will be reloading with FLIP fuel in August 1976. As we are the first Mark II TRIGA with a circular grid pattern and graphite reflector to utilize FLIP fuel, the safety analysis calculations performed at other facilities using FLIP were only of limited use to us. A multigroup, multiregion, one-dimensional diffusion theory code was used to calculate power densities in six different operational cores - mixed to full FLIP. Pulsing characteristics were obtained from a computer code based on point kinetics, with adiabatic heating of the fuel, linear temperature dependence of the specific heat, and prompt fuel temperature feedback coefficient. The results of all pertinent calculations will be presented. (author)

Thermal power calibration of the TRIGA Mark I IPR-R1 reactor during the upgrading tests to 250 kW

International Nuclear Information System (INIS)

Mesquita, Amir Zacarias; Maretti, Fausto Junior; Rezende, Hugo Cesar

2002-01-01

This paper presents the results and the methodology used to calibrate the thermal power of the TRIGA MARK I IPR-R1 Reactor in CDTN, Belo Horizonte, Brazil. This calibration was realized during the operation tests carried out to allow the reactor power upgrade from the current 100 kW to 250 kW. The methodology consisted in the measurement of the inlet and outlet temperature and the water flow in the primary cooling loop. The thermal balance together with the thermal losses gave the thermal power. There were made three sequences of tests. The first rising of the thermal power was made with the usual configuration of the core (59 fuel elements). After the changing of the ion chambers position and the control rod and the increase of the number of fuels (63 fuel elements), a new evaluation of the thermal power was accomplished, having been obtained a thermal power of 234 kW, for an indication of 250 kW in the lineal channel. After the return of the core to the initial configuration (59 fuel elements), it took place a new test, getting back the reactor to the power level of 100 kW. (author)

Criticality and shielding calculations for containers in dry of spent fuel of TRIGA Mark III reactor of ININ; Calculos de criticidad y blindaje para contenedores en seco de combustible gastado del reactor Triga Mark III del ININ

Energy Technology Data Exchange (ETDEWEB)

Barranco R, F.

2015-07-01

In this thesis criticality and shielding calculations to evaluate the design of a container of dry storage of spent nuclear fuel generated in research reactors were made. The design of such container was originally proposed by Argentina and Brazil, and the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico. Additionally, it is proposed to modify the design of this container to store spent fuel 120 that are currently in the pool of TRIGA Mark III reactor, the Nuclear Center of Mexico and calculations and analyzes are made to verify that the settlement of these fuel elements is subcritical limits and dose rates to workers and the general public are not exceeded. These calculations are part of the design criteria for security protection systems in dry storage system (Dss for its acronym in English) proposed by the Nuclear Regulatory Commission (NRC) of the United States. To carry out these calculations simulation codes of Monte Carlo particle transport as MCNPX and MCNP5 were used. The initial design (design 1) 78 intended to store spent fuel with a maximum of 115. The ININ has 120 fuel elements and spent 3 control rods (currently stored in the reactor pool). This leads to the construction of two containers of the original design, but for economic reasons was decided to modify (design 2) to store in a single container. Criticality calculations are performed to 78, 115 and fresh fuel elements 124 within the container, to the two arrangements described in Chapter 4, modeling the three-dimensional geometry assuming normal operating conditions and accident. These calculations are focused to demonstrate that the container will remain subcritical, that is, that the effective multiplication factor is less than 1, in particular not greater than 0.95 (as per specified by the NRC). Spent fuel 78 and 124 within the container, both gamma radiation to neutron shielding calculations for only two cases were simulated. First actinides and fission products generated

A narratological analysis of Mark 12:1-12: The plot of the Gospel of ...

African Journals Online (AJOL)

The purpose of this article is an attempt to read Mark 12: 1-12 in terms of the plot of the Gospel. Firstly a brief survey is given of the development of the term plot from Aristotle to the present, thereafter an own methodological point of departure concerning plot is formulated in order to study the plot of Mark. The conclusions ...

Multi-fuel reformers for fuel cells used in transportation. Phase 1: Multi-fuel reformers

Science.gov (United States)

1994-05-01

DOE has established the goal, through the Fuel Cells in Transportation Program, of fostering the rapid development and commercialization of fuel cells as economic competitors for the internal combustion engine. Central to this goal is a safe feasible means of supplying hydrogen of the required purity to the vehicular fuel cell system. Two basic strategies are being considered: (1) on-board fuel processing whereby alternative fuels such as methanol, ethanol or natural gas stored on the vehicle undergo reformation and subsequent processing to produce hydrogen, and (2) on-board storage of pure hydrogen provided by stationary fuel processing plants. This report analyzes fuel processor technologies, types of fuel and fuel cell options for on-board reformation. As the Phase 1 of a multi-phased program to develop a prototype multi-fuel reformer system for a fuel cell powered vehicle, the objective of this program was to evaluate the feasibility of a multi-fuel reformer concept and to select a reforming technology for further development in the Phase 2 program, with the ultimate goal of integration with a DOE-designated fuel cell and vehicle configuration. The basic reformer processes examined in this study included catalytic steam reforming (SR), non-catalytic partial oxidation (POX) and catalytic partial oxidation (also known as Autothermal Reforming, or ATR). Fuels under consideration in this study included methanol, ethanol, and natural gas. A systematic evaluation of reforming technologies, fuels, and transportation fuel cell applications was conducted for the purpose of selecting a suitable multi-fuel processor for further development and demonstration in a transportation application.

Enamel-based mark performance for marking Chinese mystery snail Bellamya chinensis

Science.gov (United States)

Wong, Alec; Allen, Craig R.; Hart, Noelle M.; Haak, Danielle M.; Pope, Kevin L.; Smeenk, Nicholas A.; Stephen, Bruce J.; Uden, Daniel R.

2013-01-01

The exoskeleton of gastropods provides a convenient surface for carrying marks, and i the interest of improving future marking methods our laboratory assessed the performance of an enamel paint. The endurance of the paint was also compared to other marking methods assessed in the past. We marked the shells of 30 adult Chinese mystery snails Bellamya chinensis and held them in an aquarium for 181 days. We observed no complete degradation of any enamel-paint mark during the 181 days. The enamel-paint mark was superior to a nai;-polish mark, which lasted a median of 100 days. Enamel-paint marks also have a lower rate of loss (0.00 month-1 181 days) than plastic bee tags (0.01 month-1, 57 days), gouache paint (0.07 month-1, 18.5 days), or car body paint from studies found in scientific literature. Legibility of enamel-paint marks had a median lifetime of 102 days. The use of enamel paint on the shells of gastropods is a viable option for studies lasting up to 6 months. Furthermore, visits to capture-mark-recapture site 1 year after application of enamel-paint marks on B. chinesnis shells produced several individuals on which the enamel paint was still visible, although further testing is required to clarify durability over longer periods.

Fuel Summary for Peach Bottom Unit 1 High-Temperature Gas-Cooled Reactor Cores 1 and 2

Energy Technology Data Exchange (ETDEWEB)

Karel I. Kingrey

2003-04-01

This fuel summary report contains background and summary information for the Peach Bottom Unit 1, High-Temperature, Gas-Cooled Reactor Cores 1 and 2. This report contains detailed information about the fuel in the two cores, the Peach Bottom Unit 1 operating history, nuclear parameters, physical and chemical characteristics, and shipping and storage canister related data. The data in this document have been compiled from a large number of sources and are not qualified beyond the qualification of the source documents. This report is intended to provide an overview of the existing data pertaining to spent fuel management and point to pertinent reference source documents. For design applications, the original source documentation must be used. While all referenced sources are available as records or controlled documents at the Idaho National Engineering and Environmental Laboratory (INEEL), some of the sources were marked as informal or draft reports. This is noted where applicable. In some instances, source documents are not consistent. Where they are known, this document identifies those instances and provides clarification where possible. However, as stated above, this document has not been independently qualified and such clarifications are only included for information purposes. Some of the information in this summary is available in multiple source documents. An effort has been made to clearly identify at least one record document as the source for the information included in this report.

Helicobacter pylori CagA Inhibits PAR1-MARK Family Kinases by Mimicking Host Substrates

Energy Technology Data Exchange (ETDEWEB)

Nesic, D.; Miller, M; Quinkert, Z; Stein, M; Chait, B; Stebbins, C

2010-01-01

The CagA protein of Helicobacter pylori interacts with numerous cellular factors and is associated with increased virulence and risk of gastric carcinoma. We present here the cocrystal structure of a subdomain of CagA with the human kinase PAR1b/MARK2, revealing that a CagA peptide mimics substrates of this kinase family, resembling eukaryotic protein kinase inhibitors. Mutagenesis of conserved residues central to this interaction renders CagA inactive as an inhibitor of MARK2.

SLARette Mark 2 system

International Nuclear Information System (INIS)

Burnett, D.J.

1992-01-01

The SLAR (Spacer Location and Repositioning) program has developed the technology and tooling necessary to locate and reposition the fuel channel spacers that separate the pressure tube from the calandria tube in a CANDU reactor. The in-channel SLAR tool contains all the inspection probes, and is capable of moving spacers under remote control. The SLAR inspection computer system translates all eddy currents and ultrasonic signals from the in-channel tool into various graphic displays. The in-channel SLAR tool can be delivered and manipulated in a fuel channel by either a SLAR delivery machine or a SLARette delivery machine. The SLAR delivery machine consists of a modified fuelling machine, and is capable of operating under totally remote control in automatic or semi-automatic mode. The SLARette delivery machine is a smaller less automated version, which was designed to be quickly installed, operated, and removed from a limited number of fuel channels during regular annual maintenance outages. This paper describes the design and operation of the SLARette Mark 2 system. 5 figs

EcoMark 2.0

DEFF Research Database (Denmark)

Guo, Chenjuan; Yang, Bin; Andersen, Ove

2015-01-01

Eco-routing is a simple yet effective approach to substantially reducing the environmental impact, e.g., fuel consumption and greenhouse gas (GHG) emissions, of vehicular transportation. Eco-routing relies on the ability to reliably quantify the environmental impact of vehicles as they travel...... in a spatial network. The procedure of quantifying such vehicular impact for road segments of a spatial network is called eco-weight assignment. EcoMark 2.0 proposes a general framework for eco-weight assignment to enable eco-routing. It studies the abilities of six instantaneous and five aggregated models......, and experiments for assessing the utility of the impact models in assigning eco-weights. The application of EcoMark 2.0 indicates that the instantaneous model EMIT and the aggregated model SIDRA-Running are suitable for assigning eco-weights under varying circumstances. In contrast, other instantaneous models...

Burn-up TRIGA Mark II benchmark experiment

International Nuclear Information System (INIS)

Persic, A.; Ravnik, M.; Zagar, T.

1998-01-01

Different reactor codes are used for calculations of reactor parameters. The accuracy of the programs is tested through comparison of the calculated values with the experimental results. Well-defined and accurately measured benchmarks are required. The experimental results of reactivity measurements, fuel element reactivity worth distribution and fuel-up measurements are presented in this paper. The experiments were performed with partly burnt reactor core. The experimental conditions were well defined, so that the results can be used as a burn-up benchmark test case for a TRIGA Mark II reactor calculations.(author)

DUPIC fuel cycle economics assessment (1)

International Nuclear Information System (INIS)

Choi, H. B.; Roh, G. H.; Kim, D. H.

1999-04-01

This is a state-of-art report that describes the current status of the DUPIC fuel cycle economics analysis conducted by the DUPIC fuel compatibility assessment group of the DUPIC fuel development project. For the DUPIC fuel cycle economics analysis, the DUPIC fuel compatibility assessment group has organized the 1st technical meeting composed of 8 domestic specialists from government, academy, industry, etc. and a foreign specialist of hot-cell design from TRI on July 16, 1998. This report contains the presentation material of the 1st technical meeting, published date used for the economics analysis and opinions of participants, which could be utilized for further DUPIC fuel cycle and back-end fuel cycle economics analyses. (author). 11 refs., 7 charts

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

International Nuclear Information System (INIS)

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

Kinase Associated-1 Domains Drive MARK/PAR1 Kinases to Membrane Targets by Binding Acidic Phospholipids

Energy Technology Data Exchange (ETDEWEB)

Moravcevic, Katarina; Mendrola, Jeannine M.; Schmitz, Karl R.; Wang, Yu-Hsiu; Slochower, David; Janmey, Paul A.; Lemmon, Mark A. (UPENN-MED)

2011-09-28

Phospholipid-binding modules such as PH, C1, and C2 domains play crucial roles in location-dependent regulation of many protein kinases. Here, we identify the KA1 domain (kinase associated-1 domain), found at the C terminus of yeast septin-associated kinases (Kcc4p, Gin4p, and Hsl1p) and human MARK/PAR1 kinases, as a membrane association domain that binds acidic phospholipids. Membrane localization of isolated KA1 domains depends on phosphatidylserine. Using X-ray crystallography, we identified a structurally conserved binding site for anionic phospholipids in KA1 domains from Kcc4p and MARK1. Mutating this site impairs membrane association of both KA1 domains and intact proteins and reveals the importance of phosphatidylserine for bud neck localization of yeast Kcc4p. Our data suggest that KA1 domains contribute to coincidence detection, allowing kinases to bind other regulators (such as septins) only at the membrane surface. These findings have important implications for understanding MARK/PAR1 kinases, which are implicated in Alzheimer's disease, cancer, and autism.

Fuel shipment experience, fuel movements from the BMI-1 transport cask

International Nuclear Information System (INIS)

Bauer, Thomas L.; Krause, Michael G.

1986-01-01

The University of Texas at Austin received two shipments of irradiated fuel elements from Northrup Aircraft Corporation on April 11 and 16, 1985. A total of 59 elements consisting of standard and instrumented TRIGA fuel were unloaded from the BMI-1 shipping cask. At the time of shipment, the Northrup core burnup was approximately 50 megawatt days with fuel element radiation levels, after a cooling time of three months, of approximately 1.75 rem/hr at 3 feet. In order to facilitate future planning of fuel shipment at the UT facility and other facilities, a summary of the recent transfer process including several factors which contributed to its success are presented. Numerous color slides were made of the process for future reference by UT and others involved in fuel transfer and handling of the BMI-1 cask

Eddy current examination of the nuclear fuel elements of IPR-R1 research reactor

International Nuclear Information System (INIS)

Silva, Roger F.; Frade, Rangel T.; Oliveira, Paulo F.; Silva, Marlucio A.; Silva Junior, Silverio F.

2015-01-01

Tubes of AISI 304 stainless steel as well as tubes of Aluminum 1100-F are used as cladding of the fuel elements of TRIGA MARK 1 nuclear research reactor. Usually, these tubes are periodically inspected by means of visual test and sipping test. The visual test allows the detection of changes occurred at the external fuel elements surface, such as those promoted by corrosion processes. However, this test method cannot be used for detection of internal discontinuities at the tube walls. Sipping test allows the detection of fuel elements in which the cladding has failed, but it is not able to determine the place where the discontinuity is located. In turn, eddy current testing, an electromagnetic nondestructive test method, allows the detection of discontinuities and monitoring their growth. In this paper, a study about the use of eddy current testing for detection and characterization of discontinuities in the fuel elements cladding is proposed. The study involves the development of probes able to operate in underwater inspections, the design and manufacture of reference standards and the development of a test methodology to perform the evaluations. (author)

Eddy current examination of the nuclear fuel elements of IPR-R1 research reactor

Energy Technology Data Exchange (ETDEWEB)

Silva, Roger F.; Frade, Rangel T.; Oliveira, Paulo F.; Silva, Marlucio A.; Silva Junior, Silverio F., E-mail: rfs@cdtn.br, E-mail: rtf@cdtn.br, E-mail: pfo@cdtn.br, E-mail: mas@cdtn.br, E-mail: silvasf@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2015-07-01

Tubes of AISI 304 stainless steel as well as tubes of Aluminum 1100-F are used as cladding of the fuel elements of TRIGA MARK 1 nuclear research reactor. Usually, these tubes are periodically inspected by means of visual test and sipping test. The visual test allows the detection of changes occurred at the external fuel elements surface, such as those promoted by corrosion processes. However, this test method cannot be used for detection of internal discontinuities at the tube walls. Sipping test allows the detection of fuel elements in which the cladding has failed, but it is not able to determine the place where the discontinuity is located. In turn, eddy current testing, an electromagnetic nondestructive test method, allows the detection of discontinuities and monitoring their growth. In this paper, a study about the use of eddy current testing for detection and characterization of discontinuities in the fuel elements cladding is proposed. The study involves the development of probes able to operate in underwater inspections, the design and manufacture of reference standards and the development of a test methodology to perform the evaluations. (author)

Fabrication of fuel elements interplay between typical SNR Mark Ia specifications and the fuel element fabrication

International Nuclear Information System (INIS)

Biermann, W.K.; Heuvel, H.J.; Pilate, S.; Vanderborck, Y.; Pelckmans, E.; Vanhellemont, G.; Roepenack, H.; Stoll, W.

1987-01-01

The core and fuel were designed for the SNR-300 first core by Interatom GmbH and Belgonucleaire. The fuel was fabricated by Alkem/RBU and Belgonucleaire. Based on the preparation of drawings and specifications and on the results of the prerun fabrication, an extensive interplay took place between design requirements, specifications, and fabrication processes at both fuel plants. During start-up of pellet and pin fabrication, this solved such technical questions as /sup 239/Pu equivalent linear weight, pellet density, stoichiometry of the pellets, and impurity content. Close cooperation of designers and manufacturers has allowed manufacture of 205 fuel assemblies without major problems

HTGR fuel particle crusher: Mark 2 design

Energy Technology Data Exchange (ETDEWEB)

Baer, J.W.

1979-06-01

The double-roll crusher for fracturing the silicon carbide coatings of high-temperature gas-cooled reactor (HTGR) fuel particles has been redesigned to improve the equipment. The housing was simplified and reduced to a two-piece assembly; the bearings were changed to accommodate thermal effects; the bearing protection seals were improved with triple redundancy; the bearing preload arrangement was simplified and improved; and localized wear areas were reinforced with better materials or special treatment. In addition, the crusher drive was changed for impoved characteristics and an increase in power.

HTGR fuel particle crusher: Mark 2 design

International Nuclear Information System (INIS)

Baer, J.W.

1979-06-01

The double-roll crusher for fracturing the silicon carbide coatings of high-temperature gas-cooled reactor (HTGR) fuel particles has been redesigned to improve the equipment. The housing was simplified and reduced to a two-piece assembly; the bearings were changed to accommodate thermal effects; the bearing protection seals were improved with triple redundancy; the bearing preload arrangement was simplified and improved; and localized wear areas were reinforced with better materials or special treatment. In addition, the crusher drive was changed for impoved characteristics and an increase in power

Criticality calculation in TRIGA MARK II PUSPATI Reactor using Monte Carlo code

International Nuclear Information System (INIS)

Rafhayudi Jamro; Redzuwan Yahaya; Abdul Aziz Mohamed; Eid Abdel-Munem; Megat Harun Al-Rashid; Julia Abdul Karim; Ikki Kurniawan; Hafizal Yazid; Azraf Azman; Shukri Mohd

2008-01-01

A Monte Carlo simulation of the Malaysian nuclear reactor has been performed using MCNP Version 5 code. The purpose of the work is the determination of the multiplication factor (k e ff) for the TRIGA Mark II research reactor in Malaysia based on Monte Carlo method. This work has been performed to calculate the value of k e ff for two cases, which are the control rod either fully withdrawn or fully inserted to construct a complete model of the TRIGA Mark II PUSPATI Reactor (RTP). The RTP core was modeled as close as possible to the real core and the results of k e ff from MCNP5 were obtained when the control fuel rods were fully inserted, the k e ff value indicates the RTP reactor was in the subcritical condition with a value of 0.98370±0.00054. When the control fuel rods were fully withdrawn the value of k e ff value indicates the RTP reactor is in the supercritical condition, that is 1.10773±0.00083. (Author)

Nuclear and radiological safety in the substitution process of the fuel HEU to LEU 30/20 in the Reactor TRIGA Mark III of the ININ

International Nuclear Information System (INIS)

Hernandez G, J.

2012-10-01

Inside the safety initiative in the international ambit, with the purpose of reducing the risks associated with the use of high enrichment nuclear fuels (HEU) for different proposes to the peaceful uses of the nuclear energy, Mexico contributes by means of the substitution of the high enrichment fuel HEU for low enrichment fuel LEU 30/20 in the TRIGA Mark III Reactor, belonging to Instituto Nacional de Investigaciones Nucleares (ININ). The conversion process was carried out by means of the following activities: analysis of the proposed core, reception and inspection of the fuel LEU 30/20, the discharge of the fuels of the mixed reactor core, shipment of the fuels HEU fresh and irradiated to the origin country, reload activities with the fuels LEU 30/20 and parameters measurement of the core operation. In order to maintaining the personnel's integrity and infrastructure associated to the Reactor, during the whole process the measurements of nuclear and radiological safety were controlled to detail, in execution with the license requirements of the installation. This work describes the covering activities and radiological inspections more relevant, as well as the measurements of radiological control implemented with base in the estimate of the equivalent dose of the substitution process. (Author)

Applicable regulations and development of surveillance experiments of criticality approach in the TRIGA III Mark reactor; Normativa aplicable y desarrollo de experimentos de vigilancia de aproximacion a criticidad en el reactor Triga Mark III

Energy Technology Data Exchange (ETDEWEB)

Gonzalez M, J L; Aguilar H, F; Rivero G, T; Sainz M, E [Instituto nacional de Investigaciones Nucleares, Departamento de Automatizacion, A.P. 18-1027, Col. Escandon, 11801 Mexico D.F. (Mexico)

2000-07-01

In the procedure elaborated to repair the vessel of TRIGA III Mark reactor is required to move toward two tanks of temporal storage the fuel elements which are in operation and the spent fuel elements which are in decay inside the reactor pool. The National Commission of Nuclear Safety and Safeguards (CNSNS) has requested as protection measure that it is carried out a surveillance of the criticality approach of the temporal storages. This work determines the main regulation aspects that entails an experiment of criticality approach, moreover, informing about the results obtained in the developing of this experiments. The regulation aspects are not exclusives for this work in the TRIGA Mark III reactor but they also apply toward any assembling of fissile material. (Author)

Core map generation for the ITU TRIGA Mark II research reactor using Genetic Algorithm coupled with Monte Carlo method

Energy Technology Data Exchange (ETDEWEB)

Türkmen, Mehmet, E-mail: tm@hacettepe.edu.tr [Nuclear Engineering Department, Hacettepe University, Beytepe Campus, Ankara (Turkey); Çolak, Üner [Energy Institute, Istanbul Technical University, Ayazağa Campus, Maslak, Istanbul (Turkey); Ergün, Şule [Nuclear Engineering Department, Hacettepe University, Beytepe Campus, Ankara (Turkey)

2015-12-15

Highlights: • Optimum core maps were generated for the ITU TRIGA Mark II Research Reactor. • Calculations were performed using a Monte Carlo based reactor physics code, MCNP. • Single-Objective and Multi-Objective Genetic Algorithms were used for the optimization. • k{sub eff} and ppf{sub max} were considered as the optimization objectives. • The generated core maps were compared with the fresh core map. - Abstract: The main purpose of this study is to present the results of Core Map (CM) generation calculations for the İstanbul Technical University TRIGA Mark II Research Reactor by using Genetic Algorithms (GA) coupled with a Monte Carlo (MC) based-particle transport code. Optimization problems under consideration are: (i) maximization of the core excess reactivity (ρ{sub ex}) using Single-Objective GA when the burned fuel elements with no fresh fuel elements are used, (ii) maximization of the ρ{sub ex} and minimization of maximum power peaking factor (ppf{sub max}) using Multi-Objective GA when the burned fuels with fresh fuels are used. The results were obtained when all the control rods are fully withdrawn. ρ{sub ex} and ppf{sub max} values of the produced best CMs were provided. Core-averaged neutron spectrum, and variation of neutron fluxes with respect to radial distance were presented for the best CMs. The results show that it is possible to find an optimum CM with an excess reactivity of 1.17 when the burned fuels are used. In the case of a mix of burned fuels and fresh fuels, the best pattern has an excess reactivity of 1.19 with a maximum peaking factor of 1.4843. In addition, when compared with the fresh CM, the thermal fluxes of the generated CMs decrease by about 2% while change in the fast fluxes is about 1%.Classification: J. Core physics.

Performance of PARR-1 with LEU Fuel

International Nuclear Information System (INIS)

Pervez, S.; Latif, M.; Bokhari, I.H.; Bakhtyar, S.

2005-01-01

Pakistan Research Reactor (PARR-1) went critical in 1965 with HEU fuel. The reactor core was converted to LEU fuel with power upgradation from 5 MW to 10 MW in 1992. The reactor has been operated with LEU fuel for about 10,000 hours and has produced about 66,000 MWh energy up to now. Average burn up of the irradiated fuel is about 42 %. The fuel performance during the last 12 years has been excellent. Post irradiation visual inspection of the fuel has revealed no abnormality. During operation there have been no signs of releases in the pool water establishing the full integrity of this fuel. The reactor has been mainly utilized for radioisotope production, beam tube experiments including neutron diffraction studies, neutron radiography etc. Studies have been completed to operate the reactor with a mixed core (HEU + LEU) to utilize the less burned HEU fuel elements. A major project of production of fission Moly using PARR-1 is in the final stages. (author)

A cold demonstration of fuel consolidation. Part 1

International Nuclear Information System (INIS)

Matheson, J.E.

1989-01-01

Spent fuel consolidation is an option for increasing spent fuel storage capacities being considered by many utilities. The process of consolidating fuel involves separating the fuel rods from the structural frame which holds them in a square array. The rods are then repackaged into a tightly packed bundle which occupies about half the cross-sectional area of fuel assembly. Thus approximately twice as much fuel can be stored in the underwater racks at a spent fuel storage pool. There have been several demonstrations of fuel consolidation to date. The focus of this paper is the development and subsequent demonstration program of a shear/compactor

Marks on the petroleum fiscality; Reperes sur la fiscalite petroliere

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-02-15

This document offers some marks on the petroleum fiscality in France: the taxes as the 'accises' and the 'TVA', the part of the taxes in the sale price at the service station, the comparison with other countries of Europe, the tax revenues and the Government budget. It provides also marks on the fuels prices formation (margins), the world petroleum markets (supply and demand) and the part of the petroleum companies on the petroleum market. (A.L.B.)

Fuel cells 101

Energy Technology Data Exchange (ETDEWEB)

Taylor, B.

2003-06-01

A capsule history of fuel cells is given, beginning with the first discovery in 1839 by William Grove, a Welsh judge who, when experimenting with electrolysis discovered that by re-combining the two components of electrolysis (water and oxygen) an electric charge was produced. A century later, in 1958, Francis Thomas Bacon, a British scientist demonstrated the first working fuel cell stack, a technology which was licensed and used in the Apollo spacecraft. In Canada, early research on the development of fuel cells was carried out at the University of Toronto, the Defence Research Establishment and the National Research Council. Most of the early work concentrated on alkaline and phosphoric acid fuel cells. In 1983, Ballard Research began the development of the electrolyte membrane fuel cell, which marked the beginning of Canada becoming a world leader in fuel cell technology development. The paper provides a brief account of how fuel cells work, describes the distinguishing characteristics of the various types of fuel cells (alkaline, phosphoric acid, molten-carbonate, solid oxide, and proton exchange membrane types) and their principal benefits. The emphasis is on proton exchange membrane fuel cells because they are the only fuel cell technology that is appropriate for providing primary propulsion power onboard a vehicle. Since vehicles are by far the greatest consumers of fossil fuels, it follows that proton exchange membrane fuel cells will have the greatest potential impact on both environmental matters and on our reliance on oil as our primary fuel. Various on-going and planned fuel cell demonstration projects are also described. 1 fig.

Fuel oil and kerosene sales 1994

International Nuclear Information System (INIS)

1995-01-01

This publication contains the 1994 survey results of the ''Annual Fuel Oil and Kerosene Sales Report'' (Form EIA-821). This is the sixth year that the survey data have appeared in a separate publication. Prior to the 1989 report, the statistics appeared in the Petroleum Marketing Annual (PMA)for reference year 1988 and the Petroleum Marketing Monthly (PMM) for reference years 1984 through 1987. The 1994 edition marks the 11th annual presentation of the results of the ongoing ''Annual Fuel Oil and Kerosene Sales Report'' survey. Distillate and residual fuel oil sales continued to move in opposite directions during 1994. Distillate sales rose for the third year in a row, due to a growing economy. Residual fuel oil sales, on the other hand, declined for the sixth year in a row, due to competitive natural gas prices, and a warmer heating season than in 1993. Distillate fuel oil sales increased 4.4 percent while residual fuel oil sales declined 1.6 percent. Kerosene sales decreased 1.4 percent in 1994

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

International Nuclear Information System (INIS)

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

1977-01-01

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

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

International Nuclear Information System (INIS)

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

1984-10-01

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

Dissolution of FFTF vendor fuel

International Nuclear Information System (INIS)

Lerch, R.E.

1979-08-01

Dissolution experiments were performed on FFTF vendor fuel (both mechanically mixed and coprecipitated) during 1974, 1975, and 1976. A marked improvement was noted in the completeness of fuel dissolution from 1974 to 1976. The reason for this is unknown but may have been attributable to slight changes in fuel fabrication conditions. In general, the bulk of the fuel pellets tested dissolved to greater than 99.9% in nitric acid alone

Dissolution of FFTF vendor fuel

Energy Technology Data Exchange (ETDEWEB)

Lerch, R.E.

1979-08-01

Dissolution experiments were performed on FFTF vendor fuel (both mechanically mixed and coprecipitated) during 1974, 1975, and 1976. A marked improvement was noted in the completeness of fuel dissolution from 1974 to 1976. The reason for this is unknown but may have been attributable to slight changes in fuel fabrication conditions. In general, the bulk of the fuel pellets tested dissolved to greater than 99.9% in nitric acid alone.

Probabilistic evaluation of fuel element performance by the combined use of a fast running simplistic and a detailed deterministic fuel performance code

International Nuclear Information System (INIS)

Misfeldt, I.

1980-01-01

A comprehensive evaluation of fuel element performance requires a probabilistic fuel code supported by a well bench-marked deterministic code. This paper presents an analysis of a SGHWR ramp experiment, where the probabilistic fuel code FRP is utilized in combination with the deterministic fuel models FFRS and SLEUTH/SEER. The statistical methods employed in FRP are Monte Carlo simulation or a low-order Taylor approximation. The fast-running simplistic fuel code FFRS is used for the deterministic simulations, whereas simulations with SLEUTH/SEER are used to verify the predictions of FFRS. The ramp test was performed with a SGHWR fuel element, where 9 of the 36 fuel pins failed. There seemed to be good agreement between the deterministic simulations and the experiment, but the statistical evaluation shows that the uncertainty on the important performance parameters is too large for this ''nice'' result. The analysis does therefore indicate a discrepancy between the experiment and the deterministic code predictions. Possible explanations for this disagreement are discussed. (author)

A survey of processes for producing hydrogen fuel from different sources for automotive-propulsion fuel cells

Energy Technology Data Exchange (ETDEWEB)

Brown, L.F.

1996-03-01

Seven common fuels are compared for their utility as hydrogen sources for proton-exchange-membrane fuel cells used in automotive propulsion. Methanol, natural gas, gasoline, diesel fuel, aviation jet fuel, ethanol, and hydrogen are the fuels considered. Except for the steam reforming of methanol and using pure hydrogen, all processes for generating hydrogen from these fuels require temperatures over 1000 K at some point. With the same two exceptions, all processes require water-gas shift reactors of significant size. All processes require low-sulfur or zero-sulfur fuels, and this may add cost to some of them. Fuels produced by steam reforming contain {approximately}70-80% hydrogen, those by partial oxidation {approximately}35-45%. The lower percentages may adversely affect cell performance. Theoretical input energies do not differ markedly among the various processes for generating hydrogen from organic-chemical fuels. Pure hydrogen has severe distribution and storage problems. As a result, the steam reforming of methanol is the leading candidate process for on-board generation of hydrogen for automotive propulsion. If methanol unavailability or a high price demands an alternative process, steam reforming appears preferable to partial oxidation for this purpose.

A Cherenkov viewing device for used-fuel verification

International Nuclear Information System (INIS)

Attas, E.M.; Chen, J.D.; Young, G.J.

1990-01-01

A Cherenkov viewing device (CVD) has been developed to help verify declared inventories of used nuclear fuel stored in water bays. The device detects and amplifies the faint ultraviolet Cherenkov glow from the water surrounding the fuel, producing a real-time visible image on a phosphor screen. Quartz optics, a UV-pass filter and a microchannel-plate image-intensifier tube serve to form the image, which can be photographed or viewed directly through an eyepiece. Normal fuel bay lighting does not interfere with the Cherenkov light image. The CVD has been successfully used to detect anomalous PWR, BWR and CANDU (CANada Deuterium Uranium: registered trademark) fuel assemblies in the presence of normal-burnup assemblies stored in used-fuel bays. The latest version of the CVD, known as Mark IV, is being used by inspectors from the International Atomic Energy agency for verification of light-water power-reactor fuel. Its design and operation are described, together with plans for further enhancements of the instrumentation. (orig.)

Characterization of the TRIGA Mark II reactor full-power steady state

Energy Technology Data Exchange (ETDEWEB)

Cammi, Antonio, E-mail: antonio.cammi@polimi.it [Politecnico di Milano – Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via La Masa 34, 20156 Milano (Italy); Zanetti, Matteo [Politecnico di Milano – Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via La Masa 34, 20156 Milano (Italy); Chiesa, Davide; Clemenza, Massimiliano; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica [University of Milano-Bicocca, Physics Department “G. Occhialini” and INFN Section, Piazza dell’Ateneo Nuovo, 20126 Milan (Italy); Magrotti, Giovanni; Prata, Michele; Salvini, Andrea [University of Pavia, Applied Nuclear Energy Laboratory (L.E.N.A.), Via Gaspare Aselli 41, 27100 Pavia (Italy)

2016-04-15

Highlights: • Full-power steady state characterization of the TRIGA Mark II reactor. • Monte Carlo and Multiphysics simulation of the TRIGA Mark II reactor. • Sub-cooled boiling effects in the TRIGA Mark II reactor. • Thermal feedback effects in the TRIGA Mark II reactor. • Experimental data based validation. - Abstract: In this paper, the characterization of the full-power steady state of the TRIGA Mark II nuclear reactor at the University of Pavia is achieved by coupling the Monte Carlo (MC) simulation for neutronics with the “Multiphysics” model for thermal-hydraulics. Neutronic analyses have been carried out with a MCNP5 based MC model of the entire reactor system, already validated in fresh fuel and zero-power configurations (in which thermal effects are negligible) and using all available experimental data as a benchmark. In order to describe the full-power reactor configuration, the temperature distribution in the core must be established. To evaluate this, a thermal-hydraulic model has been developed, using the power distribution results from the MC simulation as input. The thermal-hydraulic model is focused on the core active region and takes into account sub-cooled boiling effects present at full reactor power. The obtained temperature distribution is then entered into the MC model and a benchmark analysis is carried out to validate the model in fresh fuel and full-power configurations. An acceptable correspondence between experimental data and simulation results concerning full-power reactor criticality proves the reliability of the adopted methodology of analysis, both from the perspective of neutronics and thermal-hydraulics.

The study of time-dependent neutronics parameters of the 2MW TRIGA Mark II Moroccan research reactor using BUCAL1 computer code

International Nuclear Information System (INIS)

Bakkari, B. El; Nacir, B.; El Younoussi, C.; Boulaich, Y.; Riyach, I.; Otmani, S.; Marcih, I.; Elbadri, H.; El Bardouni, T; Merroun, O.; Boukhal, H.; Zoubair, M.; Htet, A.; Chakir, M.

2010-01-01

The 2-MW TRIGA MARK II research reactor at Centre National de l'Energie, des Sciences et des Techniques Nucleaires (CNESTEN) achieved initial criticality on May 2, 2007 with 71 fuel elements. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower and training and production of radioisotopes for their use in agriculture, industry and medicine. This work aims to study the time-dependent neutronics parameters of the TRIGA reactor for elaborating and planning of an in-core fuel management strategy to maximize the utilization of the TRIGA fluxes, using a new elaborated burnup computer code called 'BUCAL1'. The code can be used to aid in analysis, prediction, and optimization of fuel burnup performance in a nuclear reactor. It was developed to incorporate the neutron absorption tally/reaction information generated directly by MCNP5 code in the calculation of fissioned or neutron-transmuted isotopes for multi-fueled regions. The use of Monte Carlo method and punctual cross section data characterizing the MCNP code allows an accurate simulation of neutron life cycle in the reactor, and the integration of data on the entire energy spectrum, thus a more accurate estimation of results than deterministic code can do. Also, for the purpose of this study, a full-model of the TRIGA reactor was developed using the MCNP5 code. The validation of the MCNP model of the TRIGA reactor was made by benchmarking the reactivity experiments. (author)

Decommissioning of the ICI TRIGA Mark I reactor

International Nuclear Information System (INIS)

Parry, D.R.; England, M.R.; Ward, A.; Green, D.

2000-01-01

This paper considers the fuel removal, transportation and subsequent decommissioning of the ICI TRIGA Mark I Reactor at Billingham, UK. BNFL Waste Management and Decommissioning carried out this work on behalf of ICI. The decommissioning methodology was considered in the four stages to be described, namely Preparatory Works, Reactor Defueling, Intermediate Level Waste Removal and Low Level Waste Removal. This paper describes the principal methodologies involved in the defueling of the reactor and subsequent decommissioning operations, highlighting in particular the design and safety case methodologies used in order to achieve a solution which was completed without incident or accident and resulted in a cumulative radiation dose to personnel of only 1.57 mSv. (author)

21 CFR 1.382 - What labeling or marking requirements apply to a detained article of food?

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 1 2010-04-01 2010-04-01 false What labeling or marking requirements apply to a detained article of food? 1.382 Section 1.382 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES GENERAL GENERAL ENFORCEMENT REGULATIONS Administrative Detention of Food for...

Review of oxidation rates of DOE spent nuclear fuel : Part 1 : nuclear fuel

International Nuclear Information System (INIS)

Hilton, B.A.

2000-01-01

The long-term performance of Department of Energy (DOE) spent nuclear fuel (SNF) in a mined geologic disposal system depends highly on fuel oxidation and subsequent radionuclide release. The oxidation rates of nuclear fuels are reviewed in this two-volume report to provide a baseline for comparison with release rate data and technical rationale for predicting general corrosion behavior of DOE SNF. The oxidation rates of nuclear fuels in the DOE SNF inventory were organized according to metallic, Part 1, and non-metallic, Part 2, spent nuclear fuels. This Part 1 of the report reviews the oxidation behavior of three fuel types prototypic of metallic fuel in the DOE SNF inventory: uranium metal, uranium alloys and aluminum-based dispersion fuels. The oxidation rates of these fuels were evaluated in oxygen, water vapor, and water. The water data were limited to pure water corrosion as this represents baseline corrosion kinetics. Since the oxidation processes and kinetics discussed in this report are limited to pure water, they are not directly applicable to corrosion rates of SNF in water chemistry that is significantly different (such as may occur in the repository). Linear kinetics adequately described the oxidation rates of metallic fuels in long-term corrosion. Temperature dependent oxidation rates were determined by linear regression analysis of the literature data. As expected the reaction rates of metallic fuels dramatically increase with temperature. The uranium metal and metal alloys have stronger temperature dependence than the aluminum dispersion fuels. The uranium metal/water reaction exhibited the highest oxidation rate of the metallic fuel types and environments that were reviewed. Consequently, the corrosion properties of all DOE SNF may be conservatively modeled as uranium metal, which is representative of spent N-Reactor fuel. The reaction rate in anoxic, saturated water vapor was essentially the same as the water reaction rate. The long-term intrinsic

A comparative analysis of in vitro toxicity of diesel exhaust particles from combustion of 1st- and 2nd-generation biodiesel fuels in relation to their physicochemical properties-the FuelHealth project.

Science.gov (United States)

Lankoff, Anna; Brzoska, Kamil; Czarnocka, Joanna; Kowalska, Magdalena; Lisowska, Halina; Mruk, Remigiusz; Øvrevik, Johan; Wegierek-Ciuk, Aneta; Zuberek, Mariusz; Kruszewski, Marcin

2017-08-01

Biodiesels represent more carbon-neutral fuels and are introduced at an increasing extent to reduce emission of greenhouse gases. However, the potential impact of different types and blend concentrations of biodiesel on the toxicity of diesel engine emissions are still relatively scarce and to some extent contradictory. The objective of the present work was to compare the toxicity of diesel exhaust particles (DEP) from combustion of two 1st-generation fuels: 7% fatty acid methyl esters (FAME; B7) and 20% FAME (B20) and a 2nd-generation 20% FAME/HVO (synthetic hydrocarbon biofuel (SHB)) fuel. Our findings indicate that particulate emissions of each type of biodiesel fuel induce cytotoxic effects in BEAS-2B and A549 cells, manifested as cell death (apoptosis or necrosis), decreased protein concentrations, intracellular ROS production, as well as increased expression of antioxidant genes and genes coding for DNA damage-response proteins. The different biodiesel blend percentages and biodiesel feedstocks led to marked differences in chemical composition of the emitted DEP. The different DEPs also displayed statistically significant differences in cytotoxicity in A549 and BEAS-2B cells, but the magnitude of these variations was limited. Overall, it seems that increasing biodiesel blend concentrations from the current 7 to 20% FAME, or substituting 1st-generation FAME biodiesel with 2nd-generation HVO biodiesel (at least below 20% blends), affects the in vitro toxicity of the emitted DEP to some extent, but the biological significance of this may be moderate.

Extended burnup with SEU fuel in Atucha-1 NPP

International Nuclear Information System (INIS)

Alvarez, L.; Casario, J.; Fink, J.; Perez, R.; Higa, M.

2002-01-01

Atucha-1 is a Pressurized Heavy Water Reactor originally fuelled with natural uranium. Fuel Assemblies consist of 36 fuel rods and the active length is 5300 mm. The total length of the fuel assembly is about 6 m. The average discharge burnup of natural UO 2 fuel is 5900 MWd/tU. After the deregulation of the Argentine electricity market there was an important incentive to reduce the impact of fuel cost on the cost of generation. To keep the competitiveness of the nuclear energy against another sources of electricity it was necessary to reduce the cost of the nuclear fuel. With this objective a program to introduce SEU (0.85 % 235 U) fuel in Atucha-1 was launched in 1993. As a result of this program the average SEU fuel discharge burnup increased to more than 11000 MWd/tU. The first SEU fuels were introduced in Atucha-1 in 1995 and, in the present stage of the program, 71% of core positions are loaded with this type of fuel. This paper describes key aspects of Atucha-1 fuel design and their relevance limiting the burnup extension and shows relevant data regarding the SEU in-reactor performance. At the present time 125 SEU Fuel Assemblies have been irradiated without failures associated with the extended burnup or unfavorable influences on the operation of the power station. (author)

Chooz B1 start-up marked by total computer control

International Nuclear Information System (INIS)

Anon.

1995-01-01

The start-up of Chooz B1, the first of EdF's latest generation 1450 MWe N4 reactors, marks the first use in a nuclear power plant of a fully computerised control room. Working in partnership with EdF, Sema Group designed and supplied the advanced command and control system for this plant. (Author)

Structural insights into the regulation and the recognition of histone marks by the SET domain of NSD1

International Nuclear Information System (INIS)

Morishita, Masayo; Di Luccio, Eric

2011-01-01

Highlights: → NSD1, NSD2/MMSET/WHSC1, and NSD3/WHSC1L1 are histone methyltransferases linked to numerous cancers. → Little is known about the NSD pathways and HMTase inhibitors are sorely needed in the epigenetic therapy of cancers. → We investigate the regulation and the recognition of histone marks by the SET domain of NSD1. → A unique and key mechanism is driven by a loop at the interface of the SET and postSET region. → Implications for developing specific and selective HMTase inhibitors are presented. -- Abstract: The development of epigenetic therapies fuels cancer hope. DNA-methylation inhibitors, histone-deacetylase and histone-methyltransferase (HMTase) inhibitors are being developed as the utilization of epigenetic targets is emerging as an effective and valuable approach to chemotherapy as well as chemoprevention of cancer. The nuclear receptor binding SET domain (NSD) protein is a family of three HMTases, NSD1, NSD2/MMSET/WHSC1, and NSD3/WHSC1L1 that are critical in maintaining the chromatin integrity. A growing number of studies have reported alterations or amplifications of NSD1, NSD2, or NSD3 in numerous carcinogenic events. Reducing NSDs activity through specific lysine-HMTase inhibitors appears promising to help suppressing cancer growth. However, little is known about the NSD pathways and our understanding of the histone lysine-HMTase mechanism is partial. To shed some light on both the recognition and the regulation of epigenetic marks by the SET domain of the NSD family, we investigate the structural mechanisms of the docking of the histone-H4 tail on the SET domain of NSD1. Our finding exposes a key regulatory and recognition mechanism driven by the flexibility of a loop at the interface of the SET and postSET region. Finally, we prospect the special value of this regulatory region for developing specific and selective NSD inhibitors for the epigenetic therapy of cancers.

Spent Fuel Management Newsletter. No. 1

International Nuclear Information System (INIS)

1990-03-01

This Newsletter has been prepared in accordance with the recommendations of the International Regular Advisory Group on Spent Fuel Management and the Agency's programme (GC XXXII/837, Table 76, item 14). The main purpose of the Newsletter is to provide Member States with new information about the state-of-the-art in one of the most important parts of the nuclear fuel cycle - Spent Fuel Management. The contents of this publication consists of two parts: (1) IAEA Secretariat contribution -work and programme of the Nuclear Materials and Fuel Cycle Technology Section of the Division of Nuclear Fuel Cycle and Waste Management, recent and planned meetings and publications, Technical Co-operation projects, Co-ordinated Research programmes, etc. (2) Country reports - national programmes on spent fuel management: current and planned storage and reprocessing capacities, spent fuel arisings, safety, transportation, storage, treatment of spent fuel, some aspects of uranium and plutonium recycling, etc. The IAEA expects to publish the Newsletter once every two years between the publications of the Regular Advisory Group on Spent Fuel Management. Figs and tabs

BRAKER1: Unsupervised RNA-Seq-Based Genome Annotation with GeneMark-ET and AUGUSTUS.

Science.gov (United States)

Hoff, Katharina J; Lange, Simone; Lomsadze, Alexandre; Borodovsky, Mark; Stanke, Mario

2016-03-01

Gene finding in eukaryotic genomes is notoriously difficult to automate. The task is to design a work flow with a minimal set of tools that would reach state-of-the-art performance across a wide range of species. GeneMark-ET is a gene prediction tool that incorporates RNA-Seq data into unsupervised training and subsequently generates ab initio gene predictions. AUGUSTUS is a gene finder that usually requires supervised training and uses information from RNA-Seq reads in the prediction step. Complementary strengths of GeneMark-ET and AUGUSTUS provided motivation for designing a new combined tool for automatic gene prediction. We present BRAKER1, a pipeline for unsupervised RNA-Seq-based genome annotation that combines the advantages of GeneMark-ET and AUGUSTUS. As input, BRAKER1 requires a genome assembly file and a file in bam-format with spliced alignments of RNA-Seq reads to the genome. First, GeneMark-ET performs iterative training and generates initial gene structures. Second, AUGUSTUS uses predicted genes for training and then integrates RNA-Seq read information into final gene predictions. In our experiments, we observed that BRAKER1 was more accurate than MAKER2 when it is using RNA-Seq as sole source for training and prediction. BRAKER1 does not require pre-trained parameters or a separate expert-prepared training step. BRAKER1 is available for download at http://bioinf.uni-greifswald.de/bioinf/braker/ and http://exon.gatech.edu/GeneMark/ katharina.hoff@uni-greifswald.de or borodovsky@gatech.edu Supplementary data are available at Bioinformatics online. © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please e-mail: journals.permissions@oup.com.

Simulation of TRIGA Mark II Benchmark Experiment using WIMSD4 and CITATION codes

International Nuclear Information System (INIS)

Dalle, Hugo Moura; Pereira, Claubia

2000-01-01

This paper presents a simulation of the TRIGA Mark II Benchmark Experiment, Part I: Steady-State Operation and is part of the calculation methodology validation developed to the neutronic calculation of the CDTN's TRIGA IPR - R1 reactor. A version of the WIMSD4, obtained in the Centro de Tecnologia Nuclear, in Cuba, was used in the cells calculation. In the core calculations was adopted the diffusion code CITATION. Was adopted a 3D representation of the core and the calculations were carried out at two energy groups. Many of the experiments were simulated, including, K eff , control rods reactivity worth, fuel elements reactivity worth distribution and the fuel temperature reactivity coefficient. The comparison of the obtained results, with the experimental results, shows differences in the range of the accuracy of the measurements, to the control rods worth and fuel temperature reactivity coefficient, or on an acceptable range, following the literature, to the K eff and fuel elements reactivity worth distribution and the fuel temperature reactivity coefficient. The comparison of the obtained results, with the experimental. results, shows differences in the range of the accuracy of the measurements, to the control rods worth and fuel temperature reactivity coefficient, or in an acceptable range, following the literature, to the K eff and fuel elements reactivity worth distribution. (author)

Minimal Marking: A Success Story

Science.gov (United States)

McNeilly, Anne

2014-01-01

The minimal-marking project conducted in Ryerson's School of Journalism throughout 2012 and early 2013 resulted in significantly higher grammar scores in two first-year classes of minimally marked university students when compared to two traditionally marked classes. The "minimal-marking" concept (Haswell, 1983), which requires…

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-07-19

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

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

International Nuclear Information System (INIS)

Montierth, Leland M.

2016-01-01

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

1st Fire Behavior and Fuels Conference: Fuels Management-How to Measure Success

Science.gov (United States)

Patricia L. Andrews

2006-01-01

The 1st Fire Behavior and Fuels Conference: Fuels Management -- How to Measure Success was held in Portland, Oregon, March 28-30, 2006. The International Association of Wildland Fire (IAWF) initiated a conference on this timely topic primarily in response to the needs of the U.S. National Interagency Fuels Coordinating Group (http://www.nifc.gov/).

Standardization of Alternative Fuels. Phase 1

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-08-15

There are different interpretations of the term 'alternative fuels', depending on the part of the world in which the definition is used. In this report, alternative fuels mainly stand for fuels that can replace gasoline and diesel oil and at the same time contribute to lowered emissions with impact on health, environment and climate. The use of alternative vehicle fuels has increased during the last 30 years. However, the increase has developed slowly and today the use is very limited, compared to the use of conventional fuels. Although, the use in some special applications, often in rather small geographical areas, can be somewhat larger. The main interest for alternative fuels has for a long time been driven by supply security issues and the possibility to reduce emissions with a negative impact on health and environment. However, the development of reformulated gasoline and low sulphur diesel oil has contributed to substantially decreased emissions from these fuels without using any alternative fuel. This has reduced the environmental impact driving force for the introduction of alternative fuels. In line with the increased interest for climate effects and the connections between these effects and the emission of greenhouse gases, and then primarily carbon dioxide, the interest for biomass based alternative fuels has increased during the 1990s. Even though one of the driving forces for alternative fuels is small today, alternative fuels are more commonly accepted than ever before. The European Commission has for example in May 2003 agreed on a directive for the promotion of the use of bio fuels. In the directive there are goals for the coming 7 years that will increase the use of alternative fuels in Europe rather dramatically, from below 1 percent now up to almost 6 percent of the total vehicle fuel consumption in 2010. The increased use of alternative fuels in Europe and the rest of the world will create a need for a common interpretation of what we

Communication impairments in mice lacking Shank1: reduced levels of ultrasonic vocalizations and scent marking behavior.

Directory of Open Access Journals (Sweden)

Markus Wöhr

Full Text Available Autism is a neurodevelopmental disorder with a strong genetic component. Core symptoms are abnormal reciprocal social interactions, qualitative impairments in communication, and repetitive and stereotyped patterns of behavior with restricted interests. Candidate genes for autism include the SHANK gene family, as mutations in SHANK2 and SHANK3 have been detected in several autistic individuals. SHANK genes code for a family of scaffolding proteins located in the postsynaptic density of excitatory synapses. To test the hypothesis that a mutation in SHANK1 contributes to the symptoms of autism, we evaluated Shank1(-/- null mutant mice for behavioral phenotypes with relevance to autism, focusing on social communication. Ultrasonic vocalizations and the deposition of scent marks appear to be two major modes of mouse communication. Our findings revealed evidence for low levels of ultrasonic vocalizations and scent marks in Shank1(-/- mice as compared to wildtype Shank1(+/+ littermate controls. Shank1(-/- pups emitted fewer vocalizations than Shank1(+/+ pups when isolated from mother and littermates. In adulthood, genotype affected scent marking behavior in the presence of female urinary pheromones. Adult Shank1(-/- males deposited fewer scent marks in proximity to female urine than Shank1(+/+ males. Call emission in response to female urinary pheromones also differed between genotypes. Shank1(+/+ mice changed their calling pattern dependent on previous female interactions, while Shank1(-/- mice were unaffected, indicating a failure of Shank1(-/- males to learn from a social experience. The reduced levels of ultrasonic vocalizations and scent marking behavior in Shank1(-/- mice are consistent with a phenotype relevant to social communication deficits in autism.

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

Energy Technology Data Exchange (ETDEWEB)

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

A study for providing additional storage spaces to ET-RR-1 spent fuel

International Nuclear Information System (INIS)

El-Kady, A.; Ashoub, N.; Saleh, H.G.

1995-01-01

The ET-RR-1 reactor spent fuel storage pool is a trapezoidal aluminum tank concrete shield and of capacity 10 m 3 . It can hold up to 60 fuel assemblies. The long operation history of the ET-RR-1 reactor resulted in a partially filled spent fuel storage with the remaining spaces not enough to host a complete load from the reactor. This work have been initiated to evaluate possible alternative solutions for providing additional storage spaces to host the available EK-10 fuel elements after irradiation and any foreseen fuel in case of reactor upgrading. Several alternate solutions have been reviewed and decision on the most suitable one is under study. These studies include criticality calculation of some suggested alternatives like reracking the present spent fuel storage pool and double tiering by the addition of a second level storage rack above the existing rack. The two levels may have different factor. Criticality calculation of the double tiering possible accident was also studied. (author)

Investigation of fuel lean reburning process in a 1.5 MW boiler

International Nuclear Information System (INIS)

Kim, Hak Young; Baek, Seung Wook; Kim, Se Won

2012-01-01

Highlights: → We examine a detailed study of fuel lean reburning process in a 1.5 MW gas-fired boiler. → Experimental and numerical researches are conducted. → We investigate change in the level of NO X and CO emission. → The recirculation flow is important in the fuel lean reburning process. -- Abstract: This paper examines a detailed study of fuel lean reburning process applied to a 1.5 MW gas-fired boiler. Experimental and numerical studies were carried out to investigate the effect of the fuel lean reburning process on the NO X reduction and CO emission. Natural gas (CH 4 ) was used as the reburn as well as the main fuel. The amount of the reburn fuel, injection location and thermal load of boiler were considered as experimental parameters. The flue gas data revealed that the fuel lean reburning process led to NO X reduction up to 43%, while CO emission was limited to less than 30 ppm for the 100% thermal load condition. The commercial computational fluid dynamics code FLUENT 6.3, which included turbulence, chemical reaction, radiation and NO modeling, was used to predict the fluid flow and heat transfer characteristics under various operational conditions in the boiler. Subsequently, predicted results were validated with available measured data such as gas temperature distributions and local mean NO X concentrations. The detailed numerical results showed that the recirculation flow developed inside the boiler was found to play an important role in improving the effectiveness of fuel lean reburning process.

Fuel cells. Pt. 1; Celle a combustibile. Pt. 1

Energy Technology Data Exchange (ETDEWEB)

Campanari, S; Macchi, E [Milan Politecnico (Italy). Dip. di Energetica

1999-01-01

Direct conversion of chemical energy into electricity (without intermediate heat generation) is a long-established method to improve the efficiency of power generation, as well as to reduce polluting emissions from thermal plants. The origins of fuel cells, as well as their operating principles, are dealt with. Then, various types of cells are taken into consideration, on the basis of both their characteristics and the operating principles of electrolytes. Finally, structure and operation of Polymer Electrolyte Membrane Fuel Cells (PEMFC), Alkaline Fuel Cells (AFC) and Phosphoric Acid Fuel Cells (PAFC) are described. [Italiano] La conversione diretta dell`energia chimica del combustibile in energia elettrica, senza passare attraverso la produzione di calore, rappresenta una via ormai ampiamente collaudata per migliorare l`efficienza della produzione di energia elettrica e per contenere le emissioni generate dagli impianti termoelettrici. L`articolo, dopo una breve presentazione della storia dello sviluppo nel tempo delle celle a combustibile, espone i principi di funzionamento delle stesse. Si esaminano quindi i vari tipi di cella a partire dalle caratteristiche e dalle modalita` di funzionamento degli elettroliti che ne definiscono la classificazione. Successivamente vengono illustrate le caratteristiche costruttive e funzionali delle celle ad elettrolita polimerico (PEMFC), delle celle alcaline (AFC) e delle celle ad acido fosforico (PAFC).

Fuel burnup analysis for the Moroccan TRIGA research reactor

International Nuclear Information System (INIS)

El Bakkari, B.; El Bardouni, T.; Nacir, B.; El Younoussi, C.; Boulaich, Y.; Boukhal, H.; Zoubair, M.

2013-01-01

Highlights: ► A fuel burnup analysis of the 2 MW TRIGA MARK II Moroccan research reactor was established. ► Burnup calculations were done by means of the in-house developed burnup code BUCAL1. ► BUCAL1 uses the MCNP tallies directly in the calculation of the isotopic inventories. ► The reactor life time was found to be 3360 MW h considering full power operating conditions. ► Power factors and fluxes of the in-core irradiation positions are strongly affected by burnup. -- Abstract: The fundamental advantage and main reason to use Monte Carlo methods for burnup calculations is the possibility to generate extremely accurate burnup dependent one group cross-sections and neutron fluxes for arbitrary core and fuel geometries. Yet, a set of values determined for a material at a given position and time remains accurate only in a local region, in which neutron spectrum and flux vary weakly — and only for a limited period of time, during which changes of the local isotopic composition are minor. This paper presents the approach of fuel burnup evaluation used at the Moroccan TRIGA MARK II research reactor. The approach is essentially based upon the utilization of BUCAL1, an in-house developed burnup code. BUCAL1 is a FORTRAN computer code designed to aid in analysis, prediction, and optimization of fuel burnup performance in nuclear reactors. The code was developed to incorporate the neutron absorption reaction tally information generated directly by MCNP5 code in the calculation of fissioned or neutron-transmuted isotopes for multi-fueled regions. The fuel cycle length and changes in several core parameters such as: core excess reactivity, control rods position, fluxes at the irradiation positions, axial and radial power factors and other parameters are estimated. Besides, this study gives valuable insight into the behavior of the reactor and will ensure better utilization and operation of the reactor during its life-time and it will allow the establishment of

Communication Impairments in Mice Lacking Shank1: Reduced Levels of Ultrasonic Vocalizations and Scent Marking Behavior

Science.gov (United States)

Wöhr, Markus; Roullet, Florence I.; Hung, Albert Y.; Sheng, Morgan; Crawley, Jacqueline N.

2011-01-01

Autism is a neurodevelopmental disorder with a strong genetic component. Core symptoms are abnormal reciprocal social interactions, qualitative impairments in communication, and repetitive and stereotyped patterns of behavior with restricted interests. Candidate genes for autism include the SHANK gene family, as mutations in SHANK2 and SHANK3 have been detected in several autistic individuals. SHANK genes code for a family of scaffolding proteins located in the postsynaptic density of excitatory synapses. To test the hypothesis that a mutation in SHANK1 contributes to the symptoms of autism, we evaluated Shank1 −/− null mutant mice for behavioral phenotypes with relevance to autism, focusing on social communication. Ultrasonic vocalizations and the deposition of scent marks appear to be two major modes of mouse communication. Our findings revealed evidence for low levels of ultrasonic vocalizations and scent marks in Shank1 −/− mice as compared to wildtype Shank1 +/+ littermate controls. Shank1 −/− pups emitted fewer vocalizations than Shank1+/+ pups when isolated from mother and littermates. In adulthood, genotype affected scent marking behavior in the presence of female urinary pheromones. Adult Shank1 −/− males deposited fewer scent marks in proximity to female urine than Shank1+/+ males. Call emission in response to female urinary pheromones also differed between genotypes. Shank1+/+ mice changed their calling pattern dependent on previous female interactions, while Shank1 −/− mice were unaffected, indicating a failure of Shank1 −/− males to learn from a social experience. The reduced levels of ultrasonic vocalizations and scent marking behavior in Shank1 −/− mice are consistent with a phenotype relevant to social communication deficits in autism. PMID:21695253

South Korea's nuclear fuel industry

International Nuclear Information System (INIS)

Clark, R.G.

1990-01-01

March 1990 marked a major milestone for South Korea's nuclear power program, as the country became self-sufficient in nuclear fuel fabrication. The reconversion line (UF 6 to UO 2 ) came into full operation at the Korea Nuclear Fuel Company's fabrication plant, as the last step in South Korea's program, initiated in the mid-1970s, to localize fuel fabrication. Thus, South Korea now has the capability to produce both CANDU and pressurized water reactor (PWR) fuel assemblies. This article covers the nuclear fuel industry in South Korea-how it is structures, its current capabilities, and its outlook for the future

High efficiency metal marking with CO2 laser and glass marking with excimer laser

DEFF Research Database (Denmark)

Bastue, Jens; Olsen, Flemmming Ove

1997-01-01

with a thoroughly tested ray-tracing model is presented and compared with experimental results. Special emphasis is put on two different applications namely marking in metal with TEA-CO2 laser and marking in glass with excimer laser. The results are evaluated on the basis of the achievable energy enhancement......Today, mask based laser materials processing and especially marking is widely used. However, the energy efficiency in such processes is very low [1].This paper gives a review of the results, that may be obtained using the energy enhancing technique [1]. Results of simulations performed...

Marked augmentation of PLGA nanoparticle-induced metabolically beneficial impact of γ-oryzanol on fuel dyshomeostasis in genetically obese-diabetic ob/ob mice.

Science.gov (United States)

Kozuka, Chisayo; Shimizu-Okabe, Chigusa; Takayama, Chitoshi; Nakano, Kaku; Morinaga, Hidetaka; Kinjo, Ayano; Fukuda, Kotaro; Kamei, Asuka; Yasuoka, Akihito; Kondo, Takashi; Abe, Keiko; Egashira, Kensuke; Masuzaki, Hiroaki

2017-11-01

Our previous works demonstrated that brown rice-specific bioactive substance, γ-oryzanol acts as a chaperone, attenuates exaggerated endoplasmic reticulum (ER) stress in brain hypothalamus and pancreatic islets, thereby ameliorating metabolic derangement in high fat diet (HFD)-induced obese diabetic mice. However, extremely low absorption efficiency from intestine of γ-oryzanol is a tough obstacle for the clinical application. Therefore, in this study, to overcome extremely low bioavailability of γ-oryzanol with super-high lipophilicity, we encapsulated γ-oryzanol in polymer poly (DL-lactide-co-glycolide) (PLGA) nanoparticles (Nano-Orz), and evaluated its metabolically beneficial impact in genetically obese-diabetic ob/ob mice, the best-known severest diabetic model in mice. To our surprise, Nano-Orz markedly ameliorated fuel metabolism with an unexpected magnitude (∼1000-fold lower dose) compared with regular γ-oryzanol. Furthermore, such a conspicuous impact was achievable by its administration once every 2 weeks. Besides the excellent impact on dysfunction of hypothalamus and pancreatic islets, Nano-Orz markedly decreased ER stress and inflammation in liver and adipose tissue. Collectively, nanotechnology-based developments of functional foods oriented toward γ-oryzanol shed light on the novel approach for the treatment of a variety of metabolic diseases in humans.

Irradiation behavior of uranium-molybdenum dispersion fuel: Fuel performance data from RERTR-1 and RERTR-2

International Nuclear Information System (INIS)

Meyer, M.K.; Clark, C.R.; Hayes, S.L.; Strain, R.V.; Hofman, G.L.; Snelgrove, J.L.; Park, J.M.; Kim, K.H.

1999-01-01

This paper presents quantitative data on the irradiation behavior of uranium-molybdenum fuels from the low temperature RERTR-1 and -2 experiments. Fuel swelling measurements of U-Mo fuels at ∼40% and ∼70% burnup are presented. The rate of fuel-matrix interaction layer growth is estimated. Microstructures of fuel in the pre- and postirradiation condition were compared. Based on these data, a qualitative picture of the evolution of the U-Mo fuel microstructure during irradiation has been developed. Estimates of uranium-molybdenum fuel swelling and fuel-matrix interaction under high-power research reactor operating conditions are presented. (author)

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

International Nuclear Information System (INIS)

Edelmann, M.

1995-12-01

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

Equilibrium fuel-management simulations for 1.2% SEU in a CANDU 6

International Nuclear Information System (INIS)

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

1989-06-01

Fuel-management simulations have been performed for 1.2% SEU in a CANDU 6 reactor at equilibrium, for three fuel-management options: axial shuffling; a regular 2-bundling shift with the adjuster rods removed from the core; and a regular 2-bundle shift with the adjuster rods present. Both time-average and time-dependent simulations were performed, from which the physics characteristics of the cores at equilibrium were estimated. Power and power-boost envelopes were derived for both 37-element fuel, and the advanced CANFLEX bundle

Probabilistic Safety Assessment Of It TRIGA Mark-II Reactor

International Nuclear Information System (INIS)

Ergun, E; Kadiroglu, O.S.

1999-01-01

The probabilistic safety assessment for Istanbul Technical University (ITU) TRIGA Mark-II reactor is performed. Qualitative analysis, which includes fault and event trees and quantitative analysis which includes the collection of data for basic events, determination of minimal cut sets, calculation of quantitative values of top events, sensitivity analysis and importance measures, uncertainty analysis and radiation release from fuel elements are considered

Design and construction of the SIPPING for fuels of the TRIGA Mark III reactor

International Nuclear Information System (INIS)

Castaneda J, G.; Delfin L, A.; Alvarado P, R.; Mazon R, R.; Ortega V, B.

2003-01-01

The sipping technique, it has been used by several possessors of nuclear research reactors in its irradiated nuclear fuels, likewise in some fuel storage sites, with the objective of to determine the quantity of radioactivity that the fuel liberates in the means in that it is. The irradiated fuel in storage of some nuclear research reactors, its can have cracks that cross the cladding of the same one, generating the liberation of fission products that its need to determine to maintain safety measures appropriate as much as the fuel as of the facilities where they are. It doesn't exist until now, some method published for the non destructive sipping test technique. Based on that described, the Reactor Department of the National Institute of Nuclear Research, it has designed and built an inspection system of irradiated fuel that it will allow the detection of gassy fission products in site, and solids by means of the measurement of the activity of the Cs-137 contained in water samples. (Author)

Nonlinear analysis and evaluation of a reinforced concrete spent fuel storage pool for accidental thermal loads

International Nuclear Information System (INIS)

Kabir, A.F.; Bolourchi, S.

1991-01-01

A feasibility study was conducted for addition of consolidated fuel racks to an existing reinforced concrete spent fuel storage pool of a Mark I BWR plant. Nonlinear analysis of a detailed three-dimensional model of the fuel pool, considering cracking in concrete under gravity and thermal load conditions, showed that the pool has reserve capacities to carry the additional loads. (author)

27 CFR 28.103 - Export marks.

Science.gov (United States)

2010-04-01

... 27 Alcohol, Tobacco Products and Firearms 1 2010-04-01 2010-04-01 false Export marks. 28.103... Manufacturing Bonded Warehouse § 28.103 Export marks. (a) General. In addition to the marks and brands required... provisions of part 19 of this chapter, the proprietor shall mark the word “Export” on the Government side of...

27 CFR 28.144 - Export marks.

Science.gov (United States)

2010-04-01

... 27 Alcohol, Tobacco Products and Firearms 1 2010-04-01 2010-04-01 false Export marks. 28.144... § 28.144 Export marks. (a) General Requirement. In addition to the marks and brands required to be... brewer shall mark the word “Export” on each container or case of beer, or the words “Beer concentrate for...

27 CFR 28.154 - Export marks.

Science.gov (United States)

2010-04-01

... 27 Alcohol, Tobacco Products and Firearms 1 2010-04-01 2010-04-01 false Export marks. 28.154..., for Exportation or Transfer to a Foreign-Trade Zone § 28.154 Export marks. In addition to the marks... provisions of part 19 of this chapter, the proprietor shall mark the word “Export” on the Government side of...

Fuel System Compatibility Issues for Prometheus-1

International Nuclear Information System (INIS)

DC-- Noe; KB Gibbard; MH Krohn

2006-01-01

Compatibility issues for the Prometheus-1 fuel system have been reviewed based upon the selection of UO 2 as the reference fuel material. In particular, the potential for limiting effects due to fuel- or fission product-component (cladding, liner, spring, etc) chemical interactions and clad-liner interactions have been evaluated. For UO 2 -based fuels, fuel-component interactions are not expected to significantly limit performance. However, based upon the selection of component materials, there is a potential for degradation due to fission products. In particular, a chemical liner may be necessary for niobium, tantalum, zirconium, or silicon carbide-based systems. Multiple choices exist for the configuration of a chemical liner within the cladding; there is no clear solution that eliminates all concerns over the mechanical performance of a clad/liner system. A series of tests to evaluate the performance of candidate materials in contact with real and simulated fission products is outlined

Fuel bundle to pressure tube fretting in Bruce and Darlington

Energy Technology Data Exchange (ETDEWEB)

Norsworthy, A G; Ditschun, A [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1996-12-31

As the fuel channel elongates due to creep, the fuel string moves relative to the inlet until the fuel pads at the inboard end eventually separate from the spacer sleeve, and the fuel resides on the burnish mark of the pressure tube. The bundle is then supported in a fashion which contributes to increased levels of vibration. Those pads which (due to geometric variation) have contact loads with the pressure tube within a certain range, vibrate, and cause significant fretting on the burnish mark, and further along at the midplane of the bundle. Inspection of the pressure tubes in Bruce A, Bruce B, and Darlington has revealed fret damage up to 0.55 mm at the burnish mark and slightly lower than this at the inlet bundle midplane. To date, all fret marks have been dealt with successfully without the need for tube replacement, but a program of work has been initiated to understand the mechanism and reduce the fretting. Such understanding is necessary to guide future design changes to the fuel bundle, to guide future inspection programs, to guide maintenance programs, and for longer term strategic planning. This paper discusses how the understanding of fretting has evolved and outlines a current hypothesis for the mechanism of fretting. The role of bundle geometry, excitation forces, and reactor conditions are reviewed, along with options under consideration to mitigate damage. (author). 4 refs., 2 tabs., 13 figs.

Fuel bundle to pressure tube fretting in Bruce and Darlington

International Nuclear Information System (INIS)

Norsworthy, A.G.; Ditschun, A.

1995-01-01

As the fuel channel elongates due to creep, the fuel string moves relative to the inlet until the fuel pads at the inboard end eventually separate from the spacer sleeve, and the fuel resides on the burnish mark of the pressure tube. The bundle is then supported in a fashion which contributes to increased levels of vibration. Those pads which (due to geometric variation) have contact loads with the pressure tube within a certain range, vibrate, and cause significant fretting on the burnish mark, and further along at the midplane of the bundle. Inspection of the pressure tubes in Bruce A, Bruce B, and Darlington has revealed fret damage up to 0.55 mm at the burnish mark and slightly lower than this at the inlet bundle midplane. To date, all fret marks have been dealt with successfully without the need for tube replacement, but a program of work has been initiated to understand the mechanism and reduce the fretting. Such understanding is necessary to guide future design changes to the fuel bundle, to guide future inspection programs, to guide maintenance programs, and for longer term strategic planning. This paper discusses how the understanding of fretting has evolved and outlines a current hypothesis for the mechanism of fretting. The role of bundle geometry, excitation forces, and reactor conditions are reviewed, along with options under consideration to mitigate damage. (author). 4 refs., 2 tabs., 13 figs

Treatment for dismantled radioactive solid waste from the TRIGA Mark-2 and 3

International Nuclear Information System (INIS)

Park, Seung Kook; Jung, Kyung Hwan

1999-06-01

Radioactive wastes are generally classified into 3 type depending on their physical property: liquid, solid and gaseous type. State-of -the art concerning liquid waste treatment has already been published; KAERI/TR-1315/99. Solid wastes classification package and treatment method will be studied to effectively manage them during the practical decommissioning work. All of the spent fuel produced during the operation of the TRIGA Mark-2 and 3 have been transported to the US last year, 1998, according to the spent fuel management strategy set-up by the US government for the non-proliferation of nuclear energy. Solid wastes are mainly all equipment existing inside of the reactors, activated concrete among the bio-shielded concrete, pipes, pimps, resin filter and it's housings, heat-exchangers, liquid waste storage tanks, to radioactive waste storage treatment facilities and so on. Solid wastes are generally low-level. They are classified according to the national regulation and nuclear law and IAEA Safety Standard Series ST-1(1996). Medium level radioactive wastes from reactor structures, mainly stainless steel component from the Rotary Specimen Rack(RSR) will be properly dismantled and stored in a shield container such as TIF(TRIGA Irradiated Fuel) container. While, low-level solid waste will be treated and packed in a ISO container(4m 3 ISO container for example) according to the IAEA recommendation. And combustible solid waste such as cloths, gloves, paper etc. will be packed in a 200 liters drum. This state-of-the art shows a general feature of the solid radioactive waste management which will be produced during the decommissioning of the TRIGA Mark-2 and 3 research reactors. (author). 17 refs., 17 tabs., 2 figs

The probability of Mark-1 liner failure

International Nuclear Information System (INIS)

Theofanous, T.G.; Yan, H.; Ratnam, U.; Amarasooriya, W.H.

1991-01-01

The authors are proposing a probabilistic methodology, the risk-oriented accident analysis methodology (ROAAM) as an overall systematic, disciplined approach for addressing the Mark-1 liner attack issue. The probabilistic framework encompasses the key features of the phenomenology, yet it is flexible enough to allow independent quantification of individual components as it may arise from independent research efforts. As a first step in this direction, the authors assembled, discussed, and took into consideration in the quantification proposed all relevant prior work. Furthermore, as an essential aspect of the overall methodology, most of those whose work has been referenced and/or used in this report have been asked to comment. The details of this work, the comments received, and the authors' responses are included in NUREG/CR-5423. As an even more important characteristic of the methodology, it is hoped that other quantifications (or information relevant to such) of independent components will become available in the future so that one can aim for convergence and closure

Fuel oil and kerosene sales 1992

International Nuclear Information System (INIS)

1993-01-01

This publication contains the 1992 survey results of the ''Annual Fuel Oil and Kerosene Sales Report'' (Form EIA-821). This is the fourth year that the survey data have appeared in a separate publication. Prior to the 1989 report, the statistics appeared in the Petroleum Marketing Annual (PMA) for reference year 1988 and the Petroleum Marketing Monthly (PMM for reference years 1984 through 1987. The 1992 edition marks the ninth annual presentation of the results of the ongoing ''Annual Fuel Oil and Kerosene Sales Report'' survey. Except for the kerosene and on-highway diesel information, data presented in Tables 1 through 12 (Sales of Fuel Oil and Kerosene) present results of the EIA-821 survey. Tables 13 through 24 (Adjusted Sales of Fuel Oil and Kerosene) include volumes that are based on the EIA-821 survey but have been adjusted to equal the products supplied volumes published in the Petroleum Supply Annual (PSA)

Development and use of GREET 1.6 fuel-cycle model for transportation fuels and vehicle technologies

International Nuclear Information System (INIS)

Wang, M. Q.

2001-01-01

Since 1995, with funds from the U.S. Department of Energy's (DOE's) Office of Transportation Technologies (OTT), Argonne National Laboratory has been developing the Greenhouse gases, Regulated Emissions, and Energy use in Transportation (GREET) model. The model is intended to serve as an analytical tool for use by researchers and practitioners in estimating fuel-cycle energy use and emissions associated with alternative transportation fuels and advanced vehicle technologies. Argonne released the first version of the GREET model--GREET 1.0--in June 1996. Since then, it has released a series of GREET versions with revisions, updates, and upgrades. In February 2000, the latest public version of the model--GREET 1.5a--was posted on Argonne's Transportation Technology Research and Development Center (TTRDC) Web site (www.transportation.anl.gov/ttrdc/greet). Major publications that address GREET development are listed. These reports document methodologies, development, key default assumptions, applications, and results of the GREET model. They are also posted, along with additional materials for the GREET model, on the TTRDC Web site. For a given transportation fuel/technology combination, the GREET model separately calculates: (A)--Fuel-cycle energy consumption for the following three source categories: (1) Total energy (all energy sources), (2) Fossil fuels (petroleum, natural gas [NG], and coal), and (3) Petroleum. (B)--Fuel-cycle emissions of the following three greenhouse gases (GHGs): (1) Carbon dioxide (CO 2 ) (with a global warming potential [GWP] of 1), (2) Methane (CH 4 ) (with a GWP of 21), and (3) Nitrous oxide (N 2 O) (with a GWP of 310). (C)--Fuel-cycle emissions of the following five criteria pollutants (separated into total [T] and urban [U] emissions): (1) Volatile organic compounds (VOCs), (2) Carbon monoxide (CO), (3) Nitrogen oxides (NO x ), (4) Particulate matter with a mean aerodynamic diameter of 10 (micro)m or less (PM 10 ), and (5) Sulfur oxides

Predicted HIFAR fuel element temperatures for postulated loss-of-coolant accidents

International Nuclear Information System (INIS)

Green, W.J.

1987-04-01

A two-dimensional theoretical heat transfer model of a HIFAR Mark IV/Va fuel element has been developed and validated by comparing predicted thermal performances with experimental temperature responses obtained from irradiated fuel elements during simulated accident conditions. Full details of the model's development and its verification have been reported elsewhere. In this report, the model has been further used to ascertain acceptable limits of fuel element decay power for the start of two specific LOCAs which have been identified by the Regulatory Bureau of the AAEC. For a single fuel element which is positioned within a fuel load/unload flask and is not subjected to any forced convective air cooling, the model indicates that fission product decay powers must not exceed 1.86 kW if fuel surface temperatures are not to exceed 450 deg C. In the case of a HIFAR core LOCA in which the complete inventory of heavy water is lost, it is calculated that the maximum fission product decay power of a central element must not exceed 1.1 kW if fuel surface temperatures are not to exceed 450 deg C anywhere in the core

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

International Nuclear Information System (INIS)

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

1977-01-01

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

World nuclear fuel cycle

International Nuclear Information System (INIS)

Anon.

1979-01-01

A coloured pull-out wall chart is presented showing the fuel cycle interests of the world. Place names are marked and symbols are used to indicate regions associated with uranium or thorium deposits, mining, milling, enrichment, reprocessing and fabrication. (UK)

Applicable regulations and development of surveillance experiments of criticality approach in the TRIGA III Mark reactor

International Nuclear Information System (INIS)

Gonzalez M, J.L.; Aguilar H, F.; Rivero G, T.; Sainz M, E.

2000-01-01

In the procedure elaborated to repair the vessel of TRIGA III Mark reactor is required to move toward two tanks of temporal storage the fuel elements which are in operation and the spent fuel elements which are in decay inside the reactor pool. The National Commission of Nuclear Safety and Safeguards (CNSNS) has requested as protection measure that it is carried out a surveillance of the criticality approach of the temporal storages. This work determines the main regulation aspects that entails an experiment of criticality approach, moreover, informing about the results obtained in the developing of this experiments. The regulation aspects are not exclusives for this work in the TRIGA Mark III reactor but they also apply toward any assembling of fissile material. (Author)

Development of fuel number reader by ultrasonic imaging techniques

International Nuclear Information System (INIS)

Omote, T.; Yoshida, T.

1991-01-01

This paper reports on a spent fuel ID number reader using ultrasonic imaging techniques that has been developed to realize efficient and automatic verification of fuel numbers, thereby to reduce mental load and radiation exposure for operators engaged in the verification task. The ultrasonic imaging techniques for automatic fuel number recognition are described. High-speed and high reliability imaging of the spent fuel ID number are obtained by using linear array type ultrasonic probe. The ultrasonic wave is scanned by switching array probe in vertical direction, and scanned mechanically in horizontal direction. Time for imaging of spent fuel ID number on assembly was confirmed less than three seconds by these techniques. And it can recognize spent fuel ID number even if spent fuel ID number can not be visualized by an optical method because of depositing fuel number regions by soft card. In order to recognize spent fuel ID number more rapidly and more reliably, coded fuel number expressed by plural separate recesses form is developed. Every coded fuel number consists of six small holes (about 1 mm dia.) and can be marked adjacent to the existing fuel number expressed by letters and numbers

27 CFR 28.123 - Export marks.

Science.gov (United States)

2010-04-01

... 27 Alcohol, Tobacco Products and Firearms 1 2010-04-01 2010-04-01 false Export marks. 28.123..., or Transportation to a Manufacturing Bonded Warehouse § 28.123 Export marks. (a) General. In addition... filled under the provisions of part 24 of this chapter, the proprietor shall mark the word “Export” on...

Irradiation experience of IPEN fuel at IEA-R1 research reactor

International Nuclear Information System (INIS)

Perrotta, Jose A.; Neto, Adolfo; Durazzo, Michelangelo; Souza, Jose A.B. de; Frajndlich, Roberto

1998-01-01

IPEN/CNEN-SP produces, for its IEA-R1 Research Reactor, MTR fuel assemblies based on U 3 O 8 -Al dispersion fuel type. Since 1985 a qualification program on these fuel assemblies has been performed. Average 235 U burnup of 30% and peak burnup of 50% was already achieved by these fuel assemblies. This paper presents some results acquire, by these fuel assemblies, under irradiation at IEA-R1 Research Reactor. (author)

In-reactor cladding breach of EBR-II driver-fuel elements

International Nuclear Information System (INIS)

Seidel, B.R.; Einziger, R.E.

1977-01-01

Knowledge of performance and minimum useful element lifetime of Mark-II driver-fuel elements is required to maintain a high plant operating capacity factor with maximum fuel utilization. To obtain such knowledge, intentional cladding breach has been obtained in four run-to-cladding-breach Mark-II experimental driver-fuel subassemblies operating under normal conditions in EBR-II. Breach and subsequent fission-product release proved benign to reactor operations. The breaches originated on the outer surface of the cladding in the root of the restrainer dimples and were intergranular. The Weibull distribution of lifetime accurately predicts the observed minimum useful element lifetime of 10 at.% burnup, with breach ensuing shortly thereafter

Mark II containment 1/6-scale pressure suppression test program: data report no. 2

International Nuclear Information System (INIS)

Kukita, Yutaka; Okazaki, Motoaki; Namatame, Ken; Shiba, Masayoshi

1979-08-01

This report documents experimental data from the first test phase of the Mark II Containment 1/6-Scale Pressure Suppression Test. The 1/6-Scale Test was initiated in December, 1976, to investigate the thermohydraulic responses of a BWR Mark II pressure suppression system to a postulated loss-of-coolant accident (LOCA), by means of scale model experiments. From January to June, 1977, a series of tests were performed for the Japanese BWR Owners' Group. These tests consisted of eight air-blowdown pool swell tests, three steam-blowdown pool swell tests, and twelve steam condensation tests. The dynamic responses of pressure and pool water level during the blowdown, pressure oscillation and chugging phenomena associated with unsteady condensation of steam were measured. (author)

Mark 4A project training evaluation

Science.gov (United States)

Stephenson, S. N.

1985-01-01

A participant evaluation of a Deep Space Network (DSN) is described. The Mark IVA project is an implementation to upgrade the tracking and data acquisition systems of the dSN. Approximately six hundred DSN operations and engineering maintenance personnel were surveyed. The survey obtained a convenience sample including trained people within the population in order to learn what training had taken place and to what effect. The survey questionnaire used modifications of standard rating scales to evaluate over one hundred items in four training dimensions. The scope of the evaluation included Mark IVA vendor training, a systems familiarization training seminar, engineering training classes, a on-the-job training. Measures of central tendency were made from participant rating responses. Chi square tests of statistical significance were performed on the data. The evaluation results indicated that the effects of different Mark INA training methods could be measured according to certain ratings of technical training effectiveness, and that the Mark IVA technical training has exhibited positive effects on the abilities of DSN personnel to operate and maintain new Mark IVA equipment systems.

Mark 4A project training evaluation

Science.gov (United States)

Stephenson, S. N.

1985-11-01

A participant evaluation of a Deep Space Network (DSN) is described. The Mark IVA project is an implementation to upgrade the tracking and data acquisition systems of the dSN. Approximately six hundred DSN operations and engineering maintenance personnel were surveyed. The survey obtained a convenience sample including trained people within the population in order to learn what training had taken place and to what effect. The survey questionnaire used modifications of standard rating scales to evaluate over one hundred items in four training dimensions. The scope of the evaluation included Mark IVA vendor training, a systems familiarization training seminar, engineering training classes, a on-the-job training. Measures of central tendency were made from participant rating responses. Chi square tests of statistical significance were performed on the data. The evaluation results indicated that the effects of different Mark INA training methods could be measured according to certain ratings of technical training effectiveness, and that the Mark IVA technical training has exhibited positive effects on the abilities of DSN personnel to operate and maintain new Mark IVA equipment systems.

Standalone BISON Fuel Performance Results for Watts Bar Unit 1, Cycles 1-3

Energy Technology Data Exchange (ETDEWEB)

Clarno, Kevin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pawlowski, Roger [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Stimpson, Shane [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-03-07

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is moving forward with more complex multiphysics simulations and increased focus on incorporating fuel performance analysis methods. The coupled neutronics/thermal-hydraulics capabilities within the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) have become relatively stable, and major advances have been made in analysis efforts, including the simulation of twelve cycles of Watts Bar Nuclear Unit 1 (WBN1) operation. While this is a major achievement, the VERA-CS approaches for treating fuel pin heat transfer have well-known limitations that could be eliminated through better integration with the BISON fuel performance code. Several approaches are being implemented to consider fuel performance, including a more direct multiway coupling with Tiamat, as well as a more loosely coupled one-way approach with standalone BISON cases. Fuel performance typically undergoes an independent analysis using a standalone fuel performance code with manually specified input defined from an independent core simulator solution or set of assumptions. This report summarizes the improvements made since the initial milestone to execute BISON from VERA-CS output. Many of these improvements were prompted through tighter collaboration with the BISON development team at Idaho National Laboratory (INL). A brief description of WBN1 and some of the VERA-CS data used to simulate it are presented. Data from a small mesh sensitivity study are shown, which helps justify the mesh parameters used in this work. The multi-cycle results are presented, followed by the results for the first three cycles of WBN1 operation, particularly the parameters of interest to pellet-clad interaction (PCI) screening (fuel-clad gap closure, maximum centerline fuel temperature, maximum/minimum clad hoop stress, and cumulative damage index). Once the mechanics of this capability are functioning, future work will target cycles with

Hybrid laser arc welding of a used fuel container

Energy Technology Data Exchange (ETDEWEB)

Boyle, C., E-mail: cboyle@nwmo.ca [Nuclear Waste Management Organization, Toronto, ON (Canada); Martel, P. [Novika Solutions, La Pocatiere, QC (Canada)

2015-07-01

The Nuclear Waste Management Organization (NWMO) has designed a novel Used Fuel Container (UFC) optimized for CANDU used nuclear fuel. The Mark II container is constructed of nuclear grade pipe for the body and capped with hemi-spherical heads. The head-to-shell joint fit-up features an integral backing designed for external pressure, eliminating the need for a full penetration closure weld. The NWMO and Novika Solutions have developed a partial penetration, single pass Hybrid Laser Arc Weld (HLAW) closure welding process requiring no post-weld heat treatment. This paper will discuss the joint design, HLAW process, associated welding equipment, and prototype container fabrication. (author)

Hybrid laser arc welding of a used fuel container

Energy Technology Data Exchange (ETDEWEB)

Boyle, C. [Nuclear Waste Management Organization (NWMO), Toronto, Ontario (Canada); Martel, P. [Novika Solutions, La Pocatiere, Quebec (Canada)

2015-09-15

The Nuclear Waste Management Organization (NWMO) has designed a novel Used Fuel Container (UFC) optimized for CANDU used nuclear fuel. The Mark II container is constructed of nuclear grade pipe for the body and capped with hemi-spherical heads. The head-to-shell joint fit-up features an integral backing designed for external pressure, eliminating the need for a full penetration closure weld. The NWMO and Novika Solutions have developed a partial penetration, single pass Hybrid Laser Axe Weld (HLAW) closure welding process requiring no post-weld heat treatment. This paper will discuss the joint design, HLAW process, associated welding equipment, and prototype container fabrication. (author)

DNAM-1 Expression Marks an Alternative Program of NK Cell Maturation

Directory of Open Access Journals (Sweden)

Ludovic Martinet

2015-04-01

Full Text Available Natural killer (NK cells comprise a heterogeneous population of cells important for pathogen defense and cancer surveillance. However, the functional significance of this diversity is not fully understood. Here, we demonstrate through transcriptional profiling and functional studies that the activating receptor DNAM-1 (CD226 identifies two distinct NK cell functional subsets: DNAM-1+ and DNAM-1− NK cells. DNAM-1+ NK cells produce high levels of inflammatory cytokines, have enhanced interleukin 15 signaling, and proliferate vigorously. By contrast, DNAM-1− NK cells that differentiate from DNAM-1+ NK cells have greater expression of NK-cell-receptor-related genes and are higher producers of MIP1 chemokines. Collectively, our data reveal the existence of a functional program of NK cell maturation marked by DNAM-1 expression.

27 CFR 28.193 - Export marks.

Science.gov (United States)

2010-04-01

... 27 Alcohol, Tobacco Products and Firearms 1 2010-04-01 2010-04-01 false Export marks. 28.193... Drawback Filing of Notice and Removal § 28.193 Export marks. In addition to the marks and brands required... chapter, the exporter shall mark the word “Export” on the Government side of each case or Government head...

Super Phenix 1 fuel cycle, technical and economical outlooks

International Nuclear Information System (INIS)

Mougniot, J.C.; Baumier, J.; Duchatelle, L.

1982-01-01

An analysis of the costs of the various parts of the Super Phenix 1 fuel cycle is presented. The basis for calculating the mean levelized present unit cost used in French economic analyses is described. A description of the fuel cycle which follows includes the physical characteristics and management of the fuel and the costs of fuel services and raw materials. The results of calculations about Super Phenix mean levelized present fuel cycle unit cost are indicated and a comparison with two, four and six 1500 MWe units and PWR units is made. Finally conclusions are drawn about the economic possibility of FBR deployment. (U.K.)

Neutron Characterization of Encapsulated ATF-1/LANL-1 Mockup Fuel Capsules

Energy Technology Data Exchange (ETDEWEB)

Vogel, Sven C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Borges, Nicholas Paul [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Losko, Adrian Simon [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mosby, Shea Morgan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Voit, Stewart Lancaster [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); White, Joshua Taylor [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Byler, Darrin David [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dunwoody, John Tyler [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Nelson, Andrew Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mcclellan, Kenneth James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-09-28

Twenty pellets of mock-up accident tolerant fuels UN-U3Si5 were produced at LANL and loaded in two rodlet/capsule assemblies. Tomographic imaging and diffraction measurements were performed to characterize these samples at the Flight-Path 5 and HIPPO beam lines at LANSCE/LANL between November 2016 and January 2017 as well as in August 2017. The entire ~10 cm long, ~1 cm diameter fuel volume could be characterized, however due to time constraints only 2 mm slices in 4mm increments were characterized with neutron diffraction and a 28mm subset of the entire sample was characterized with energy-resolved neutron imaging. The double encapsulation of the fuel into two steel containers does not pose a problem for the neutron analysis and the methods could be applied to enriched as well irradiated fuels.

Changing Context of Trade Mark Protection in India: A Review of the Trade Marks Act, 1999

OpenAIRE

Pathak, Akhileshwar

2004-01-01

With liberalisation and globalisation of the Indian economy, it has become possible for anyone to get into production and services in most of the sectors. This has led to rampant misuse and appropriation of trade marks. In an insulated economy, with monopoly markets, law protecting trade marks had a limited role. In the changed context, however, trade mark law will be a field of much interest for academics and practitioners. Towards this, the paper explores the formation of trade mark law in ...

Fuel quality processing study, volume 1

Science.gov (United States)

Ohara, J. B.; Bela, A.; Jentz, N. E.; Syverson, H. T.; Klumpe, H. W.; Kessler, R. E.; Kotzot, H. T.; Loran, B. L.

1981-01-01

A fuel quality processing study to provide a data base for an intelligent tradeoff between advanced turbine technology and liquid fuel quality, and also, to guide the development of specifications of future synthetic fuels anticipated for use in the time period 1985 to 2000 is given. Four technical performance tests are discussed: on-site pretreating, existing refineries to upgrade fuels, new refineries to upgrade fuels, and data evaluation. The base case refinery is a modern Midwest refinery processing 200,000 BPD of a 60/40 domestic/import petroleum crude mix. The synthetic crudes used for upgrading to marketable products and turbine fuel are shale oil and coal liquids. Of these syncrudes, 50,000 BPD are processed in the existing petroleum refinery, requiring additional process units and reducing petroleum feed, and in a new refinery designed for processing each syncrude to produce gasoline, distillate fuels, resid fuels, and turbine fuel, JPGs and coke. An extensive collection of synfuel properties and upgrading data was prepared for the application of a linear program model to investigate the most economical production slate meeting petroleum product specifications and turbine fuels of various quality grades. Technical and economic projections were developed for 36 scenarios, based on 4 different crude feeds to either modified existing or new refineries operated in 2 different modes to produce 7 differing grades of turbine fuels. A required product selling price of turbine fuel for each processing route was calculated. Procedures and projected economics were developed for on-site treatment of turbine fuel to meet limitations of impurities and emission of pollutants.

Reducing fuel subsidies and the implication on fiscal balance and poverty in Indonesia: A simulation analysis

International Nuclear Information System (INIS)

Dartanto, Teguh

2013-01-01

There is an urgent need for phasing out fuel subsidies in Indonesia due to a severe budget deficit and a worsening of income distribution. Fuel subsidies, of which almost 72% are enjoyed by the 30% of the richest income groups, have consumed on average 63.8% of the total subsidies between 1998 and 2013. This paper aims at evaluating the relationship between existing fuel subsidies and fiscal balance and at analysing the poverty impact of phasing out fuel subsidies. Applying a CGE-microsimulation, this study found that removing 25% of fuel subsidies increases the incidence of poverty by 0.259 percentage points. If this money were fully allocated to government spending, the poverty incidence would decrease by 0.27 percentage points. Moreover, the 100% removal of fuel subsidies and the reallocation of 50% of them to government spending, transfers and other subsidies could decrease the incidence of poverty by 0.277 percentage points. However, these reallocation policies might not be effective in compensating for the adverse impacts of the 100% removal of fuel subsidies if economic agents try to seek profit through mark-up pricing over the increase of production costs. - Highlights: ► Massive fuel subsidies reduce fiscal spaces used to alleviate poverty in Indonesia. ► Indonesia can avoid a budget deficit by 78% cutting of fuel subsidies. ► A CGE-microsimulation is applied to analyse the impacts of fuel subsidy reallocation. ► The 50% of reallocation fuel subsidies decreases the poverty by 0.277 percentage points. ► Mark-up pricing done by economic agents reduces the effectiveness of reallocation

Development and validation of a model TRIGA Mark III reactor with code MCNP5

International Nuclear Information System (INIS)

Galicia A, J.; Francois L, J. L.; Aguilar H, F.

2015-09-01

The main purpose of this paper is to obtain a model of the reactor core TRIGA Mark III that accurately represents the real operating conditions to 1 M Wth, using the Monte Carlo code MCNP5. To provide a more detailed analysis, different models of the reactor core were realized by simulating the control rods extracted and inserted in conditions in cold (293 K) also including an analysis for shutdown margin, so that satisfied the Operation Technical Specifications. The position they must have the control rods to reach a power equal to 1 M Wth, were obtained from practice entitled Operation in Manual Mode performed at Instituto Nacional de Investigaciones Nucleares (ININ). Later, the behavior of the K eff was analyzed considering different temperatures in the fuel elements, achieving calculate subsequently the values that best represent the actual reactor operation. Finally, the calculations in the developed model for to obtain the distribution of average flow of thermal, epithermal and fast neutrons in the six new experimental facilities are presented. (Author)

Reactor physics tests of TRIGA Mark-II Reactor in Ljubljana

International Nuclear Information System (INIS)

Ravnik, M.; Mele, I.; Trkov, A.; Rant, J.; Glumac, B.; Dimic, V.

2008-01-01

TRIGA Mark-II Reactor in Ljubljana was recently reconstructed. The reconstruction consisted mainly of replacing the grid plates, the control rod mechanisms and the control unit. The standard type control rods were replaced by the fuelled follower type, the central grid location (A ring) was adapted for fuel element insertion, the triangular cutouts were introduced in the upper plate design. However, the main novelty in reactor physics and operational features of the reactor was the installation of a pulse rod. Having no previous operational experience in pulsing, a detailed and systematic sequence of tests was defined in order to check the predicted design parameters of the reactor with measurements. The following experiments are treated in this paper: initial criticality, excess reactivity measurements, control rod worth measurement, fuel temperature distribution, fuel temperature reactivity coefficient, pulse parameters measurement (peak power, prompt energy, peak temperature). Flux distributions in steady state and pulse mode were measured as well, however, they are treated only briefly due to the volume of the results. The experiments were performed with completely fresh fuel of 12 w% enriched Standard Stainless Steel type. The core configuration was uniform (one fuel element type, including fuelled followers) and compact (no irradiation channels or gaps), as such being particularly convenient for testing the computer codes for TRIGA reactor calculations. Comparison of analytical predictions, obtained with WIMS, SLXTUS, TRIGAP and PULSTRI codes to measured values showed agreement within the error of the measurement and calculation. The paper has the following contents: 1. Introduction; 2. Steady State Experiments; 2.1. Core loading and critical experiment; 2.2. Flux range determination for tests at zero power; 2.3. Digital reactivity meter checkout; 2.4. Control rod worth measurements; 2.5. Excess reactivity measurement; 2.6. Thermal power calibration; 2

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

International Nuclear Information System (INIS)

Majercik, J.

1998-01-01

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

IEA-R1 reactor - Spent fuel management

International Nuclear Information System (INIS)

Mattos, J.R.L. De

1996-01-01

Brazil currently has one Swimming Pool Research Reactor (IEA-R1) at the Instituto de Pesquisas Energeticas e Nucleares - Sao Paulo. The spent fuel produced is stored both at the Reactor Pool Storage Compartment and at the Dry Well System. The present situation and future plans for spent fuel storage are described. (author). 3 refs, 2 figs, 2 tabs

Visual inspection system and sipping design for spent fuel at TRIGA MARK III reactor of Mexico

International Nuclear Information System (INIS)

Delfin, A.; Mazon, R.

2002-01-01

In the framework of the Technical Cooperation Regional Project for Latin America RLA/4/018 for the biennium 2001-2002, one of the activities identified is the characterization of spent fuel. Of these activities an important one is not doubt the physical condition of spent fuel because an appropriate identification of the fuel status will prevent problems of fuel leaks, corrosion problems etc. As part of the activities of the project was decided that countries no having visual inspection and sipping systems should be very desirable to have them as a result of this project. The Triga reactor of Mexico does not have both of them, therefore, it was decided the need of having both system. The paper describe first the way we designed and constructed a remote Visual Inspection System and example of how is operated. Along the experience and problems we have had with the system. Also we will present the design of the Sipping system were two option were considered. First to take a sample of water after a convenient period of time passing through a circuit to a multichannel analyzer and to identify leakage by way of measuring Caesium-137. Second, exists the possibility that the Stainless Steel sleeve of the fuel has only very small failures, so it is going to be very difficult to have leakages unless the fuel is hot. Therefore we are evaluating the possibility of using heaters to increase the temperature of the fuel and succeed on detecting leakages. The results - we hope - will be ready to be presented at the meeting. (author)

Analysis of gamma heating at TRIGA mark reactor core Bandung using plate type fuel

International Nuclear Information System (INIS)

Setiyanto; Tukiran Surbakti

2016-01-01

In accordance with the discontinuation of TRIGA fuel element production by its producer, the operation of all TRIGA type reactor of at all over the word will be disturbed, as well as TRIGA reactor in Bandung. In order to support the continuous operation of Bandung TRIGA reactor, a study on utilization of fuel plate mode, as used at RSG-GAS reactor, to replace the cylindrical model has been done. Various assessments have been done, including core design calculation and its safety aspects. Based on the neutronic calculation, utilization of fuel plate shows that Bandung TRIGA reactor can be operated by 20 fuel elements only. Compared with the original core, the new reactor core configuration is smaller and it results in some empty space that can be used for in-core irradiation facilities. Due to the existing of in-core irradiation facilities, the gamma heating value became a new factor that should be evaluated for safety analysis. For this reason, the gamma heating for TRIGA Bandung reactor using fuel plate was calculated by Gamset computer code. The calculations based on linear attenuation equations, line sources and gamma propagation on space. Calculations were also done for reflector positions (Lazy Susan irradiation facilities) and central irradiation position (CIP), especially for any material samples. The calculation results show that gamma heating for CIP is significantly important (0.87 W/g), but very low value for Lazy Susan position (lest then 0.11 W/g). Based on this results, it can be concluded that the utilization of CIP as irradiation facilities need to consider of gamma heating as data for safety analysis report. (author)

Operation experience of the advanced fuel assemblies at Unit 1 of Volgodonsk NPP within four fuel cycles

International Nuclear Information System (INIS)

Vasilchenko, I.; Kobelev, S.; Kushmanov, S.

2006-01-01

The first commissioning of Volgodonsk NPP Unit 1 with standard reactor WWER-1000 (project V-320) was in 2001. The reactor core, starting from the first fuel charge, was arranged completely with Advanced Fuel Assemblies (AFAs). In this way, it is possible to obtain the experience in startup and operation of the core, completely arranged with AFAs, and also to get a possibility of performing the comprehensive check for justification of newly commissioned units and justification of design solutions accepted in the design of reactor core for Taiwan NPP, Bushehr NPP and Kudankulam NPP. The first fuel charge of the Volgodonsk NPP Unit 1 is a reference and unified for Tiawan NPP (V-428), Bushehr NPP (V-446), Kudankulam NPP(V-412) with small differences caused by design features of RP V-320. The first core charge of Unit 1 of Volgodonsk NPP was arranged of 163 AFAs, comprising 61 CPS ARs and 42 BAR bundles. The subsequent fuel charges were arranged of AFAs with gadolinium oxide integrated into fuel instead of BAR. By 2005 the results of operation of the core at Unit 1 of Volgodonsk NPP during four fuel cycles showed that AFA is sufficiently reliable and serviceable. The activity of the primary coolant of the Volgodonsk NPP is at stable low level. During the whole time of the core operation of the Volgodonsk NPP Unit 1 no leaky AFAs were revealed. The modifications of the internals, made during pre-operational work, are reasonable and effective to provide for fuel mechanical stability in the course of operation. The modifications, made in AFA structure during operation of the Volgodonsk NPP Unit 1, are aimed at improving the service and operational reliability of its components. Correctness of the solutions taken is confirmed by AFAs operation experience both at the Volgodonsk NPP, and at other operating Russian NPPs

A digital data acquisition and display system for ITU TRIGA Mark II reactor

International Nuclear Information System (INIS)

Can, B.; Omuz, S.

2008-01-01

Full text: In this study, a digital data acquisition and display system realized for ITU TRIGA Mark-II Reactor is described. This system is realized in order to help the reactor operator and to increase reactor console capacity. The system consists of two main units, which are host computers and RTI-815F, analog devices, data acquisition card. RTI-815F is multi-function analog/digital input/output board that plugs into one of the available long expansion slots in the IBM-PC, PC/XT, PC/AT, or equivalent personal computers. It has 16 analog input channels for single-ended input signals or 8 analog input channels for differential input signals. But its channel capacity can be increased to 32 input channels for single-ended input signals or 16 input channels for differential input signals. RTI-815F board contains 2 analog output channels, 8 digital input channels and 8 digital output channels. In the ITD TRIGA Mark-II Reactor, 6 fuel temperature channels, 3 water temperature channels, 3 control rod position channels and 4 power channels are chosen as analog input signals for RTI-815F. Its digital outputs are assigned to cooling tower fan, primary and secondary pump reactor scram, control rod rundown. During operation, data are automatically archived to disk and displayed on screen. The channel selection time and sampling time can be adjusted. The simulated movement and position of control rods in the reactor core can be noted and displayed. The changes of power, fuel temperature and water temperature can be displayed on the screen as a graphic. In this system both period and reactivity are calculated and displayed on the screen. (authors)

Mutations in HNF1A Result in Marked Alterations of Plasma Glycan Profile

Science.gov (United States)

Thanabalasingham, Gaya; Huffman, Jennifer E.; Kattla, Jayesh J.; Novokmet, Mislav; Rudan, Igor; Gloyn, Anna L.; Hayward, Caroline; Adamczyk, Barbara; Reynolds, Rebecca M.; Muzinic, Ana; Hassanali, Neelam; Pucic, Maja; Bennett, Amanda J.; Essafi, Abdelkader; Polasek, Ozren; Mughal, Saima A.; Redzic, Irma; Primorac, Dragan; Zgaga, Lina; Kolcic, Ivana; Hansen, Torben; Gasperikova, Daniela; Tjora, Erling; Strachan, Mark W.J.; Nielsen, Trine; Stanik, Juraj; Klimes, Iwar; Pedersen, Oluf B.; Njølstad, Pål R.; Wild, Sarah H.; Gyllensten, Ulf; Gornik, Olga; Wilson, James F.; Hastie, Nicholas D.; Campbell, Harry; McCarthy, Mark I.; Rudd, Pauline M.; Owen, Katharine R.; Lauc, Gordan; Wright, Alan F.

2013-01-01

A recent genome-wide association study identified hepatocyte nuclear factor 1-α (HNF1A) as a key regulator of fucosylation. We hypothesized that loss-of-function HNF1A mutations causal for maturity-onset diabetes of the young (MODY) would display altered fucosylation of N-linked glycans on plasma proteins and that glycan biomarkers could improve the efficiency of a diagnosis of HNF1A-MODY. In a pilot comparison of 33 subjects with HNF1A-MODY and 41 subjects with type 2 diabetes, 15 of 29 glycan measurements differed between the two groups. The DG9-glycan index, which is the ratio of fucosylated to nonfucosylated triantennary glycans, provided optimum discrimination in the pilot study and was examined further among additional subjects with HNF1A-MODY (n = 188), glucokinase (GCK)-MODY (n = 118), hepatocyte nuclear factor 4-α (HNF4A)-MODY (n = 40), type 1 diabetes (n = 98), type 2 diabetes (n = 167), and nondiabetic controls (n = 98). The DG9-glycan index was markedly lower in HNF1A-MODY than in controls or other diabetes subtypes, offered good discrimination between HNF1A-MODY and both type 1 and type 2 diabetes (C statistic ≥0.90), and enabled us to detect three previously undetected HNF1A mutations in patients with diabetes. In conclusion, glycan profiles are altered substantially in HNF1A-MODY, and the DG9-glycan index has potential clinical value as a diagnostic biomarker of HNF1A dysfunction. PMID:23274891

Fuel performance analysis for the HAMP-1 mini plate test

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

Detection of a leaking boron-carbide control rod in a TRIGA Mark I reactor

Energy Technology Data Exchange (ETDEWEB)

Blotcky, A J; Arsenault, L J [General Medical Research, Veterans Administration Hospital, Omaha (United States)

1974-07-01

During a routine quarterly inspection of the boron-carbide control rods of the Omaha Veterans Administration Hospital 18 kW Triga Mark I reactor, a pin hole leak was detected approximately 3 mm from the chamfered edge. The leak was found by observing bubbles when the rod was withdrawn from the reactor tank for visual observation, and could not be seen with the naked eye. This suggests that pin hole leaks could occur and not be visually detected in control rods and fuel elements examined underwater. A review of the rod calibrations showed that the leak had not caused a loss in rod worth. Slides will be presented showing the bubbles observed during the inspection, together with an unmagnified and magnified view of the pin hole. (author)

Detection of a leaking boron-carbide control rod in a TRIGA Mark I reactor

International Nuclear Information System (INIS)

Blotcky, A.J.; Arsenault, L.J.

1974-01-01

During a routine quarterly inspection of the boron-carbide control rods of the Omaha Veterans Administration Hospital 18 kW Triga Mark I reactor, a pin hole leak was detected approximately 3 mm from the chamfered edge. The leak was found by observing bubbles when the rod was withdrawn from the reactor tank for visual observation, and could not be seen with the naked eye. This suggests that pin hole leaks could occur and not be visually detected in control rods and fuel elements examined underwater. A review of the rod calibrations showed that the leak had not caused a loss in rod worth. Slides will be presented showing the bubbles observed during the inspection, together with an unmagnified and magnified view of the pin hole. (author)

Postirradiation examination of Kori-1 nuclear power plant fuels

International Nuclear Information System (INIS)

Ro, S.G.; Kim, E.K.; Lee, K.S.; Min, D.K.

1994-01-01

Full size fuels discharged from Kori-1 PWR nuclear power plant have been subjected to postirradiation examination. The fuels under investigation were irradiated for one- to four-reactor cycles. Nondestructive examination and dismantling of the fuel assemblies have been conducted in the pool of the postirradiation examination facility (PIEF) of Korea Atomic Energy Research Institue. Subsequently nondestructive and destructive examinations of fuel rods have been performed in the hot cells of the PIEF. An evaluation of fuel burnup behaviors was based on the postirradiation examination data and the nominal design values. The results did not show any evidence of abnormalities in the fuel integrity. (orig.)

Postirradiation examination of Kori-1 nuclear power plant fuels

Science.gov (United States)

Seung-Gy, Ro; Eun-Ka, Kim; Key-Soon, Lee; Duck-Kee, Min

1994-05-01

Full size fuels discharged from Kori-1 PWR nuclear power plant have been subjected to postirradiation examination. The fuels under investigation were irradiated for one- to four-reactor cycles. Nondestructive examination and dismantling of the fuel assemblies have been conducted in the pool of the postirradiation examination facility (PIEF) of Korea Atomic Energy Research Institute. Subsequently nondestructive and destructive examinations of fuel rods have been performed in the hot cells of the PIEF. An evaluation of fuel burnup behaviors was based on the postirradiation examination data and the nominal design values. The results did not show any evidence of abnormalities in the fuel integrity.

Modeling and Implementation of a 1 kW, Air Cooled HTPEM Fuel Cell in a Hybrid Electrical Vehicle

DEFF Research Database (Denmark)

Andreasen, Søren Juhl; Ashworth, Leanne; Remón, Ian Natanael

2008-01-01

This work is a preliminary study of using the PBI-based, HTPEM fuel cell technology in automotive applications. This issue was investigated through computational modeling and an experimental investigation. A hybrid fuel cell system, consisting of a 1 kW stack and lead acid batteries, was implemen......This work is a preliminary study of using the PBI-based, HTPEM fuel cell technology in automotive applications. This issue was investigated through computational modeling and an experimental investigation. A hybrid fuel cell system, consisting of a 1 kW stack and lead acid batteries......, was implemented in a small electrical vehicle. A dynamic model was developed using Matlab-Simulink to describe the system characteristics, select operating conditions and to size system components. Preheating of the fuel cell stack with electrical resistors was investigated and found to be an unrealistic approach...

Design and construction of the SIPPING for fuels of the TRIGA Mark III reactor; Diseno y construccion del SIPPING para combustibles del reactor TRIGA Mark III

Energy Technology Data Exchange (ETDEWEB)

Castaneda J, G.; Delfin L, A.; Alvarado P, R.; Mazon R, R.; Ortega V, B. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: adl@nuclear.inin.mx

2003-07-01

The sipping technique, it has been used by several possessors of nuclear research reactors in its irradiated nuclear fuels, likewise in some fuel storage sites, with the objective of to determine the quantity of radioactivity that the fuel liberates in the means in that it is. The irradiated fuel in storage of some nuclear research reactors, its can have cracks that cross the cladding of the same one, generating the liberation of fission products that its need to determine to maintain safety measures appropriate as much as the fuel as of the facilities where they are. It doesn't exist until now, some method published for the non destructive sipping test technique. Based on that described, the Reactor Department of the National Institute of Nuclear Research, it has designed and built an inspection system of irradiated fuel that it will allow the detection of gassy fission products in site, and solids by means of the measurement of the activity of the Cs-137 contained in water samples. (Author)

[Werkgartner's muzzle imprint mark--a literature study].

Science.gov (United States)

Geserick, Gunther; Vendura, Klaus; Wirth, Ingo

2009-01-01

Since Werkgartner described and correctly interpreted the muzzle imprint mark around the gunshot entrance wound in 1922, this finding has been generally accepted as a sign of a contact shot. In further studies, it could finally be clarified that the muzzle imprint mark is caused by the expansive power of the powder gases with pressure on and abrasion of the skin at the muzzle (weapon imprint). Its shape depends on the firearm, the ammunition and the anatomical conditions, but does not require a bullet. Examinations under a magnifying glass microscope and histological investigations can complete the macroscopic findings. Occasionally, the muzzle imprint mark requires a certain "drying period" in order to become clearly visible. In rare cases, muzzle imprint marks also form on textiles perforated by the projectile. Characteristically shaped muzzled imprint marks can provide clues to the type of the firearm and its position at the time of discharge.

Mark I 1/5-scale boiling water reactor pressure suppression experiment. Quick-look report for test numbers 1.0(a) and 1.0(b) performed on March 4 and 8, 1977

International Nuclear Information System (INIS)

McCauley, E.W.; Pitts, J.H.

1977-01-01

The experimental results obtained from pressure suppression experiment numbers 1.0(a) and 1.0(b) that were performed on the Lawrence Livermore Laboratory's 1 / 5 -scale boiling water reactor (BWR) Mark I pressure suppression experimental facility are summarized

Development and validation of a model TRIGA Mark III reactor with code MCNP5; Desarrollo y validacion de un modelo del reactor Triga Mark III con el codigo MCNP5

Energy Technology Data Exchange (ETDEWEB)

Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Aguilar H, F., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2015-09-15

The main purpose of this paper is to obtain a model of the reactor core TRIGA Mark III that accurately represents the real operating conditions to 1 M Wth, using the Monte Carlo code MCNP5. To provide a more detailed analysis, different models of the reactor core were realized by simulating the control rods extracted and inserted in conditions in cold (293 K) also including an analysis for shutdown margin, so that satisfied the Operation Technical Specifications. The position they must have the control rods to reach a power equal to 1 M Wth, were obtained from practice entitled Operation in Manual Mode performed at Instituto Nacional de Investigaciones Nucleares (ININ). Later, the behavior of the K{sub eff} was analyzed considering different temperatures in the fuel elements, achieving calculate subsequently the values that best represent the actual reactor operation. Finally, the calculations in the developed model for to obtain the distribution of average flow of thermal, epithermal and fast neutrons in the six new experimental facilities are presented. (Author)

Postirradiation examinations of fuel pins from the GCFR F-1 series of mixed-oxide fuel pins at 5.5 at. % burnup

International Nuclear Information System (INIS)

Strain, R.V.; Johnson, C.E.

1978-05-01

Postirradiation examinations were performed on five fuel pins from the Gas-Cooled Fast-Breeder Reactor F-1 experiment irradiated in EBR-II to a peak burnup of approximately 5.5 at. %. These encapsulated fuel pins were irradiated at peak-power linear ratings from approximately 13 to 15 kW/ft and peak cladding inside diameter temperatures from approximately 625 to 760 0 C. The maximum diametral change that occurred during irradiation was 0.2% ΔD/D 0 . The maximum fuel-cladding chemical interaction depth was 2.6 mils in fuel pin G-1 and 1 mil or less in the other three pins examined destructively. Significant migration of the volatile fission products occurred axially to the fuel-blanket interfaces. Teh postirradiation examination data indicate that fuel melted at the inner surface of the annular fuel pellets in the two highest power rating fuel pins, but little axial movement of fuel occurred

Postirradiation examinations of fuel pins from the GCFR F-1 series of mixed-oxide fuel pins at 5. 5 at. % burnup

Energy Technology Data Exchange (ETDEWEB)

Strain, R V; Johnson, C E

1978-05-01

Postirradiation examinations were performed on five fuel pins from the Gas-Cooled Fast-Breeder Reactor F-1 experiment irradiated in EBR-II to a peak burnup of approximately 5.5 at. %. These encapsulated fuel pins were irradiated at peak-power linear ratings from approximately 13 to 15 kW/ft and peak cladding inside diameter temperatures from approximately 625 to 760/sup 0/C. The maximum diametral change that occurred during irradiation was 0.2% ..delta..D/D/sub 0/. The maximum fuel-cladding chemical interaction depth was 2.6 mils in fuel pin G-1 and 1 mil or less in the other three pins examined destructively. Significant migration of the volatile fission products occurred axially to the fuel-blanket interfaces. Teh postirradiation examination data indicate that fuel melted at the inner surface of the annular fuel pellets in the two highest power rating fuel pins, but little axial movement of fuel occurred.

Methanol supply issues for alternative fuels demonstration programs

International Nuclear Information System (INIS)

Teague, J.M.; Koyama, K.K.

1995-01-01

This paper surveys issues affecting the supply of fuel-grade methanol for the California Energy Commission's alternative fuels demonstration programs and operations by other public agencies such as transit and school districts. Establishing stable and reasonably priced sources of methanol (in particular) and of alternative fuels generally is essential to their demonstration and commercialization. Development both of vehicle technologies and of fuel supply and distribution are complementary and must proceed in parallel. However, the sequence of scaling up supply and distribution is not necessarily smooth; achievement of volume thresholds in demand and through-put of alternative fuels are marked by different kinds of challenges. Four basic conditions should be met in establishing a fuel supply: (1) it must be price competitive with petroleum-based fuels, at least when accounting for environmental and performance benefits; (2) bulk supply must meet volumes required at each phase; necessitating resilience among suppliers and a means of designating priority for high value users; (3) distribution systems must be reliable, comporting with end users' operational schedules; (4) volatility in prices to the end user for the fuel must be minimal. Current and projected fuel volumes appear to be insufficient to induce necessary economies of scale in production and distribution for fuel use. Despite their benefits, existing programs will suffer absent measures to secure economical fuel supplies. One solution is to develop sources that are dedicated to fuel markets and located within the end-use region

Radioactive waste management plan during the TRIGA Mark II and III decommissioning

International Nuclear Information System (INIS)

Jung, K.J.; Park, S.K.; Geong, G.H.; Lee, K.W.; Chung, U.S.; Paik, S.T.

2001-01-01

The decontamination and decommissioning (D and D) project of TRIGA Mark-I and Mark-II (KRR 1 and 2) was started in January 1997 and will be completed by December 2002. In the first year of the project, work was performed in preparation of the decommissioning plan, start of the environmental impact assessment and setup licensing procedure and documentation for the project with cooperation of the Korea Institute of Nuclear Safety (KINS). In the second year, Hyundai Engineering Company (HEC) with British Nuclear Fuels pie (BNFL) as technical assisting partner was designated as the contractor to do design and licensing documentation for the D and D of both reactors. After pre-design, a hazard and operability (HAZOP) study checked each step of the work. At the end of 1998, the decommissioning plan documentation including environmental impact assessment report was finished and submitted to the Ministry of Science and Technology (MOST) for licensing. It is expected to be issued by the end of September 1999. Practical work will then be started around the end of 1999. The safe treatment and management of the radioactive waste arising from the D and D activities is of utmost importance for successful completion of the practical dismantling work. This paper summarizes general aspects of radioactive waste treatment and management plan for the TRIGA Mark-I and II decommissioning work. (author)

27 CFR 28.223 - Export marks.

Science.gov (United States)

2010-04-01

... 27 Alcohol, Tobacco Products and Firearms 1 2010-04-01 2010-04-01 false Export marks. 28.223... Export marks. In addition to the marks and brands required to be placed on kegs, barrels, cases, crates... “Export” on each container or case before removal for export, for use on vessels or aircraft, or for...

Minimal Marking: A Success Story

Directory of Open Access Journals (Sweden)

Anne McNeilly

2014-11-01

Full Text Available The minimal-marking project conducted in Ryerson’s School of Journalism throughout 2012 and early 2013 resulted in significantly higher grammar scores in two first-year classes of minimally marked university students when compared to two traditionally marked classes. The “minimal-marking” concept (Haswell, 1983, which requires dramatically more student engagement, resulted in more successful learning outcomes for surface-level knowledge acquisition than the more traditional approach of “teacher-corrects-all.” Results suggest it would be effective, not just for grammar, punctuation, and word usage, the objective here, but for any material that requires rote-memory learning, such as the Associated Press or Canadian Press style rules used by news publications across North America.

RECH-1 test fuel irradiation status report

International Nuclear Information System (INIS)

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

2005-01-01

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

Performance and management of IPR-R1 fuel elements

International Nuclear Information System (INIS)

Stasiulevicius, R.; Maretti Junior, F.

1983-01-01

The performance of fuel elements during the 23 years of the reactor operation, is presented aiming to introduce improvements in the fuel load distribution and consequent increase of the reactivity. A computer code CORE was developed aiming to calculate the individual burnup of the fuel elements and the value of the reactivity for several core configurations, establishing a routine to control the nuclear material in the IPR-R1. The values calculated were compared with the experimental results. Some alternatives to augment the reactivity of the present core are presented foreseeing the fuel load availability for operation with 100Km and, for angmenting the power reaction in a next stage. (E.G.) [pt

Reuse of discharged fuel in Bohunice-1,2 units

International Nuclear Information System (INIS)

Chrapciak, V.; Majercik, J.; Kacmar, M.

2003-01-01

During the reconstruction of Bohunice-1,2 units (1997 - 2001), their cycle lengths dropped to very short values. Because of 4-year limit to fuel residence time, refuelling with fresh 2.4 % enriched assemblies seemed to be a solution of the problem. The paper describes the implementation of a final decision to reuse 3.6 % enriched fuel discharged after 3-year irradiation in previous cycles. This decision led to a large-scale moving of discharged assemblies from spent fuel pools back to reactors (Authors)

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

International Nuclear Information System (INIS)

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

1989-11-01

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

Power Reactor Fuel Reprocessing Plant-1: a stepping stone in Indian PHWR spent fuel reprocessing

International Nuclear Information System (INIS)

Pradhan, Sanjay; Dubey, K.; Qureshi, F.T.; Lokeswar, S.P.

2017-01-01

India has low reserves of uranium and high reserves of thorium. In order to optimize resource utilization India has adopted a closed fuel cycle to ensure long-term energy security. The optimum resource utilization is feasible only by adopting reprocessing, conditioning and recycle options. It is very much imperative to view spent fuel as a vital resource material and not a waste to be disposed off. Thus, spent nuclear fuel reprocessing forms an integral part of the Indian Nuclear Energy Programme. Aqueous reprocessing based on PUREX technology is in use for more than 50 years and has reached a matured status

Validation of KENO V.a for criticality safety calculations involving WR-1 fast-neutron fuel arrangements

Energy Technology Data Exchange (ETDEWEB)

Gauld, I. C.

1991-07-15

The KENO V.a criticality safety code, used with the SCALE 27-energy-group ENDF/B-IV-based cross-section library, has been validated for low-enriched uranium carbide (UC) WR-1 fast-neutron (FN) fuel arrangements. Because of a lack of relevant experimental data for UC fuel in the published literature, the validation is based primarily on calculational comparisons with critical experiments for fuel types with a range of enrichments and densities that cover those of the FN UC fuel. The ability of KENO V.a to handle the unique annular pin arrangement of the WR-1 FN fuel bundle was established using a comparison with the MCNP3B code used with a continuous-energy ENDF/B-V-based cross-section library. This report is part of the AECL--10146 report series documenting the validation of the KENO V.a criticality safety code.

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

International Nuclear Information System (INIS)

Bukhari, K.M.

1990-09-01

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

TRESK background K(+ channel is inhibited by PAR-1/MARK microtubule affinity-regulating kinases in Xenopus oocytes.

Directory of Open Access Journals (Sweden)

Gabriella Braun

Full Text Available TRESK (TWIK-related spinal cord K(+ channel, KCNK18 is a major background K(+ channel of sensory neurons. Dominant-negative mutation of TRESK is linked to familial migraine. This important two-pore domain K(+ channel is uniquely activated by calcineurin. The calcium/calmodulin-dependent protein phosphatase directly binds to the channel and activates TRESK current several-fold in Xenopus oocytes and HEK293 cells. We have recently shown that the kinase, which is responsible for the basal inhibition of the K(+ current, is sensitive to the adaptor protein 14-3-3. Therefore we have examined the effect of the 14-3-3-inhibited PAR-1/MARK, microtubule-associated-protein/microtubule affinity-regulating kinase on TRESK in the Xenopus oocyte expression system. MARK1, MARK2 and MARK3 accelerated the return of TRESK current to the resting state after the calcium-dependent activation. Several other serine-threonine kinase types, generally involved in the modulation of other ion channels, failed to influence TRESK current recovery. MARK2 phosphorylated the primary determinant of regulation, the cluster of three adjacent serine residues (S274, 276 and 279 in the intracellular loop of mouse TRESK. In contrast, serine 264, the 14-3-3-binding site of TRESK, was not phosphorylated by the kinase. Thus MARK2 selectively inhibits TRESK activity via the S274/276/279 cluster, but does not affect the direct recruitment of 14-3-3 to the channel. TRESK is the first example of an ion channel phosphorylated by the dynamically membrane-localized MARK kinases, also known as general determinants of cellular polarity. These results raise the possibility that microtubule dynamics is coupled to the regulation of excitability in the neurons, which express TRESK background potassium channel.

Physics operating experience and fuel management of RAPS-1

International Nuclear Information System (INIS)

Nakra, A.N.; Purandare, H.D.; Srinivasan, K.R.; Rastogi, B.P.

1976-01-01

Rajasthan Atomic Power Station Unit-1 achieved criticality on August 11, 1972. Thereafter the reactor was brought to power, in November, 1972. Due to non-availability of the depleted fuel, the loading of which was necessary to obtain full power to begin with, the core was loaded with all natural uranium fuel and only 70% of the full power could be achieved. During the reactor operation for the last three years, the reactor has seen more than one effective full power year and about 1400 fresh fuel bundles have been loaded in the core. The reactor was subjected to about 150 power cycles resulting in more than 30% variation in operating power level and about 10 fuel bundles have failed. For satisfactory fuel management and refuelling decisions, a three dimensional simulator TRIVENI was developed. This was extensively tested during the start-up experiments and was found to be a satisfactory tool for day to day operation of the plant. In this paper, a brief account of analysis of the start-up experiments, approach to full power, power distortions and flux peaking, fuel management service and analysis of the failed fuel data has been given. (author)

Used mixed oxide fuel reprocessing at RT-1 plant

Energy Technology Data Exchange (ETDEWEB)

Kolupaev, D.; Logunov, M.; Mashkin, A.; Bugrov, K.; Korchenkin, K. [FSUE PA ' Mayak' , 30, Lenins str, Ozersk, 460065 (Russian Federation); Shadrin, A.; Dvoeglazov, K. [ITCP ' PRORYV' , 2/8 Malaya Krasmoselskay str, Moscow, 107140 (Russian Federation)

2016-07-01

Reprocessing of the mixed uranium-plutonium spent nuclear fuel of the BN-600 reactor was performed at the RT-1 plant twice, in 2012 and 2014. In total, 8 fuel assemblies with a burn-up from 73 to 89 GW day/t and the cooling time from 17 to 21 years were reprocessed. The reprocessing included the stages of dissolution, clarification, extraction separation of U and Pu with purification from the fission products, refining of uranium and plutonium at the relevant refining cycles. Dissolution of the fuel composition of MOX used nuclear fuel (UNF) in nitric acid solutions in the presence of fluoride ion has occurred with the full transfer of actinides into solution. Due to the high content of Pu extraction separation of U and Pu was carried out on a nuclear-safe equipment designed for the reprocessing of highly enriched U spent nuclear fuel and Pu refining. Technological processes of extraction, separation and refining of actinides proceeded without deviations from the normal mode. The output flow of the extraction outlets in their compositions corresponded to the regulatory norms and remained at the level of the compositions of the streams resulting from the reprocessing of fuel types typical for the RT-1 plant. No increased losses of Pu into waste have been registered during the reprocessing of BN-600 MOX UNF an compare with VVER-440 uranium UNF reprocessing. (authors)

Dilatational behaviour of ZrNb1 fuel cans of a WWER-type reactor during a loss-of-coolant accident

International Nuclear Information System (INIS)

Adam, E.; Stephan, M.; Wetzel, L.

1987-01-01

Based on an assessment of various factors of influence on the performance of fuel cans during normal operation and imaginable accidents, the necessity of studying creep and burst behaviour of WWER-type fuel cans of ZrNb1 under simulated LOCA conditions has been proved and an experimental facility designed for this purpose is described. Control of fuel can temperature is accomplished through a minicomputer during the creep and bursts experiments. With this, various temperature loading profiles of the fuel cans can be realized. Experimental results on dilatational behaviour of ZrNb1 fuel cans from isothermal creep and burst experiments in air are presented and compared with values for Zircaloy. (author)

19 CFR 134.43 - Methods of marking specific articles.

Science.gov (United States)

2010-04-01

... 19 Customs Duties 1 2010-04-01 2010-04-01 false Methods of marking specific articles. 134.43...; DEPARTMENT OF THE TREASURY COUNTRY OF ORIGIN MARKING Method and Location of Marking Imported Articles § 134.43 Methods of marking specific articles. (a) Marking previously required by certain provisions of the...

Irradiation-induced dimensional changes of fuel compacts and graphite sleeves of OGL-1 fuel assemblies

International Nuclear Information System (INIS)

Hayashi, Kimio; Minato, Kazuo; Kobayashi, Fumiaki; Tobita, Tsutomu; Kikuchi, Teruo; Kurobane, Shiro; Adachi, Mamoru; Fukuda, Kousaku

1988-06-01

Experimental data are summarized on irradiation-induced dimensional changes of fuel compacts and graphite sleeves of the first to ninth OGL-1 fuel assemblies. The range of fast-neutron fluence is up to 4 x 10 24 n/m 2 (E > 0.18 MeV); and that of irradiation temperature is 900 - 1400 deg C for fuel compacts and 800 - 1050 deg C for graphite sleeves. The dimensional change of the fuel compacts was shrinkage under these test conditions, and the shrinkage fraction increased almost linearly with fast-neutron fluence. The shrinkage fraction of the fuel compacts was larger by 20 % in the axial direction than in the radial direction. Influence of the irradiation temperature on the dimensional-change behavior of the fuel compacts was not observed clearly; presumably the influence was hidden by scatter of the data because of low level of the fast-neutron fluence and the resultant small dimensional changes. (author)

A natural-gas fuel processor for a residential fuel cell system

Science.gov (United States)

Adachi, H.; Ahmed, S.; Lee, S. H. D.; Papadias, D.; Ahluwalia, R. K.; Bendert, J. C.; Kanner, S. A.; Yamazaki, Y.

A system model was used to develop an autothermal reforming fuel processor to meet the targets of 80% efficiency (higher heating value) and start-up energy consumption of less than 500 kJ when operated as part of a 1-kWe natural-gas fueled fuel cell system for cogeneration of heat and power. The key catalytic reactors of the fuel processor - namely the autothermal reformer, a two-stage water gas shift reactor and a preferential oxidation reactor - were configured and tested in a breadboard apparatus. Experimental results demonstrated a reformate containing ∼48% hydrogen (on a dry basis and with pure methane as fuel) and less than 5 ppm CO. The effects of steam-to-carbon and part load operations were explored.

Mark Raidpere näitused Pariisis ja Napolis / Mark Raidpere ; interv. Harry Liivrand

Index Scriptorium Estoniae

Raidpere, Mark

2008-01-01

Mark Raidpere videod "Vekovka", "Dedication / Pühendus", "Majestoso Mystico" näitusel Pariisis Michel Reini galeriis. Osaleb koos saksa fotograafi Sven Johnega näitusel Napolis. Kreekas Thessalonikis valminud filmist "1:1:1"

Construction and tests of a gamma device for experimental measurements of burnup of nuclear reactor fuel

International Nuclear Information System (INIS)

Brandao Junior, F.A.

1982-01-01

The gamma-scanning method is an important tool for the measurement of burnup of nuclear reactor fuel. The adequate knowledge of burnup allows for a better inventory of 'sensitive' fissile materials, better fuel management and provides insight on fuel behaviour and safety margins. This paper is related to the description, construction and operation of a first gamma scanning device, tested by irradiation of prototype PWR fuel pins, 14 cm long, in a Triga Mark-I reactor at very low power. Despite the limitations imposed by the low burnup, the experiment permitted a good checking of the main physical concepts and devices involved in the method. (Author) [pt

Coupling of a 2.5 kW steam reformer with a 1 kW el PEM fuel cell

Science.gov (United States)

Mathiak, J.; Heinzel, A.; Roes, J.; Kalk, Th.; Kraus, H.; Brandt, H.

The University of Duisburg-Essen has developed a compact multi-fuel steam reformer suitable for natural gas, propane and butane. This steam reformer was combined with a polymer electrolyte membrane fuel cell (PEM FC) and a system test of the process chain was performed. The fuel processor comprises a prereformer step, a primary reformer, water gas shift reactors, a steam generator, internal heat exchangers in order to achieve an optimised heat integration and an external burner for heat supply as well as a preferential oxidation step (PROX) as CO purification. The fuel processor is designed to deliver a thermal hydrogen power output from 500 W to 2.5 kW. The PEM fuel cell stack provides about 1 kW electrical power. In the following paper experimental results of measurements of the single components PEM fuel cell and fuel processor as well as results of the coupling of both to form a process chain are presented.

19 CFR 11.9 - Special marking on certain articles.

Science.gov (United States)

2010-04-01

... 19 Customs Duties 1 2010-04-01 2010-04-01 false Special marking on certain articles. 11.9 Section... OF THE TREASURY PACKING AND STAMPING; MARKING Marking § 11.9 Special marking on certain articles. (a... of additional U.S. Note 4, Chapter 91. If any article so required to be marked is found not to be...

LMFBR fuel analysis. Task A: Oxide fuel dynamics. Final report, October 1, 1976--September 30, 1977

International Nuclear Information System (INIS)

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

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

International Nuclear Information System (INIS)

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)

COMPUTER HARDWARE MARKING

CERN Multimedia

Groupe de protection des biens

2000-01-01

As part of the campaign to protect CERN property and for insurance reasons, all computer hardware belonging to the Organization must be marked with the words 'PROPRIETE CERN'.IT Division has recently introduced a new marking system that is both economical and easy to use. From now on all desktop hardware (PCs, Macintoshes, printers) issued by IT Division with a value equal to or exceeding 500 CHF will be marked using this new system.For equipment that is already installed but not yet marked, including UNIX workstations and X terminals, IT Division's Desktop Support Service offers the following services free of charge:Equipment-marking wherever the Service is called out to perform other work (please submit all work requests to the IT Helpdesk on 78888 or helpdesk@cern.ch; for unavoidable operational reasons, the Desktop Support Service will only respond to marking requests when these coincide with requests for other work such as repairs, system upgrades, etc.);Training of personnel designated by Division Leade...

Rokkashomura: debut of the nuclear fuel cycle business

International Nuclear Information System (INIS)

Anon.

1991-01-01

Japan Nuclear Fuel Industries and local governments signed the safety agreement, and the work began to initiate the operation of a uranium enrichment plant. In this way, the national Rokkashomura project to be constructed with the total cost of 1.2 trillion yen marked the debut of nuclear fuel cycle business in Japan. The public hearing concerning the low level radioactive waste storage facility was finished. However, a fuel reprocessing plant has not advanced since the national government did not clarify the policy for the management of high level rad-waste from the plant. Gubernatorial election was the best thing to happen for the public acceptance, and the local opposition movement lost steam. The operation of the uranium enrichment plant is to begin next January, and the construction of the low level waste storage facility proceeds on schedule. Regarding the fuel reprocessing plant, the public hearing is to be held in autumn, but it faces difficulties. The siting of nuclear fuel cycle facilities has already produced benefits for the local economy. 18 business establishments representing 15 firms have so far decided to open in Aomori Prefecture. JNFI and JNFS began the specific study for merger. (K.I.)

Perception of scent over-marks by golden hamsters (Mesocricetus auratus): novel mechanisms for determining which individual's mark is on top.

Science.gov (United States)

Johnston, R E; Bhorade, A

1998-09-01

Hamsters preferentially remember or value the top scent of a scent over-mark. What cues do they use to do this? Using habituation-discrimination techniques, we exposed male golden hamsters (Mesocricetus auratus) on 3 to 4 trials to genital over-marks from 2 females and then tested subjects for their familiarity with these 2 scents compared with that of a novel female's secretion. Preferential memory for 1 of the 2 individuals' scents did not occur if the 2 marks did not overlap or did not overlap but differed in age, but it did occur if a region of overlap existed or 1 mark apparently occluded another (but did not overlap it). Thus, hamsters use regions of overlap and the spatial configuration of scents to evaluate over-marks. These phenomena constitute evidence for previously unsuspected perceptual abilities, including olfactory scene analysis, which is analogous to visual and auditory scene analysis.

Structural integrity assessment and stress measurement of CHASNUPP-1 fuel assembly

Directory of Open Access Journals (Sweden)

Waseem

2016-01-01

Full Text Available Fuel assembly of the PWR nuclear power plant is a long and flexible structure. This study has been made in an attempt to find the structural integrity of the fuel assembly (FA of Chashma Nuclear Power Plant-1 (CHASNUPP-1 at room temperature in air. The non-linear contact and structural tensile analysis have been performed using ANSYS 13.0, in order to determine the fuel assembly (FA elongation behaviour as well as the location and values of the stress intensity and stresses developed in axial direction under applied tensile load of 9800 N or 2 g being the fuel assembly handling or lifting load [Y. Zhang et al., Fuel assembly design report, SNERDI, China, 1994]. The finite element (FE model comprises spacer grids, fuel rods, flexible contacts between the fuel rods and grid's supports system and guide thimbles with dash-pots and flow holes, in addition to the spot welds between spacer grids and guide thimbles, has been developed using Shell181, Conta174 and Targe170 elements. FA is a non-straight structure. The actual behavior of the geometry is non-linear due to its curvature or design tolerance. It has been observed that fuel assembly elongation values obtained through FE analysis and experiment [SNERDI Tech. Doc., Mechanical strength and calculation for fuel assembly, Technical Report, F3.2.1, China, 1994] under applied tensile load are comparable and show approximately linear behaviors. Therefore, it seems that the permanent elongation of fuel assembly may not occur at the specified load. Moreover, the values of stresses obtained at different locations of the fuel assembly are also comparable with the stress values of the experiment determined at the same locations through strain gauges. Since the results of both studies (analytical and experimental are comparable, therefore, validation of the FE methodology is confirmed. The stress intensity of the FE model and maximum stresses developed along the guide thimbles in axial direction are

Activation calculation of steel of the control rods of TRIGA Mark III reactor; Calculo de activacion del acero de las barras de control del reactor TRIGA Mark III

Energy Technology Data Exchange (ETDEWEB)

Garcia M, T.; Cruz G, H. S.; Ruiz C, M. A.; Angeles C, A., E-mail: teodoro.garcia@inin.gob.mx [ININ, Carretera Mexico-Toluca sn, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2014-10-15

In the pool of TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ), there are control rods that were removed from the core, and which are currently on shelves of decay. These rods were part of the reactor core when only had fuel standard (from 1968-1989). To conduct a proper activation analysis of the rods, is very important to have well-characterized the materials which are built, elemental composition of the same ones, the atomic densities and weight fractions of the elements that constitute them. To determine the neutron activation of the control rods MCNP5 code was used, this code allows us to have well characterized the radionuclides inventory that were formed during irradiation of the control rods. This work is limited to determining the activation of the steel that is part of the shielding of the control rods, the nuclear fuel that is in the fuel follower does not include. The calculation model of the code will be validated with experimental measurements and calculating the activity of fission products of the fuel follower which will take place at the end of 2014. (Author)

Mark I 1/5-scale boiling water reactor pressure suppresion experiment quick-look report

International Nuclear Information System (INIS)

Lai, W.; Collins, E.K.

1977-01-01

This report is intended as a ''quick-look'' report summarizing the experimental results obtained from pressure suppression experiment numbers 2.1, 2.2, and 2.3 that were performed on the Lawrence Livermore Laboratory's 1/5-scale boiling water reactor (BWR) Mark I pressure suppression experimental facility on April 26, 1977. A brief description of the general nature of the tests and a summary of the actual tests that were performed are given

46 CFR 199.176 - Markings on lifesaving appliances.

Science.gov (United States)

2010-10-01

... ARRANGEMENTS LIFESAVING SYSTEMS FOR CERTAIN INSPECTED VESSELS Requirements for All Vessels § 199.176 Markings on lifesaving appliances. (a) Lifeboats and rescue boats. Each lifeboat and rescue boat must be plainly marked as follows: (1) Each side of each lifeboat and rescue boat bow must be marked in block...

Project fuel development

International Nuclear Information System (INIS)

Stratton, R.W.

1981-05-01

The activities continued on lab-scale production of uranium-plutonium carbide fuel for the fast reactor using gelation methods, irradiation testing and performance evaluation. Whereas in earlier years a balance was maintained between research and development or with emphasis on research, 1980 was marked by a concentrated equipment development effort for an increased throughput with improved process control and product reproducability and installation of new equipment for large pin fabrication. (Auth.)

AKORT-1 on-line system for technological control of the mixed fuel distribution in fuel elements

International Nuclear Information System (INIS)

Baklanov, V.S.; Besednov, G.Yu.; Gadzhiev, G.I.

1982-01-01

An on-line system for technological control of experimental fuel elements with vibrocompacted UO 2 -PuO 2 fuel rods fabricated for the BOR-60 reactor is described. Equipment and performance specific features of the system mechanical part and electronic circuits are considered. The results of the system performance testing are given. The fuel element quality sorting is made on the base of the analysis of Pu and fuel density distributions in the rod length. Gamma-absorption method for density measuring and the method of Pu content determination by its own gamma radiation are used in the system simultaneously. The system has the following main characteristics: tested fuel element diameter is 6 mm; the range of fuel rod mean densities is 7-10 g/cm 3 ; Pu content in the fuel is more than 20%; gamma detectors are the NaI(Tl) detectors with dimensions 40x40 and 25x25 mm; energy resolution is 137 Cs gamma line. Electronic circuits of the system operating on-line with the D3-28 microcomputer are made using the VECTOR standard. The system testing has shown that the error in the fuel density determination is less than 1%, that for Pu content measuring is 4%, the system capacity is 6 fuel elements per hour

International fuel cycle centres offer large economics and easier financing

International Nuclear Information System (INIS)

Smith, D.

1977-01-01

The summary report of the IAEA study project on multi-national regional nuclear fuel cycle indicates that for facilities of reasonable size such projects offer very decisive advantages in fuel cycle costs and resource availability over national facilities in general, and more markedly over the other alternative of the open ended, non-recycle fuel route. The economic evaluation of alternative fuel cycle strategies, one of the basic studies summarised in the report, is considered. (author)

25 CFR 141.16 - Price marking.

Science.gov (United States)

2010-04-01

... 25 Indians 1 2010-04-01 2010-04-01 false Price marking. 141.16 Section 141.16 Indians BUREAU OF... AND ZUNI RESERVATIONS General Business Practices § 141.16 Price marking. The price of each article... visible to the customer and that affords the customer a reasonable opportunity to learn the price of the...

Validation of a Waste Heat Recovery Model for a 1kW PEM Fuel Cell using Thermoelectric Generator

Science.gov (United States)

Saufi Sulaiman, M.; Mohamed, W. A. N. W.; Singh, B.; Fitrie Ghazali, M.

2017-08-01

Fuel cell is a device that generates electricity through electrochemical reaction between hydrogen and oxygen. A major by-product of the exothermic reaction is waste heat. The recovery of this waste heat has been subject to research on order to improve the overall energy utilization. However, nearly all of the studies concentrate on high temperature fuel cells using advanced thermodynamic cycles due to the high quality of waste heat. The method, characteristics and challenges in harvesting waste heat from a low temperature fuel cell using a direct energy conversion device is explored in this publication. A heat recovery system for an open cathode 1kW Proton Exchange Membrane fuel cell (PEM FC) was developed using a single unit of thermoelectric generator (TEG) attached to a heat pipe. Power output of the fuel cell was varied to obtain the performance of TEG at different stack temperatures. Natural and forced convections modes of cooling were applied to the TEG cold side. This is to simulate the conditions of a mini fuel cell vehicle at rest and in motion. The experimental results were analysed and a mathematical model based on the thermal circuit analogy was developed and compared. Forced convection mode resulted in higher temperature difference, output voltage and maximum power which are 3.3°C, 33.5 mV, and 113.96mW respectively. The heat recovery system for 1 kW Proton Exchange Membrane fuel cell (PEM FC) using single TEG was successfully established and improved the electrical production of fuel cell. Moreover, the experimental results obtained was in a good agreement with theoretical results.

Marked hyperandrogenicity in a 60-year-old woman

Directory of Open Access Journals (Sweden)

Khaled Aljenaee

2017-09-01

Full Text Available Markedly elevated androgen levels can lead to clinical virilization in females. Clinical features of virilization in a female patient, in association with biochemical hyperandrogenism, should prompt a search for an androgen-producing tumor, especially of ovarian or adrenal origin. We herein report the case of a 60-year-old woman of Pakistani origin who presented with the incidental finding of male pattern baldness and hirsutism. Her serum testosterone level was markedly elevated at 21 nmol/L (normal range: 0.4–1.7 nmol/L, while her DHEAS level was normal, indicating a likely ovarian source of her elevated testosterone. Subsequently, a CT abdomen-pelvis was performed, which revealed a bulky right ovary, confirmed on MRI of the pelvis as an enlarged right ovary, measuring 2.9 × 2.2 cm transaxially. A laparoscopic bilateral salpingo-oophorectomy was performed, and histopathological examination and immunohistochemistry confirmed the diagnosis of a Leydig cell tumor, a rare tumor accounting for 0.1% of ovarian tumors. Surgical resection led to normalization of testosterone levels.

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

International Nuclear Information System (INIS)

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

1985-01-01

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

Research reactor fuel - an update

International Nuclear Information System (INIS)

Finlay, M.R.; Ripley, M.I.

2003-01-01

In the two years since the last ANA conference there have been marked changes in the research reactor fuel scene. A new low-enriched uranium (LEU) fuel, 'monolithic' uranium molybdenum, has shown such promise in initial trials that it may be suitable to meet the objectives of the Joint Declaration signed by Presidents Bush and Putin to commit to converting all US and Russian research reactors to LEU by 2012. Development of more conventional aluminium dispersion UMo LEU fuel has continued in the meantime and is entering the final qualification stage of multiple full sized element irradiations. Despite this progress, the original 2005 timetable for UMo fuel qualification has slipped and research reactors, including the RRR, may not convert from silicide to UMo fuel before 2007. The operators of the Swedish R2 reactor have been forced to pursue the direct route of qualifying a UMo lead test assembly (LTA) in order to meet spent fuel disposal requirements of the Swedish law. The LTA has recently been fabricated and is expected to be loaded shortly into the R2 reactor. We present an update of our previous ANA paper and details of the qualification process for UMo fuel

Qualification process of dispersion fuels in the IEAR1 research reactor

International Nuclear Information System (INIS)

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

2010-01-01

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

Reconfiguring trade mark law

DEFF Research Database (Denmark)

Elsmore, Matthew James

2013-01-01

-border setting, with a particular focus on small business and consumers. The article's overall message is to call for a rethink of received wisdom suggesting that trade marks are effective trade-enabling devices. The case is made for reassessing how we think about European trade mark law.......First, this article argues that trade mark law should be approached in a supplementary way, called reconfiguration. Second, the article investigates such a reconfiguration of trade mark law by exploring the interplay of trade marks and service transactions in the Single Market, in the cross...

Bite marks on skin and clay: A comparative analysis

Directory of Open Access Journals (Sweden)

R.K. Gorea

2014-12-01

Full Text Available Bite marks are always unique because teeth are distinctive. Bite marks are often observed at the crime scene in sexual and in physical assault cases on the skin of the victims and sometimes on edible leftovers in burglary cases. This piece of evidence is often ignored, but if properly harvested and investigated, bite marks may prove useful in apprehending and successfully prosecuting the criminals. Due to the importance of bite marks, we conducted a progressive randomised experimental study conducted on volunteers. A total of 188 bite marks on clay were studied. Based on these findings, 93.34% of the volunteers could be identified from the bite marks on the clay. In addition, 201 impressions on skin were studied, and out of these cases, 41.01% of the same volunteers could be identified based on the bite mark impressions on the skin.

PREOPERATIVE ENDOSCOPIC MARKING OF UNPALPABLE COLONIC TUMORS

Directory of Open Access Journals (Sweden)

A. L. Goncharov

2013-01-01

Full Text Available The identification of small colon lesions is one of the major problems in laparoscopic colonic resection.Research objective: to develop a technique of visualization of small tumors of a colon by preoperative endoscopic marking of a tumor.Materials and methods. In one day prior to operation to the patient after bowel preparation the colonoscopy is carried out. In the planned point near tumor on antimesentery edge the submucous infiltration of marking solution (Micky Sharpz blue tattoo pigment, UK is made. The volume of entered solution of 1–3 ml. In only 5 months of use of a technique preoperative marking to 14 patients with small (the size of 1–3 cm malignant tumors of the left colon is performed.Results. The tattoo mark was well visualized by during operation at 13 of 14 patients. In all cases we recorded no complications. Time of operation with preoperative marking averaged 108 min, that is significantly less in comparison with average time of operation with an intra-operative colonoscopy – 155 min (р < 0.001.Conclusions. The first experience of preoperative endoscopic marking of non palpable small tumors of a colon is encouraging. Performance of a technique wasn't accompanied by complications and allowed to reduce significantly time of operation and to simplify conditions of performance of operation.

Cumulative damage fraction design approach for LMFBR metallic fuel elements

International Nuclear Information System (INIS)

Johnson, D.L.; Einziger, R.E.; Huchman, G.D.

1979-01-01

The cumulative damage fraction (CDF) analytical technique is currently being used to analyze the performance of metallic fuel elements for proliferation-resistant LMFBRs. In this technique, the fraction of the total time to rupture of the cladding is calculated as a function of the thermal, stress, and neutronic history. Cladding breach or rupture is implied by CDF = 1. Cladding wastage, caused by interactions with both the fuel and sodium coolant, is assumed to uniformly thin the cladding wall. The irradiation experience of the EBR-II Mark-II driver fuel with solution-annealed Type 316 stainless steel cladding provides an excellent data base for testing the applicability of the CDF technique to metallic fuel. The advanced metal fuels being considered for use in LMFBRs are U-15-Pu-10Zr, Th-20Pu and Th-2OU (compositions are given in weight percent). The two cladding alloys being considered are Type 316 stainless steel and a titanium-stabilized Type 316 stainless steel. Both are in the cold-worked condition. The CDF technique was applied to these fuels and claddings under the assumed steady-state operating conditions

Postirradiation examination and evaluation of Fort St. Vrain fuel element 1-0743

International Nuclear Information System (INIS)

Saurwein, J.J.; Miller, C.M.; Young, C.A.

1981-05-01

Fort St. Vrain (FSV) fuel element 1-0743 was irradiated in core location 17.04.F.06 from July 3, 1976 until February 1, 1979. The element experienced an average fast neutron exposure of about 0.95 x 10 25 n/m 2 (E > 29 fJ)/sub HTGR/, a time-and-volume-averaged fuel temperature in the vicinity of 680 0 C, fissile and fertile particle burnups of approximately 6.2% and 0.3%, respectively, and a total burnup of 12,210 MWd/tonne. The postirradiation examination revealed that the element was in excellent condition. No cracks were observed on any of the element surfaces. The structural integrity of the fuel rods was good. No evidence of mechanical interaction between the fuel rods and fuel body was observed. Calculated irradiation parameters obtained with HTGR design codes were compared with measured data. Radial and axial power distributions, irradiation temperatures, neutron fluences, and fuel burnups were in good agreement with measurements. Calculated fuel rod strains were about a factor of three greater than were observed

A life-cycle perspective on automotive fuel cells

International Nuclear Information System (INIS)

Simons, Andrew; Bauer, Christian

2015-01-01

Highlights: • Individual inventories for each fuel cell system component, current and future. • Environmental and human health burdens from fuel cell production and end-of-life. • Comparison passenger transport in fuel cell and conventional vehicles. • Fuel cell can be more critical to overall burdens than hydrogen production. • Fuel cell developments require radical but possible changes to reduce burdens. - Abstract: The production and end-of-life (EoL) processes for current and future proton exchange membrane fuel cell (PEMFC) systems for road passenger vehicle applications were analysed and quantified in the form of life cycle inventories. The current PEMFC technology is characterised by highly sensitive operating conditions and a high system mass. For each core component of PEMFC there are a range of materials under development and the research aimed to identify those considered realistic for a 2020 future scenario and according to commercial goals of achieving higher performance, increased power density, greater stability and a marked reduction of costs. End-of-life scenarios were developed in consideration of the materials at the focus of recovery efforts. The life cycle impact assessment (LCIA) addressed the production and EoL of the fuel cell systems with inclusion of a sensitivity analysis to assess influences on the results from the key fuel cell parameters. The second part to the LCIA assessed the environmental and human health burdens from passenger transport in a fuel cell vehicle (FCV) with comparison between the 2012 and 2020 fuel cell scenarios and referenced to an internal combustion engine vehicle (ICEV) of Euro5 emission standard. It was seen that whilst the drivetrain (and therefore the fuel cell system) is a major contributor to the emissions in all the indicators shown, the hydrogen use (and therefore the efficiency of the fuel cell system and the method of hydrogen production) can have a far greater influence on the environmental

AUTOMOTIVE DIESEL MAINTENANCE 1. UNIT III, MAINTAINING THE FUEL SYSTEM--DETROIT DIESEL ENGINE.

Science.gov (United States)

Human Engineering Inst., Cleveland, OH.

THIS MODULE OF A 30-MODULE COURSE IS DESIGNED TO DEVELOP AN UNDERSTANDING OF THE OPERATION AND MAINTENANCE OF THE DIESEL ENGINE FUEL SYSTEM. TOPICS ARE (1) PURPOSE OF THE FUEL SYSTEM, (2) TRACING THE FUEL FLOW, (3) MINOR COMPONENTS OF THE FUEL SYSTEM, (4) MAINTENANCE TIPS, (5) CONSTRUCTION AND FUNCTION OF THE FUEL INJECTORS, AND (6)…

Nine years of operation of ITU-TRR TRIGA Mark-II reactor

International Nuclear Information System (INIS)

Yavuz, H.; Bayuelken, A.R.; Yavuz, M.A.

1988-01-01

ITU-TRR TRIGA Mark-II reactor in Istanbul with a steady state power of 250 kW and a pulsing capability up to 1200 MW has been operating since March 11,1979 with an energy release of 107.5 MWh and a total of 72 pulses. During this nearly nine years, the reactor was in operation without any major undesired shut down. One of the major problems was faced when the instrumented fuel element in position 9 of the F ring went totally out of order. Secondly, the cooling tower of the secondary cooling system could not be operated properly during the hot summer days, and also we had a tar leakage problem with the radial beam port connection to the tank. During the regular maintenance work in this summer, the measurements of changes in nuclear and physical parameters of the reactor fuel and dummy elements have also proceeded. (author)

Fuel performance annual report for 1983. Volume 1

Energy Technology Data Exchange (ETDEWEB)

Bailey, W.J.; Dunenfeld, M.S.

1985-03-01

This annual report, the sixth in a series, provides a brief description of fuel performance during 1983 in commercial nuclear power plants. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience, fuel problems, high-burnup fuel experience, and items of general significance are provided. References to additional, more detailed information and related NRC evaluations are included.

A comparative study on the sooting tendencies of various 1-alkene fuels in counterflow diffusion flames

KAUST Repository

Wang, Yu

2018-02-19

Alkenes are important components in transportation fuels, and are known to have increased sooting tendencies compared to analogous saturated hydrocarbons with the same carbon number. This work aims to understand the sooting tendencies of various 1-alkenes through experiments and numerical simulations for counterflow diffusion flames. Soot and PAH formation tendencies of 1-alkene fuels, including ethylene (C2H4), propene (C3H6), 1-butene (1-C4H8), 1-pentene (1-C5H10), 1-hexene (1-C6H12) and 1-octene (1-C8H16), were experimentally studied using laser induced-incandescence (LII) and laser-induced fluorescence (LIF) techniques, respectively. From the LII results, 1-C4H8 was found to be the most sooting fuel, followed by C3H6 > 1-C5H10 > 1-C6H12 > 1-C8H16 > C2H4. The LIF data with a detection wavelength of 500 nm indicated the PAH formation tendencies followed the order of 1-C4H8 > 1-C5H10 ∼1-C6H12 > C3H6 > 1-C8H16 > C2H4, which were different from the order of sooting tendencies. Numerical simulations with a comprehensive chemical kinetic model including PAH growth chemistry for the tested 1-alkene fuels were conducted to elucidate the aromatic formation pathways and rationalize the experimentally observed trends. The numerical results highlighted the importance of intermediate species with odd carbon numbers in aromatic species formation, such as propargyl, allyl, cyclopentadienyl and indenyl radicals. Their concentration differences, which could be traced back to the parent fuel molecules through rate of production analysis, rationalize the experimentally observed differences in soot and PAH formation tendencies.

UID....Now That's Gonna Leave A Mark

Science.gov (United States)

Schramm, Harry F., Jr.

2008-01-01

Since 1975 bar codes on products at the retail counter have been accepted as the standard for entering product identity for price determination. Since the beginning of the 21 st century, the Data Matrix symbol has become accepted as the bar code format that is marked directly on a part, assembly or product that is durable enough to identify that item for its lifetime. NASA began the studies for direct part marking Data Matrix symbols on parts during the Return to Flight activities after the Challenger Accident. Over the 20 year period that has elapsed since Challenger, a mountain of studies, analyses and focused problem solutions developed by and for NASA have brought about world changing results. NASA Technical Standard 6002 and NASA Handbook 6003 for Direct Part Marking Data Matrix Symbols on Aerospace Parts have formed the basis for most other standards on part marking internationally. NASA and its commercial partners have developed numerous products and methods that addressed the difficulties of collecting part identification in aerospace operations. These products enabled the marking of Data Matrix symbols in virtually every situation and the reading of symbols at great distances, severe angles, under paint and in the dark without a light. Even unmarkable delicate parts now have a process to apply a chemical mixture, recently trademarked as Nanocodes, that can be converted to Data Matrix information through software. The accompanying intellectual property is protected by ten patents, several of which are licensed. Direct marking Data Matrix on NASA parts dramatically decreases data entry errors and the number of parts that go through their life cycle unmarked, two major threats to sound configuration management and flight safety. NASA is said to only have people and stuff with information connecting them. Data Matrix is one of the most significant improvements since Challenger to the safety and reliability of that connection.

Mark Tompkins Canaccord

OpenAIRE

Mark Tompkins Canaccord

2018-01-01

Mark Tompkins Canaccord is a senior technologist for ecosystem and water resources management in SEC SAID Oakland, California office. In his career which lasts over fifteen years Mark has worked on project involving lake restorations, clean water engineering, ecological engineering and management, hydrology, hydraulics, sediment transport and other projects for environmental planning all over the country. Mark Tompkins Canaccord tries to blend his skills of planning and engineering with s...

Physical and chemical comparison of soot in hydrocarbon and biodiesel fuel diffusion flames: A study of model and commercial fuels

Energy Technology Data Exchange (ETDEWEB)

Matti Maricq, M. [Research and Advanced Engineering, Ford Motor Company, Dearborn, MI (United States)

2011-01-15

Data are presented to compare soot formation in both surrogate and practical fatty acid methyl ester biodiesel and petroleum fuel diffusion flames. The approach here uses differential mobility analysis to follow the size distributions and electrical charge of soot particles as they evolve in the flame, and laser ablation particle mass spectrometry to elucidate their composition. Qualitatively, these soot properties exhibit a remarkably similar development along the flames. The size distributions begin as a single mode of precursor nanoparticles, evolve through a bimodal phase marking the onset of aggregate formation, and end in a self preserving mode of fractal-like particles. Both biodiesel and hydrocarbon fuels yield a common soot composition dominated by C{sub x}H{sub y}{sup +} ions, stabilomer PAHs, and fullerenes in the positive ion mass spectrum, and C{sub x}{sup -} and C{sub 2x}H{sup -} in the negative ion spectrum. These ion intensities initially grow with height in the diffusion flames, but then decline during later stages, consistent with soot carbonization. There are important quantitative differences between fuels. The surrogate biodiesel fuel methyl butanoate substantially reduces soot levels, but soot formation and evolution in this flame are delayed relative to both soy and petroleum fuels. In contrast, soots from soy and hexadecane flames exhibit nearly quantitative agreement in their size distribution and composition profiles with height, suggesting similar soot precursor chemistry. (author)

Degradation by radiation of the response of a thermocouple of a fuel element

International Nuclear Information System (INIS)

Rodriguez V, A.

1994-01-01

In the TRIGA Mark III Reactor of the National Institute of Nuclear Research, is necessary to use an instrumented fuel element for measurement the fuel temperature during pulses of power. This fuel element is exposed to daily temperature gradient of order to 390 Centigrade degrees in normal condition of reactor operation at 1 MW. The experience which this instrumented fuel elements is that useful life of the thermocouples is less then the fuel, because they show important changes in their chemistry composition and electrical specifications, until the point they don't give any response. So is necessary to know the factors that influenced in the shortening of the thermocouples life. The change in composition affects the thermocouple calibration depends on where the changes take place relative to the temperature gradient. The change will be dependent on the neutron flux and so the value of the neutron flux may be used as a measure or the composition change. If there is no neutron flux within the temperature gradient, there will be no composition change, and so the thermocouple calibration will no change. If the neutron flux varies within the region in which a temperature gradients exists, the composition of the thermocouple will vary and the calibration will change. But the maximum change in calibration will occur if the neutron flux is high and constant within the region of the temperature gradient. In this case, a composition change takes place which is uniform throughout the gradient and so the emf output can be expected to change. In this reactor, the thermocouples are in the second case. Then, the relative position of the thermal and neutron flux gradients are the most important factor that explain the composition change after or 2,500 times of exposing the thermocouples to the temperature gradients of order to 390 Centigrade degrees. (Author)

Calculations of IAEA-CRP-6 Benchmark Case 1 through 7 for a TRISO-Coated Fuel Particle

International Nuclear Information System (INIS)

Kim, Young Min; Lee, Y. W.; Chang, J. H.

2005-01-01

IAEA-CRP-6 is a coordinated research program of IAEA on Advances in HTGR fuel technology. The CRP examines aspects of HTGR fuel technology, ranging from design and fabrication to characterization, irradiation testing, performance modeling, as well as licensing and quality control issues. The benchmark section of the program treats simple analytical cases, pyrocarbon layer behavior, single TRISO-coated fuel particle behavior, and benchmark calculations of some irradiation experiments performed and planned. There are totally seventeen benchmark cases in the program. Member countries are participating in the benchmark calculations of the CRP with their own developed fuel performance analysis computer codes. Korea is also taking part in the benchmark calculations using a fuel performance analysis code, COPA (COated PArticle), which is being developed in Korea Atomic Energy Research Institute. The study shows the calculational results of IAEACRP- 6 benchmark cases 1 through 7 which describe the structural behaviors for a single fuel particle

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

International Nuclear Information System (INIS)

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

1988-12-01

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

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

International Nuclear Information System (INIS)

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

1995-01-01

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

Air/fuel ratio visualization in a diesel spray

Science.gov (United States)

Carabell, Kevin David

1993-01-01

To investigate some features of high pressure diesel spray ignition, we have applied a newly developed planar imaging system to a spray in an engine-fed combustion bomb. The bomb is designed to give flow characteristics similar to those in a direct injection diesel engine yet provide nearly unlimited optical access. A high pressure electronic unit injector system with on-line manually adjustable main and pilot injection features was used. The primary scalar of interest was the local air/fuel ratio, particularly near the spray plumes. To make this measurement quantitative, we have developed a calibration LIF technique. The development of this technique is the key contribution of this dissertation. The air/fuel ratio measurement was made using biacetyl as a seed in the air inlet to the engine. When probed by a tripled Nd:YAG laser the biacetyl fluoresces, with a signal proportional to the local biacetyl concentration. This feature of biacetyl enables the fluorescent signal to be used as as indicator of local fuel vapor concentration. The biacetyl partial pressure was carefully controlled, enabling estimates of the local concentration of air and the approximate local stoichiometry in the fuel spray. The results indicate that the image quality generated with this method is sufficient for generating air/fuel ratio contours. The processes during the ignition delay have a marked effect on ignition and the subsequent burn. These processes, vaporization and pre-flame kinetics, very much depend on the mixing of the air and fuel. This study has shown that poor mixing and over-mixing of the air and fuel will directly affect the type of ignition. An optimal mixing arrangement exists and depends on the swirl ratio in the engine, the number of holes in the fuel injector and the distribution of fuel into a pilot and main injection. If a short delay and a diffusion burn is desired, the best mixing parameters among those surveyed would be a high swirl ratio, a 4-hole nozzle and a

Fuel decontamination at Ringhals 1 with the new decontamination process IcedecTM

International Nuclear Information System (INIS)

Fredriksson, E.; Ivars, R.; Rosengren, A.; Granath, G.

2003-01-01

The new fuel decontamination technique ICEDEC TM , which has been developed by Westinghouse, is based on abrasion of fuel crud with ice particles. A mixture of ice and water is led continuously through the fuel assembly, which is placed in a specially designed fuel decontamination container connected to a closed loop recirculation system. The ice particles scrape off the loose crud from the fuel surfaces and a mixture of crud and water from the melted ice is then led to a filter unit were the crud is separated from the water. In this paper results of fuel decontamination tests of two-year-old and spent fuel assemblies during spring 2001 at Ringhals 1 are presented. The fuel crud was only released when ice particles passed through the fuel assembly and stopped within ten seconds after the feeding of ice particles had ceased. The activity release from the fuel could thus be performed in a controlled way making the process easy to manage and survey. Activity measurements confirmed that about 50% of the loose crud was removed from the fuel surfaces of the two-year-old assembly. Fuel inspection after the decontamination process showed no influence on the fuel integrity. Furthermore, no enhanced personnel radiation dose was involved with the fuel decontamination compared to normal fuel services. (authors)

Fight against fuel poverty. Levers, stakes and expectations of the fight against fuel poverty in housing

International Nuclear Information System (INIS)

Payen, Luc; Pamart, Isabelle; Lacroix, Olivier

2013-10-01

'Is in fuel poverty a person who feels in his particular housing difficulties have the necessary energy supply to the satisfaction of basic needs due to the inadequacy of resources or its habitat conditions'. The rising cost of energy commodities in the late 2000's, added to the poor thermal quality an important part of French homes, has led to the emergence of fuel poverty in the public debate. Legislative recognition of these situations with the law 'Grenelle II' (from which is extracted the definition above) marked a decisive step in the fight against this complex problem. Affecting nearly 5 million households in France, fuel poverty is a major challenge for societies wishing to successfully achieve their energy transition. In this new publication, ENEA reports on the main levers of the fight against fuel poverty, the obstacles encountered and the needs for new solutions

NDA measurements on spent fuel assemblies at Tihange 1 by means of the ION 1/FORK

International Nuclear Information System (INIS)

Carchon, R.; Smaers, G.; Verrecchia, G.P.D.; Arlt, R.; Stoyanova, I.; Satinet, J.

1986-06-01

This report describes field tests performed at Tihange 1 Nuclear Power Station on PWR spent fuel by means of the ION 1-FORK detector. Two detector systems and three electronics systems were used to investigate the same fuel assemblies with various burn-ups and cooling times. The purpose of the exercise was to test the performance of the instrument for as well inspection purposes as for fuel management. The results are presented and discussed. (Author)

Posttest examination results of recent treat tests on metal fuel

International Nuclear Information System (INIS)

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

1986-01-01

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

BWR Mark I pressure suppression study: bench mark experiments

International Nuclear Information System (INIS)

Lai, W.; McCauley, E.W.

1977-01-01

Computer simulations representative of the wetwell of Mark I BWR's have predicted pressures and related phenomena. However, calculational predictions for purposes of engineering decision will be possible only if the code can be verified, i.e., shown to compute in accord with measured values. Described in the report is a set of single downcomer spherical flask bench mark experiments designed to produce quantitative data to validate various air-water dynamic computations; the experiments were performed since relevant bench mark data were not available from outside sources. Secondary purposes of the study were to provide a test bed for the instrumentation and post-experiment data processing techniques to be used in the Laboratory's reactor safety research program and to provide additional masurements for the air-water scaling study

MARK TWAIN IN THE RUSSIAN PERIODICALS. Part 1

Directory of Open Access Journals (Sweden)

Ekaterina A. Stetsenko

2017-06-01

Full Text Available This article deals with the interpretation of the works by Mark Twain, famous Ame- This article deals with the interpretation of the works by Mark Twain, famous American author, in the Russian pre-revolutionary periodical press (1872–1916. The objects of research are critical articles, essays, reviews, correspondences, introductions to publica tions of Twain’s short stories and novels, obituaries, and other materials printed in central and provincial magazines and newspapers. Perception of Twain in Russia was contingent on many factors including political and cultural situation in the country, state of social thought and literary criticism, newspaper and magazine conjuncture etc., always remain ing polysemantic and conflicting. In different times, in the years of democratic rising or reaction critics looked for something in Twain’s works that corresponded to the spirit of their time and helped solve ideological and aesthetic problems. Twain had reputation of either a “pure humorist” or a great writer, philosopher, and moralist. Democrats, liberals, conservatives, feminists, adepts of realistic or naturalistic trends in art discussed Twain’s works that became a source of knowledge about the United States and inspired polemics about Russia’s further development. Twain was highly esteemed as the author of books for children and young people. Yet his works that criticized monarchism and imperialism were often ignored or abridged. The history of Twain’s interpretation in the Russian press serves as evidence of the fact that perception of foreign literature is a dynamic and bumpy pro cess, repeating itself and moving backwards but also getting to deeper levels of meanings.

Computer simulations of a 1/5-scale experiment of a Mark I boiler water reactor pressure-suppression system under hypothetical LOCA conditions

International Nuclear Information System (INIS)

Edwards, L.L.

1978-01-01

The CHAMP computer code was employed to simulate a plane-geometry cross section of a Mark I boiling water reactor toroidal pressure suppression system air discharge experiment under hypothetical loss-of-coolant accident conditions. The experiments were performed at the Lawrence Livermore Laboratory on a 1 / 5 -scale model of the Peach Bottom Nuclear Power Plant

A study on the thermal expansion characteristics of simulated spent fuel and simulated DUPIC fuel

International Nuclear Information System (INIS)

Kang, Kweon Ho; Ryu, H. J.; Kim, H. S.; Song, K. C.; Yang, M. S.

2001-10-01

Thermal expansions of simulated spent PWR fuel and simulated DUPIC fuel were studied using a dilatometer in the temperature range from 298 to 1900 K. The densities of simulated spent PWR fuel and simulated DUPIC fuel used in the measurement were 10.28 g/cm3 (95.35 % of TD) and 10.26 g/cm3 (95.14 % of TD), respectively. Their linear thermal expansions of simulated fuels are higher than that of UO2, and the difference between these fuels and UO2 increases progressively as temperature increases. However, the difference between simulated spent PWR fuel and simulated DUPIC fuel can hardly be observed. For the temperature range from 298 to 1900 K, the values of the average linear thermal expansion coefficients for simulated spent PWR fuel and simulated DUPIC fuel are 1.391 10-5 and 1.393 10-5 K-1, respectively. As temperature increases to 1900 K, the relative densities of simulated spent PWR fuel and simulated DUPIC fuel decrease to 93.81 and 93.76 % of initial densities at 298 K, respectively

Biodegradation of international jet A-1 aviation fuel by microorganisms isolated from aircraft tank and joint hydrant storage systems.

Science.gov (United States)

Itah, A Y; Brooks, A A; Ogar, B O; Okure, A B

2009-09-01

Microorganisms contaminating international Jet A-1 aircraft fuel and fuel preserved in Joint Hydrant Storage Tank (JHST) were isolated, characterized and identified. The isolates were Bacillus subtillis, Bacillus megaterium, Flavobacterium oderatum, Sarcina flava, Micrococcus varians, Pseudomonas aeruginosa, Bacillus licheniformis, Bacillus cereus and Bacillus brevis. Others included Candida tropicalis, Candida albicans, Saccharomyces estuari, Saccharomyces cerevisiae, Schizosaccharomyces pombe, Aspergillus flavus, Aspergillus niger, Aspergillus fumigatus, Cladosporium resinae, Penicillium citrinum and Penicillium frequentans. The viable plate count of microorganisms in the Aircraft Tank ranged from 1.3 (+/-0.01) x 104 cfu/mL to 2.2 (+/-1.6) x 104 cfu/mL for bacteria and 102 cfu/mL to 1.68 (+/-0.32) x 103 cfu/mL for fungi. Total bacterial counts of 1.79 (+/-0.2) x 104 cfu/mL to 2.58 (+/-0.04) x 104 cfu/mL and total fungal count of 2.1 (+/-0.1) x 103 cfu/mL to 2.28 (+/-0.5) x 103 cfu/mL were obtained for JHST. Selected isolates were re-inoculated into filter sterilized aircraft fuels and biodegradation studies carried out. After 14 days incubation, Cladosporium resinae exhibited the highest degradation rate with a percentage weight loss of 66 followed by Candida albicans (60.6) while Penicillium citrinum was the least degrader with a weight loss of 41.6%. The ability of the isolates to utilize the fuel as their sole source of carbon and energy was examined and found to vary in growth profile between the isolates. The results imply that aviation fuel could be biodegraded by hydrocarbonoclastic microorganisms. To avert a possible deterioration of fuel quality during storage, fuel pipe clogging and failure, engine component damage, wing tank corrosion and aircraft disaster, efficient routine monitoring of aircraft fuel systems is advocated.

Elemental marking of arthropod pests in agricultural systems: single and multigenerational marking

Science.gov (United States)

Jane Leslie Hayes

1991-01-01

Use of elemental markers to study movement of arthropod pests of field crops is reviewed. Trace elements, rubidium (Rb) and cesium (Cs), have provided a nondisruptive method of marking natural adult populations via developmental stage consumption of treated host plants. Multigenerational marking occurs with the transfer of elemental markers from marked adults to...

Study of fuel systems for LH2-fueled subsonic transport aircraft, volume 1

Science.gov (United States)

Brewer, G. D.; Morris, R. E.; Davis, G. W.; Versaw, E. F.; Cunnington, G. R., Jr.; Riple, J. C.; Baerst, C. F.; Garmong, G.

1978-01-01

Several engine concepts examined to determine a preferred design which most effectively exploits the characteristics of hydrogen fuel in aircraft tanks received major emphasis. Many candidate designs of tank structure and cryogenic insulation systems were evaluated. Designs of all major elements of the aircraft fuel system including pumps, lines, valves, regulators, and heat exchangers received attention. Selected designs of boost pumps to be mounted in the LH2 tanks, and of a high pressure pump to be mounted on the engine were defined. A final design of LH2-fueled transport aircraft was established which incorporates a preferred design of fuel system. That aircraft was then compared with a conventionally fueled counterpart designed to equivalent technology standards.

Nondestructive method for assessment of nuclear fuel

International Nuclear Information System (INIS)

Kristof, E.; Pregl, G.; Krajnik, J.; Glumac, B.; Jencic, I.; Kerzic, J.; Moskon, F.; Zitnik, F.

1983-01-01

Description of the development of the gamma spectrometry determination of an amount of a certain radioactive fission product considering local variations of the linear attenuation coefficient of gamma rays and preliminary experiment using fuel element of TRIGA Mark II reactor in Ljubljana is given.(author)

Crud formation evaluation at the advanced fuel operating in Angra-1 NPP

Energy Technology Data Exchange (ETDEWEB)

Gomez, Diego; Palheiros, Franklin; Gomes, Sydney, E-mail: franklin@inb.gov.br, E-mail: diegogomez@inb.gov.br, E-mail: sydney@inb.gov.br [Indústrias Nucleares do Brasil (INB), Resende, RJ (Brazil). Superintendência de Engenharia do Combustível

2017-07-01

In nuclear engineering, 'crud' is a technical term. It stands for Chalk River Unidentified Deposit, originally found on the cladding surface of some fuel rods in the referred canadian reactor, for which it was named. The deposit can be flaky, porous, or hard depending on its chemical composition. In most cases, it reduces the power output of nuclear reactors - the deposits absorb boron and the neutrons that keep the fission reaction going, as well lead to a more corrosion scenario by increasing the oxide/metal interface surface temperature. This issue might been a concern at Angra 1 where many design alterations have been performed in the new Fuel assembly design. The so called 16NGF has a smaller fuel rod diameter, different burnable absorber - gadolinium instead of pyrex borosilicate glass, hydraulic mismatch compared to 16STD fuel, new IFM grids, higher FDeltaH and several other characteristics. All those features lead to a increase in the subcooled boiling rates, which might favour particles depositions in fuel cladding forming the undesired Crud deposits. In order to evaluate how those implementations could impact negatively the new fuel performance at Angra 1, a study has ben carried out using Thermal Hydraulic calculations. With that, an existing methodology was used to assess the associated risks and what could be the done to mitigate further development of crud in 16NGF Fuel in Angra 1. (author)

40 CFR 80.550 - What is the definition of a motor vehicle diesel fuel small refiner or a NRLM diesel fuel small...

Science.gov (United States)

2010-07-01

...) REGULATION OF FUELS AND FUEL ADDITIVES Motor Vehicle Diesel Fuel; Nonroad, Locomotive, and Marine Diesel Fuel... vehicle diesel fuel small refiner or a NRLM diesel fuel small refiner under this subpart? (a) A motor...-operational between January 1, 1999, and January 1, 2000, may apply for motor vehicle diesel fuel small...

Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

Energy Technology Data Exchange (ETDEWEB)

John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

2011-03-01

The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

Energy Technology Data Exchange (ETDEWEB)

Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Maddock, Thomas L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Marshall, Margaret A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Ning [Idaho National Lab. (INL), Idaho Falls, ID (United States); Phillips, Ann Marie [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schreck, Kenneth A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Briggs, J. Blair [Idaho National Lab. (INL), Idaho Falls, ID (United States); Woolstenhulme, Eric W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bolin, John M. [General Atomics, San Diego, CA (United States); Veca, Anthony [General Atomics, San Diego, CA (United States); McKnight, Richard D. [Argonne National Lab. (ANL), Argonne, IL (United States); Lell, Richard M. [Argonne National Lab. (ANL), Argonne, IL (United States)

2014-03-01

The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

Testing of a De Nora polymer electrolyte fuel cell stack of 1 kW for naval applications

Science.gov (United States)

Schmal, D.; Kluiters, C. E.; Barendregt, I. P.

In a previous study calculations were carried out for a navy frigate with respect to the energy consumption of a propulsion/electricity generation system based on fuel cells. The fuel consumption for the 'all-fuel cell' ship was compared with the consumption of the current propulsion/electricity generation system based on gas turbines and diesel engines; it showed potential energy savings of a fuel cell based system amounting from 25 to 30%. On the basis of these results and taking into account various military aspects it was decided to start tests with a polymer electrolyte fuel cell (PEFC) stack. For this purpose a De Nora 1 kW PEFC was chosen. Results of the first tests after installation are satisfying.

Thermal Hydraulics Analysis for the 3MW TRIGA MARK-II Research Reactor Under Transient Condition

International Nuclear Information System (INIS)

Huda, M.Q.; Bhuiyan, S.I.; Mondal, M.A.W.

1996-12-01

Some important thermal hydraulic parameters of the 3 MW TRIGA MARK-II research reactor operating under transient condition were investigated using two computer codes PULTRI and TEMPUL. Major transient parameters, such as, peak power and prompt energy released after pulse, maximum fuel and coolant temperature, surface heat flux, time and radial distribution of temperature within fuel element after pulse, fuel, fuel-cladding gap width variation, etc. were computer and compared with the experimental and operational values as reported in the safety Analysis Report (SAR). It was observed that pulsing of the reactor inserting an excess reactivity of $2.00 shoots the reactor power level to 854.353 MW compared to an experimental value of 852 MW; the maximum fuel temperature corresponding to this peak power was found to be 846.76 o C which is much less than the limiting maximum value of fuel temperature of 1150 0 C as reported in SAR. During a pulse if the film boiling occurs for a peak adiabatic fuel temperature of 1000 o C, the calculated outer cladding wall temperature was observed to be 702.39 0 C compared to a value of 760 o C reported in SAR under the same condition. The investigated other results were also found to be in good agreement with the values reported in the SAR. 16 refs., 22 figs. (author)

Activation calculation of steel of the control rods of TRIGA Mark III reactor

International Nuclear Information System (INIS)

Garcia M, T.; Cruz G, H. S.; Ruiz C, M. A.; Angeles C, A.

2014-10-01

In the pool of TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ), there are control rods that were removed from the core, and which are currently on shelves of decay. These rods were part of the reactor core when only had fuel standard (from 1968-1989). To conduct a proper activation analysis of the rods, is very important to have well-characterized the materials which are built, elemental composition of the same ones, the atomic densities and weight fractions of the elements that constitute them. To determine the neutron activation of the control rods MCNP5 code was used, this code allows us to have well characterized the radionuclides inventory that were formed during irradiation of the control rods. This work is limited to determining the activation of the steel that is part of the shielding of the control rods, the nuclear fuel that is in the fuel follower does not include. The calculation model of the code will be validated with experimental measurements and calculating the activity of fission products of the fuel follower which will take place at the end of 2014. (Author)

Fission product monitoring of TRISO coated fuel for the advanced gas reactor-1 experiment

International Nuclear Information System (INIS)

Scates, Dawn M.; Hartwell, John K.; Walter, John B.; Drigert, Mark W.; Harp, Jason M.

2010-01-01

The US Department of Energy has embarked on a series of tests of TRISO coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burnup of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B's) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

Efficient computation of the inverse of gametic relationship matrix for a marked QTL

Directory of Open Access Journals (Sweden)

Iwaisaki Hiroaki

2006-04-01

Full Text Available Abstract Best linear unbiased prediction of genetic merits for a marked quantitative trait locus (QTL using mixed model methodology includes the inverse of conditional gametic relationship matrix (G-1 for a marked QTL. When accounting for inbreeding, the conditional gametic relationships between two parents of individuals for a marked QTL are necessary to build G-1 directly. Up to now, the tabular method and its adaptations have been used to compute these relationships. In the present paper, an indirect method was implemented at the gametic level to compute these few relationships. Simulation results showed that the indirect method can perform faster with significantly less storage requirements than adaptation of the tabular method. The efficiency of the indirect method was mainly due to the use of the sparseness of G-1. The indirect method can also be applied to construct an approximate G-1 for populations with incomplete marker data, providing approximate probabilities of descent for QTL alleles for individuals with incomplete marker data.

Neutronics analysis of TRIGA Mark II research reactor

Directory of Open Access Journals (Sweden)

Haseebur Rehman

2018-02-01

Full Text Available This article presents clean core criticality calculations and control rod worth calculations for TRIGA (Training, Research, Isotope production-General Atomics Mark II research reactor benchmark cores using Winfrith Improved Multi-group Scheme-D/4 (WIMS-D/4 and Program for Reactor In-core Analysis using Diffusion Equation (PRIDE codes. Cores 133 and 134 were analyzed in 2-D (r, θ and 3-D (r, θ, z, using WIMS-D/4 and PRIDE codes. Moreover, the influence of cross-section data was also studied using various libraries based on Evaluated Nuclear Data File (ENDF/B-VI.8 and VII.0, Joint Evaluated Fission and Fusion File (JEFF-3.1, Japanese Evaluated Nuclear Data Library (JENDL-3.2, and Joint Evaluated File (JEF-2.2 nuclear data. The simulation results showed that the multiplication factor calculated for all these data libraries is within 1% of the experimental results. The reactivity worth of the control rods of core 134 was also calculated with different homogenization approaches. A comparison was made with experimental and reported Monte Carlo results, and it was found that, using proper homogenization of absorber regions and surrounding fuel regions, the results obtained with PRIDE code are significantly improved.

Lujan Mark-4

Energy Technology Data Exchange (ETDEWEB)

Mocko, Michael Jeffrey [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Zavorka, Lukas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Koehler, Paul E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-11-13

This is a review of Mark-IV target neutronics design. It involved the major redesign of the upper tier, offering harder neutron spectra for upper-tier FPs; a redesign of the high-resolution (HR) moderator; and a preservation of the rest of Mark-III features.

Bio fuels and family farming in Uruguay: A feasible alliance?

International Nuclear Information System (INIS)

Carambula, M.; Chiappe, M.; Fernandez, E.; Figueredo, S.

2011-01-01

The global energy crisis caused by high levels of fossil fuels consumption and the signs of oil depletion explain the search for alternative energy to traditional sources. Progress towards bio-fuels policy is positioned in a central place in Uruguay s political agenda. This context converges with a scenario of expansion of agricultural activity, marked by a dynamism based on the domestic economic environment changes, and major transformations in the productive base. In this context, in order to assess the social impacts resulting from the expansion of crops for energy purposes, this research was carried out. It explores the social impact of bio fuels production in Uruguay taking as a reference the situation of family farm production. It assumes that the demand of land for energy crop production puts pressure on other production systems. Related to this, it is possible to establish a continuum between a view that holds that family farms are marginal to bio fuel production, and an inclusive view which encourages the incorporation of family farmers into national production chains. In this scenario, the paper attempts to provide elements to answer the question about whether this new line of national production generates opportunities or threats to family farming

Fuel Exhaling Fuel Cell.

Science.gov (United States)

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.

Distinguishing butchery cut marks from crocodile bite marks through machine learning methods.

Science.gov (United States)

Domínguez-Rodrigo, Manuel; Baquedano, Enrique

2018-04-10

All models of evolution of human behaviour depend on the correct identification and interpretation of bone surface modifications (BSM) on archaeofaunal assemblages. Crucial evolutionary features, such as the origin of stone tool use, meat-eating, food-sharing, cooperation and sociality can only be addressed through confident identification and interpretation of BSM, and more specifically, cut marks. Recently, it has been argued that linear marks with the same properties as cut marks can be created by crocodiles, thereby questioning whether secure cut mark identifications can be made in the Early Pleistocene fossil record. Powerful classification methods based on multivariate statistics and machine learning (ML) algorithms have previously successfully discriminated cut marks from most other potentially confounding BSM. However, crocodile-made marks were marginal to or played no role in these comparative analyses. Here, for the first time, we apply state-of-the-art ML methods on crocodile linear BSM and experimental butchery cut marks, showing that the combination of multivariate taphonomy and ML methods provides accurate identification of BSM, including cut and crocodile bite marks. This enables empirically-supported hominin behavioural modelling, provided that these methods are applied to fossil assemblages.

Nuclear-fuel-cycle education: Module 1. Nuclear fuel cycle overview

International Nuclear Information System (INIS)

Eckhoff, N.D.

1981-07-01

This educational module is an overview of the nuclear-fule-cycle. The overview covers nuclear energy resources, the present and future US nuclear industry, the industry view of nuclear power, the International Nuclear Fuel Cycle Evaluation program, the Union of Concerned Scientists view of the nuclear-fuel-cycle, an analysis of this viewpoint, resource requirements for a model light water reactor, and world nuclear power considerations

A report on the transport of MTR-type spent fuel assemblies of the Philippine Research Reactor (PRR-1)

International Nuclear Information System (INIS)

Yoshisaki, Magno B.; Leopando, Leonardo S.

1999-03-01

Fifty one (51) fuel assemblies of mixed enrichment from the Philippine Research Reactor (PRR-1), consisting of 50 spent and 1 fresh, were shipped to the United States last 14 March 1999 under the U.S. Return of Foreign Research Reactor (FRR) fuel policy. The shipment was in line with the U.S. initiative to implement its Record of Decision (ROD) which took effect on 13 May 1996 to accept and manage all FRR uranium fuel of U.S. origin and enriched in the United States. The shipment program would last10 years, ending midnight of 13 May 2006. The ROD provided a 3 year extension period within which to accept FRR spent nuclear fuel (SNF) withdrawn from reactors after 2006. The U.S. policy gave priority to the NPT significance of high enriched U, as the prime target of the return of FRR policy. Classified as a developing country, the Philippines, through the PNRI, signed a contract with the U.S. Department of Energy for the cost-free shipment of PRR-1 spent fuel to the United States. Spent fuel loading and transport operations to the port area lasted seven (7) days, from 8 to 14 March 1999. (Author)

The Information Literacy of Survey Mark Hunting: A Dialogue

Directory of Open Access Journals (Sweden)

Jennifer Galas

2016-11-01

Full Text Available In Brief: This article makes connections between the ACRL Framework for Information Literacy for Higher Education and the activity of survey mark hunting. After a brief review of the literature related to geographic information systems (GIS, information literacy, and gamification of learning, the authors enter into a dialogue in which they discover and describe the various ways information literacy is both required by and developed through the recreational activity of survey mark hunting. Through their dialogue they found that the activity of survey mark hunting relies on the construction of both information and its authority in ways contextualized within the communities that participate; that survey mark hunting is a conversation that builds on the past, where lived experience counts as evidence; and, that survey mark hunting is both a metaphor and embodied enactment of information literacy.

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

International Nuclear Information System (INIS)

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

1993-09-01

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

The mismatch repair protein MLH1 marks a subset of strongly interfering crossovers in tomato

NARCIS (Netherlands)

Lhuissier, F.G.P.; Offenberg, H.H.; Wittich, P.E.; Vischer, N.O.E.; Heyting, C.

2007-01-01

In most eukaryotes, the prospective chromosomal positions of meiotic crossovers are marked during meiotic prophase by protein complexes called late recombination nodules (LNs). In tomato (Solanum lycopersicum), a cytological recombination map has been constructed based on LN positions. We

A conceptual model for the fuel oxidation of defective fuel

International Nuclear Information System (INIS)

Higgs, J.D.; Lewis, B.J.; Thompson, W.T.; He, Z.

2007-01-01

A mechanistic conceptual model has been developed to predict the fuel oxidation behaviour in operating defective fuel elements for water-cooled nuclear reactors. This theoretical work accounts for gas-phase transport and sheath reactions in the fuel-to-sheath gap to determine the local oxygen potential. An improved thermodynamic analysis has also been incorporated into the model to describe the equilibrium state of the oxidized fuel. The fuel oxidation kinetics treatment accounts for multi-phase transport including normal diffusion and thermodiffusion for interstitial oxygen migration in the solid, as well as gas-phase transport in the fuel pellet cracks. The fuel oxidation treatment is further coupled to a heat conduction equation. A numerical solution of the coupled transport equations is obtained by a finite-element technique with the FEMLAB 3.1 software package. The model is able to provide radial-axial profiles of the oxygen-to-uranium ratio and the fuel temperatures as a function of time in the defective element for a wide range of element powers and defect sizes. The model results are assessed against coulometric titration measurements of the oxygen-to-metal profile for pellet samples taken from ten spent defective elements discharged from the National Research Universal Reactor at the Chalk River Laboratories and commercial reactors

A Raman-Based Portable Fuel Analyzer

Science.gov (United States)

Farquharson, Stuart

2010-08-01

Fuel is the single most import supply during war. Consider that the US Military is employing over 25,000 vehicles in Iraq and Afghanistan. Most fuel is obtained locally, and must be characterized to ensure proper operation of these vehicles. Fuel properties are currently determined using a deployed chemical laboratory. Unfortunately, each sample requires in excess of 6 hours to characterize. To overcome this limitation, we have developed a portable fuel analyzer capable of determine 7 fuel properties that allow determining fuel usage. The analyzer uses Raman spectroscopy to measure the fuel samples without preparation in 2 minutes. The challenge, however, is that as distilled fractions of crude oil, all fuels are composed of hundreds of hydrocarbon components that boil at similar temperatures, and performance properties can not be simply correlated to a single component, and certainly not to specific Raman peaks. To meet this challenge, we measured over 800 diesel and jet fuels from around the world and used chemometrics to correlate the Raman spectra to fuel properties. Critical to the success of this approach is laser excitation at 1064 nm to avoid fluorescence interference (many fuels fluoresce) and a rugged interferometer that provides 0.1 cm-1 wavenumber (x-axis) accuracy to guarantee accurate correlations. Here we describe the portable fuel analyzer, the chemometric models, and the successful determination of these 7 fuel properties for over 100 unknown samples provided by the US Marine Corps, US Navy, and US Army.

RIA Fuel Codes Benchmark - Volume 1

International Nuclear Information System (INIS)

Marchand, Olivier; Georgenthum, Vincent; Petit, Marc; Udagawa, Yutaka; Nagase, Fumihisa; Sugiyama, Tomoyuki; Arffman, Asko; Cherubini, Marco; Dostal, Martin; Klouzal, Jan; Geelhood, Kenneth; Gorzel, Andreas; Holt, Lars; Jernkvist, Lars Olof; Khvostov, Grigori; Maertens, Dietmar; Spykman, Gerold; Nakajima, Tetsuo; Nechaeva, Olga; Panka, Istvan; Rey Gayo, Jose M.; Sagrado Garcia, Inmaculada C.; Shin, An-Dong; Sonnenburg, Heinz Guenther; Umidova, Zeynab; Zhang, Jinzhao; Voglewede, John

2013-01-01

Reactivity-initiated accident (RIA) fuel rod codes have been developed for a significant period of time and they all have shown their ability to reproduce some experimental results with a certain degree of adequacy. However, they sometimes rely on different specific modelling assumptions the influence of which on the final results of the calculations is difficult to evaluate. The NEA Working Group on Fuel Safety (WGFS) is tasked with advancing the understanding of fuel safety issues by assessing the technical basis for current safety criteria and their applicability to high burnup and to new fuel designs and materials. The group aims at facilitating international convergence in this area, including the review of experimental approaches as well as the interpretation and use of experimental data relevant for safety. As a contribution to this task, WGFS conducted a RIA code benchmark based on RIA tests performed in the Nuclear Safety Research Reactor in Tokai, Japan and tests performed or planned in CABRI reactor in Cadarache, France. Emphasis was on assessment of different modelling options for RIA fuel rod codes in terms of reproducing experimental results as well as extrapolating to typical reactor conditions. This report provides a summary of the results of this task. (authors)

Consolidated fuel reprocessing. Program progress report, April 1-June 30, 1980

Energy Technology Data Exchange (ETDEWEB)

1980-09-01

This progress report is compiled from major contributions from three programs: (1) the Advanced Fuel Recycle Program at ORNL; (2) the Converter Fuel Reprocessing Program at Savannah River Laboratory; and (3) the reprocessing components of the HTGR Fuel Recycle Program, primarily at General Atomic and ORNL. The coverage is generally overview in nature; experimental details and data are limited.

The KNK II/1 fuel assembly NY-205: Compilation of the irradiation history and the fuel and fuel pin fabrication data of the INTERATOM data bank system BESEX

International Nuclear Information System (INIS)

Patzer, G.; Geier, F.

1988-01-01

The fuel assembly NY-205 has been irradiated during the first and the second core of KNK II with a total residence time of 832 equivalent full-power days. A maximum burnup of 175.000 MWd/tHM or 18.6 % was reached with a maximum steel damage of 66 dpa-NRT. For the cladding the materials 1.4970 and 1.4981 have been used in different metallurgical conditions, and for the Uranium/Plutonium mixed- oxide fuel the most important variants of the major fabrication parameters had been realized. The assembly will be brought to the Hot Cells of the KfK Karlsruhe for post-irradiation examination in February 1988, so that the knowledge of the fabrication data is of interest for the selection of fuel pins and for the evaluation of the examination results. Therefore this report compiles the fuel and fuel pin fabrication data from the INTERATOM data bank system BESEX and additionally, an overview of the irradiation history of the assembly is given [de

Analysis of neutronic parameters related to reduction in fuel rod diameter for Angra-1 reactor fuel elements

International Nuclear Information System (INIS)

Faria, Eduardo F.; Sadde, Luciano M.; Sakai, Massao; Gomes, Sydney da S.

2000-01-01

The actual fuel element design for Angra-1 PWR satisfies in a very conservative way the design limits established for the critical heat flux as well as for the energy stored in the fuel rod. However, that is not an optimized design under neutronic considerations. The conservative ratio of the H and U atomic densities gives rise to a harder neutron spectrum which reduces its reactivity. In this report, a reduction in fuel rod diameters has been analyzed, keeping however the same rod pitch for geometrical compatibility reasons. By increasing the H/U ratio it is possible to obtain a net gain in reactivity. The optimized diameter in its turn should not jeopardize the reactor safety requirements. The actual trends of the nuclear industry is to extend the cycles and the enrichment by using advanced fuel design. It must be emphasized that this design change gives rise to economical advantages, for example, reduced costs for uranium utilization and enrichment with a net gain in reactivity. (author)

Test of high temperature fuel element, (1)

International Nuclear Information System (INIS)

Akino, Norio; Shiina, Yasuaki; Nekoya, Shin-ichi; Takizuka, Takakazu; Emori, Koichi

1980-11-01

Heat transfer experiment to measure the characteristics of a VHTR fuel in the same condition of the reactor core was carried out using HTGL (High Temperature Helium Gas Loop) and its test section. In this report, the details of the test section, related problems of construction and some typical results are described. The newly developed heater with graphite heat transfer surface was used as a simulated fuel element to determine the heat transfer characteristics. Following conclusions were obtained; (1) Reynolds number between turbulent and transitional region is about 2600. (2) Reynolds number between transitional and laminar region is about 4800. (3) The laminarization phenomena have not been observed and are hardly occurred in annular tubes comparing with round tube. (4) Measured Nusselt numbers agree to the established correlations in turbulent and laminar regions. (author)

Experience in producing LEU fuel elements for the RSG-GAS

International Nuclear Information System (INIS)

Suripto, A.; Soentono, S.

1991-01-01

To achieve a self-reliance in the operation of the 30 MW Multipurpose Research Reactor at Serpong (the RSG-GAS), a fuel element production facility has been constructed nearby. The main task of the facility is to produce MTR type fuel and control elements containing U 3 O 8 -Al dispersion LEU fuel for the RSG-GAS. The hot commissioning activity has started in early 1988 after completion of the cold commissioning using depleted uranium in 1987, marking the beginning of the real production activity. This paper briefly describes the main features of the fuel production facility, the production experience gained so far, and its current production activity. (orig.)

Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

Energy Technology Data Exchange (ETDEWEB)

John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

2013-03-01

The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has been evaluated as an acceptable benchmark experiment. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

A method for studying knife tool marks on bone.

Science.gov (United States)

Shaw, Kai-Ping; Chung, Ju-Hui; Chung, Fang-Chun; Tseng, Bo-Yuan; Pan, Chih-Hsin; Yang, Kai-Ting; Yang, Chun-Pang

2011-07-01

The characteristics of knife tool marks retained on hard tissues can be used to outline the shape and angle of a knife. The purpose of this study was to describe such marks on bone tissues that had been chopped with knives. A chopping stage with a gravity accelerator and a fixed bone platform was designed to reconstruct the chopping action. A digital microscope was also used to measure the knife angle (θ) and retained V-shape tool mark angle (ψ) in a pig skull. The κ value (elasticity coefficient; θ/ψ) was derived and recorded after the knife angle (θ) and the accompanied velocity were compared with the proportional impulsive force of the knife and ψ on the bone. The constant impulsive force revealed a correlation between the V-shape tool mark angle (ψ) and the elasticity coefficient (κ). These results describe the tool marks--crucial in the medicolegal investigation--of a knife on hard tissues. © 2011 American Academy of Forensic Sciences.

LIFE Materials: Overview of Fuels and Structural Materials Issues Volume 1

Energy Technology Data Exchange (ETDEWEB)

Farmer, J

2008-09-08

The National Ignition Facility (NIF) project, a laser-based Inertial Confinement Fusion (ICF) experiment designed to achieve thermonuclear fusion ignition and burn in the laboratory, is under construction at the Lawrence Livermore National Laboratory (LLNL) and will be completed in April of 2009. Experiments designed to accomplish the NIF's goal will commence in late FY2010 utilizing laser energies of 1 to 1.3 MJ. Fusion yields of the order of 10 to 20 MJ are expected soon thereafter. Laser initiated fusion-fission (LIFE) engines have now been designed to produce nuclear power from natural or depleted uranium without isotopic enrichment, and from spent nuclear fuel from light water reactors without chemical separation into weapons-attractive actinide streams. A point-source of high-energy neutrons produced by laser-generated, thermonuclear fusion within a target is used to achieve ultra-deep burn-up of the fertile or fissile fuel in a sub-critical fission blanket. Fertile fuels including depleted uranium (DU), natural uranium (NatU), spent nuclear fuel (SNF), and thorium (Th) can be used. Fissile fuels such as low-enrichment uranium (LEU), excess weapons plutonium (WG-Pu), and excess highly-enriched uranium (HEU) may be used as well. Based upon preliminary analyses, it is believed that LIFE could help meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the nation's and world's stockpile of spent nuclear fuel and excess weapons materials. LIFE takes advantage of the significant advances in laser-based inertial confinement fusion that are taking place at the NIF at LLNL where it is expected that thermonuclear ignition will be achieved in the 2010-2011 timeframe. Starting from as little as 300 to 500 MW of fusion power, a single LIFE engine will be able to generate 2000 to 3000 MWt in steady state for periods of years to decades, depending on the nuclear fuel and engine configuration. Because the fission

Transient behavior during reactivity insertion in the Moroccan TRIGA Mark II reactor using the PARET/ANL code

International Nuclear Information System (INIS)

Boulaich, Y.; Nacir, B.; El Bardouni, T.; Boukhal, H.; Chakir, E.; El Bakkari, B.; El Younoussi, C.

2015-01-01

Highlights: • PARET model for the Moroccan TRIGA MARK II reactor has been developed. • Transient behavior under reactivity insertion has been studied based on PARET code. • Power factors required by PARET code have been calculated by using MCNP5 code. • The dependence on time of the main thermal-hydraulic parameters was calculated. • Results are largely far to compromise the thermal design limits. - Abstract: A three dimensional model for the Moroccan 2 MW TRIGA MARK II reactor has been developed for thermal-hydraulic and safety analysis by using the PARET/ANL and MCNP5 codes. This reactor is located at the nuclear studies center of Mâamora (CENM), Morocco. The model has been validated through temperature measurements inside two instrumented fuel elements located near the center of the core, at various power levels, and also through the power and fuel temperature evolution after the reactor shutdown (SCRAM). The axial distributions of power factors required by the PARET code have been calculated in each fuel element rod by using MCNP5 code. Based on this thermal-hydraulic model, a safety analysis under the reactivity insertion phenomenon has been carried out and the dependence on time of the main thermal-hydraulic parameters was calculated. Results were compared to the thermal design limits imposed to maintain the integrity of the clad

Combustion characteristics of a turbocharged DI compression ignition engine fueled wth petroleum diesel fuels and biodiesel

Energy Technology Data Exchange (ETDEWEB)

Canakci, M. [Kocaeli University, Izmit (Turkey). Department of Mechanical Education

2007-04-15

In this study, the combustion characteristics and emissions of two different petroleum diesel fuels (No. 1 and No. 2) and biodiesel from soybean oil were compared. The tests were performed at steady state conditions in a four-cylinder turbocharged DI diesel engine at full load at 1400-rpm engine speed. The experimental results compared with No. 2 diesel fuel showed that biodiesel provided significant reductions in PM, CO, and unburned HC, the NO{sub x} increased by 11.2%. Biodiesel had a 13.8% increase in brake-specific fuel consumption due to its lower heating value. However, using No. 1 diesel fuel gave better emission results, NO{sub x} and brake-specific fuel consumption reduced by 16.1% and 1.2%, respectively. The values of the principal combustion characteristics of the biodiesel were obtained between two petroleum diesel fuels. The results indicated that biodiesel may be blended with No. 1 diesel fuel to be used without any modification on the engine. (author)

GREET 1.5 - transportation fuel-cycle model - Vol. 1 : methodology, development, use, and results

International Nuclear Information System (INIS)

Wang, M. Q.

1999-01-01

This report documents the development and use of the most recent version (Version 1.5) of the Greenhouse Gases, Regulated Emissions, and Energy Use in Transportation (GREET) model. The model, developed in a spreadsheet format, estimates the full fuel-cycle emissions and energy associated with various transportation fuels and advanced vehicle technologies for light-duty vehicles. The model calculates fuel-cycle emissions of five criteria pollutants (volatile organic compounds, carbon monoxide, nitrogen oxides, particulate matter with diameters of 10 micrometers or less, and sulfur oxides) and three greenhouse gases (carbon dioxide, methane, and nitrous oxide). The model also calculates total energy consumption, fossil fuel consumption, and petroleum consumption when various transportation fuels are used. The GREET model includes the following cycles: petroleum to conventional gasoline, reformulated gasoline, conventional diesel, reformulated diesel, liquefied petroleum gas, and electricity via residual oil; natural gas to compressed natural gas, liquefied natural gas, liquefied petroleum gas, methanol, Fischer-Tropsch diesel, dimethyl ether, hydrogen, and electricity; coal to electricity; uranium to electricity; renewable energy (hydropower, solar energy, and wind) to electricity; corn, woody biomass, and herbaceous biomass to ethanol; soybeans to biodiesel; flared gas to methanol, dimethyl ether, and Fischer-Tropsch diesel; and landfill gases to methanol. This report also presents the results of the analysis of fuel-cycle energy use and emissions associated with alternative transportation fuels and advanced vehicle technologies to be applied to passenger cars and light-duty trucks

User’s guide for MapMark4GUI—A graphical user interface for the MapMark4 R package

Science.gov (United States)

Shapiro, Jason

2018-05-29

MapMark4GUI is an R graphical user interface (GUI) developed by the U.S. Geological Survey to support user implementation of the MapMark4 R statistical software package. MapMark4 was developed by the U.S. Geological Survey to implement probability calculations for simulating undiscovered mineral resources in quantitative mineral resource assessments. The GUI provides an easy-to-use tool to input data, run simulations, and format output results for the MapMark4 package. The GUI is written and accessed in the R statistical programming language. This user’s guide includes instructions on installing and running MapMark4GUI and descriptions of the statistical output processes, output files, and test data files.

Build-up of actinides in irradiated fuel rods of the ET-RR-1 reactor

Energy Technology Data Exchange (ETDEWEB)

Adib, M.; Naguib, K.; Morcos, H.N

2001-09-01

The content concentrations of actinides are calculated as a function of operating reactor regime and cooling time at different percentage of fuel burn-up. The build-up transmutation equations of actinides content in an irradiated fuel are solved numerically .A computer code BAC was written to operate on a PC computer to provide the required calculations. The fuel element of 10% {sup 235}U enrichment of ET-RR-1 reactor was taken as an example for calculations using the BAC code. The results are compared with other calculations for the ET-RR-1 fuel rod. An estimation of fissile build-up content of a proposed new fuel of 20% {sup 235}U enrichment for ET-RR-1 reactor is given. The sensitivity coefficients of build-up plutonium concentrations as a function of cross-section data uncertainties are also calculated.

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

International Nuclear Information System (INIS)

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

Pollutant emissions from gasoline combustion. 1. Dependence on fuel structural functionalities.

Science.gov (United States)

Zhang, Hongzhi R; Eddings, Eric G; Sarofim, Adel F

2008-08-01

To study the formation of air pollutants and soot precursors (e.g., acetylene, 1,3-butadiene, benzene, and higher aromatics) from aliphatic and aromatic fractions of gasoline fuels, the Utah Surrogate Mechanisms is extended to include submechanisms of gasoline surrogate compounds using a set of mechanism generation techniques. The mechanism yields very good predictions of species concentrations in premixed flames of n-heptane, isooctane, benzene, cyclohexane, olefins, oxygenates, and gasoline using a 23-component surrogate formulation. The 1,3-butadiene emission comes mainly from minor fuel fractions of olefins and cyclohexane. The benzene formation potential of gasoline components shows the following trends as functions of (i) chemical class: n-paraffins produced by the real fuel should have priority when selecting candidate surrogate components for combustion simulations.

Selection of fuel design for conversion and upgradation of Pakistan Research Reactor (PARR-1)

International Nuclear Information System (INIS)

Arshad, M.

1991-01-01

The Pakistan Research Reactor (PARR-1) is being converted from the use of Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU) fuel and its power is also being upgraded. In order to select new fuel for the converted and upgraded core ten different fuel element designs were analyzed and their relative performance was compared. Results of this study were later used to select appropriate design of the new fuel for PARR-1. This paper describes the computational methodology utilized for the analysis of various fuel element designs. Criteria for selecting the new fuel element are discussed and guidelines forming the selection basis of the new fuel design are given. (author)

Effects of piston surface treatments on performance and emissions of a methanol-fueled, direct injection, stratified charge engine

Energy Technology Data Exchange (ETDEWEB)

West, B.; Green, J.B. [Oak Ridge National Lab., TN (United States)

1994-07-01

The purpose of this study was to investigate the effects of thermal barrier coatings and/or surface treatments on the performance and emissions of a methanol-fueled, direct-injection, stratified-charge (DISC) engine. A Ricardo Hydra Mark III engine was used for this work and in previous experiments at Oak Ridge National Laboratory (ORNL). The primary focus of the study was to examine the effects of various piston insert surface treatments on hydrocarbon (HC) and oxides of nitrogen (NO{sub x}) emissions. Previous studies have shown that engines of this class have a tendency to perform poorly at low loads and have high unburned fuel emissions. A blank aluminum piston was modified to employ removable piston bowl inserts. Four different inserts were tested in the experiment: aluminum, stainless steel with a 1.27-mm (0.050-in.) air gap (to act as a thermal barrier), and two stainless steel/air-gap inserts with coatings. Two stainless steel inserts were dimensionally modified to account for the coating thickness (1.27-mm) and coated identically with partially stabilized zirconia (PSZ). One of the coated inserts then had an additional seal-coat applied. The coated inserts were otherwise identical to the stainless steel/air-gap insert (i.e., they employed the same 1.27-mm air gap). Thermal barrier coatings were employed in an attempt to increase combustion chamber surface temperatures, thereby reducing wall quenching and promoting more complete combustion of the fuel in the quench zone. The seal-coat was applied to the zirconia to reduce the surface porosity; previous research suggested that despite the possibly higher surface temperatures obtainable with a ceramic coating, the high surface area of a plasma-sprayed coating may actually allow fuel to adhere to the surface and increase the unburned fuel emissions and fuel consumption.

Capture programs, analysis, data graphication for the study of the thermometry of the TRIGA Mark III reactor core

International Nuclear Information System (INIS)

Paredes G, L.C.

1991-05-01

This document covers the explanation of the capture programs, analysis and graphs of the data obtained during the measurement of the temperatures of the instrumented fuel element of the TRIGA Mark III reactor and of the coolant one near to this fuel, using the conversion card from Analogic to Digital of 'Data Translation', and using a signal conditioner for five temperature measurers with the help of thermo par type K, developed by the Simulation and Control of the nuclear systems management department, which gives a signal from 0 to 10 Vcd for an interval of temperature of 0 to 1000 C. (Author)

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

International Nuclear Information System (INIS)

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

1977-08-01

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

Using road markings as a continuous cue for speed choice.

Science.gov (United States)

Charlton, Samuel G; Starkey, Nicola J; Malhotra, Neha

2018-08-01

The potential for using road markings to indicate speed limits was investigated in a driving simulator over the course of two sessions. Two types of experimental road markings, an "Attentional" set designed to provide visually distinct cues to indicate speed limits of 60, 80 and 100 km/h, and a "Perceptual" set designed to also affect drivers' perception of speed, were compared to a standard undifferentiated set of markings. Participants (n = 20 per group) were assigned to one of four experimental groups (Attentional-Explicit, Attentional-Implicit, Perceptual-Explicit, Perceptual-Implicit) or a Control group (n = 22; standard road markings). The Explicit groups were instructed about the meaning of the road markings while those in the Implicit and Control groups did not receive any explanation. Participants drove five 10 km simulated roads containing three speed zones (60, 80 and 100 km/h) during the first session. The participants returned to the laboratory approximately 3 days later to drive five more trials including roads they had not seen before, a trial that included a secondary task, and a trial where speed signs were removed and only markings were present. The findings indicated that both types of road markings improved drivers' compliance with speed limits compared to the control group, but that explicit instruction as to the meaning of the markings was needed to realise their full benefit. Although previous research has indicated the benefit of road markings used as warnings to indicate speed reductions in advance of horizontal or vertical curves, the findings of the present experiment also suggest that systematically associating road markings with specific speed limits may be a useful way to improve speed limit compliance and increase speed homogeneity. Copyright © 2018 The Authors. Published by Elsevier Ltd.. All rights reserved.

Gel-sphere-pac reactor fuel fabrication and its application to a variety of fuels

International Nuclear Information System (INIS)

Olsen, A.R.; Judkins, R.R.

1979-12-01

The gel-sphere-pac fuel fabrication option was evaluated for its possible application to commercial scale fuel fabrication for 19 fuel element designs that use oxide fuel in metal clad rods. The dry gel spheres are prepared at the reprocessing plant and are then calcined, sintered, inspected, and loaded into fuel rods and packed by low-energy vibration. A fuel smear density of 83 to 88% theoretical can be obtained. All fuel fabrication process steps were defined and evaluated from fuel receiving to finished fuel element shipping. The evaluation also covers the feasibility of the process, the current status of technology, estimates of the required time and cost to develop the technology to commercial status, and the safety and licensability of commercial scale plants. The primary evaluation was for a Light-Water Reactor fuel element containing (U,Pu)O 2 fuel. The other 18 fuel element types - 3 for Light-Water Reactors, 1 for a Heavy-Water Reactor, 1 for a Gas-Cooled Fast Reactor, 7 for Liquid-Metal-Cooled Fast Breeder Reactors, and 3 pairs for Light-Water Prebreeder and Breeder Reactors - were compared with the Light-Water Reactor. The gel-sphere-pac option was found applicable to 17 of the 19 element types; the characteristics of a commercial scale plant were defined for these for making cost estimates for such plants. The evaluation clearly shows the gel-sphere-pac process to be a viable fuel fabrication option. Estimates indicate a significant potential fabrication cost advantage for the gel-sphere-pac process if a remotely operated and remotely maintained fuel fabrication plant is required

Mechanisms for Limiting Trade Mark Rights to Further Competition and Free Speech

DEFF Research Database (Denmark)

Ramsey, Lisa P; Schovsbo, Jens Hemmingsen

2013-01-01

This article evaluates the different mechanisms that nations use to limit trade mark rights to promote competition, free speech, and other public interests. It shows how EU and US trade mark laws seem to be converging towards a similar model which includes both (1) specific statutory defenses...... of these mechanisms for limiting trade mark rights to better protect the public interest in trade mark disputes. Finally, a proposal for reform is suggested. It consists of three parts: (1) domestic legislatures should revise their trade mark statutes to add more mandatory and specific limitations on trade mark...... to trade mark violations and (2) trade mark doctrines which give courts flexibility to permit unauthorized uses of marks that further the legitimate interests of the accused infringer and the public. Such a development should be welcomed and the article urges other nations to consider adopting one or both...

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

International Nuclear Information System (INIS)

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

LEU fuel fabrication program for the RECH-1 reactor. Status report

International Nuclear Information System (INIS)

Chavez, J.C.; Barrera, M.; Jimenez, O.; Lisboa, J.; Marin, J.

2000-01-01

In 1995 a 50 LEU U 3 Si 2 fuel elements fabrication program for the RECH-1 research reactor was established at the Comision Chilena de Energia Nuclear, CCHEN. After a fabrication process qualification stage, in 1998, four elements were early delivered to the reactor in order to start an irradiation qualification stage. The irradiation has reached an estimated 10% burn-up and no fabrication problems have been detected up to this burn-up level. During 1999 and up to the first quarter of 2000, 19 fuel elements were produced and 7 fuel elements are expected for the end of 2000. This report presents an updated summary of the main results obtained in this fuel fabrication program. A summary of other activities generated by this program, such as in core follow-up of the four leader fuel elements, ISO 9001 implementation for the fabrication process and a fabrication and qualification optimization planning, is also presented here. (author)

Factors affecting scent-marking behaviour in Eurasion beaver (Castor fiber)

NARCIS (Netherlands)

Rosell, F.; Nolet, B.A.

1997-01-01

We tested the hypothesis that a main function of territory marking in Eurasian beaver (Castor fiber) is defense of the territory. The results showed that: (1) beaver colonies with close neighbors scent-mark more often than isolated ones; (2) the number of scent markings increased significantly with

Drying results of K-Basin fuel element 1990 (Run 1)

International Nuclear Information System (INIS)

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

1998-06-01

The water-filled K-Basins in the Hanford 100-Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basins have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuels in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the first of those tests (Run 1), which was conducted on an N-Reactor inner fuel element (1990) that had been stored underwater in the K-West Basin (see Section 2.0). This fuel element was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The testing was conducted in the Whole Element Furnace Testing System, described in Section 3.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodology are given in Section 4.0, and the experimental results provided in Section 5.0. These results are further discussed in Section 6.0

A French fuel cell prototype

International Nuclear Information System (INIS)

Anon.

2001-01-01

A French prototype of a fuel cell based on the PEM (proton exchange membrane) technology has been designed by Helion, a branch of Technicatome, this fuel cell delivers 300 kW and will be used in naval applications and terrestrial transport. The main advantages of fuel cell are: 1) no contamination, even if the fuel used is natural gas the quantities of CO 2 and CO emitted are respectively 17 and 75 times as little as the maximal quantities allowed by European regulations, 2) efficiency, the electric yield is up to 60 % and can reach 80 % if we include the recovery of heat, 3) silent, the fuel cell itself does not make noise. The present price of fuel cell is the main reason that hampers its industrial development, this price is in fact strongly dependant on the cost of its different components: catalyzers, membranes, bipolar plates and the hydrogen supply. This article gives the technical characteristics of the Helion's fuel cell. (A.C.)

Material Control and Accountability Experience at the Fuel Conditioning Facility

International Nuclear Information System (INIS)

Vaden, D.; Fredrickson, G.L.

2007-01-01

The Fuel Conditioning Facility (FCF) at the Idaho National Laboratory (INL) treats spent nuclear fuel using an electrometallurgical process that separates the uranium from the fission products, sodium thermal bond, and cladding materials. Material accountancy is necessary at FCF for two reasons: 1) it provides a mechanism for detecting a potential loss of nuclear material for safeguards and security, and 2) it provides a periodic check of inventories to ensure that processes and materials are within control limits. Material Control and Accountability is also a Department of Energy (DOE) requirement (DOE Order 474.1). The FCF employs a computer based Mass Tracking (MTG) System to collect, store, retrieve, and process data on all operations that directly affect the flow of materials through the FCF. The MTG System is important for the operations of the FCF because it supports activities such as material control and accountability, criticality safety, and process modeling. To conduct material control and accountability checks and to monitor process performance, mass balances are routinely performed around the process equipment. The equipment used in FCF for pyro-processing consists of two mechanical choppers and two electro-refiners (the Mark-IV with the accompanying element chopper and Mark-V with the accompanying blanket chopper for processing driver fuel and blanket, respectively), and a cathode processor (used for processing both driver fuel and blanket) and casting furnace (mostly used for processing driver fuel). Performing mass balances requires the measurement of the masses and compositions of several process streams and equipment inventories. The masses of process streams are obtained via in-cell balances (i.e., load cells) that weigh containers entering and leaving the process equipment. Samples taken at key locations are analyzed to determine the composition of process streams and equipment inventories. In cases where equipment or containers cannot be

NAC-1 cask dose rate calculations for LWR spent fuel

International Nuclear Information System (INIS)

CARLSON, A.B.

1999-01-01

A Nuclear Assurance Corporation nuclear fuel transport cask, NAC-1, is being considered as a transport and storage option for spent nuclear fuel located in the B-Cell of the 324 Building. The loaded casks will be shipped to the 200 East Area Interim Storage Area for dry interim storage. Several calculations were performed to assess the photon and neutron dose rates. This report describes the analytical methods, models, and results of this investigation

Evaluation of U-Zr hydride fuel for a thorium fuel cycle in an RTR concept

Energy Technology Data Exchange (ETDEWEB)

Lee, Kyung Taek; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1999-12-31

In this paper, we performed a design study of a thorium fueled reactor according to the design concept of the Radkowsky Thorium Reactor (RTR) and evaluated its overall performance. To enhance its performance and alleviate its problems, we introduced a new metallic uranium fuel, uranium-zirconium hydride (U-ZrH{sub 1.6}), as a seed fuel. For comparison, typical ABB/CE-type PWR based on SYSTEM 80+and standard RTR-type thorium reactor were also studied. From the results of performance analysis, we could ascertain advantages of RTR-type thorium fueled reactor in proliferation resistance, fuel cycle economics, and back-end fuel cycle. Also, we found that enhancement of proliferation resistance and safer operating conditions may be achieved by using the U-ZrH{sub 1.6} fuel in the seed region without additional penalties in comparison with the standard RTR`s U-Zr fuel. 6 refs., 2 figs., 6 tabs. (Author)

Evaluation of U-Zr hydride fuel for a thorium fuel cycle in an RTR concept

Energy Technology Data Exchange (ETDEWEB)

Lee, Kyung Taek; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1998-12-31

In this paper, we performed a design study of a thorium fueled reactor according to the design concept of the Radkowsky Thorium Reactor (RTR) and evaluated its overall performance. To enhance its performance and alleviate its problems, we introduced a new metallic uranium fuel, uranium-zirconium hydride (U-ZrH{sub 1.6}), as a seed fuel. For comparison, typical ABB/CE-type PWR based on SYSTEM 80+and standard RTR-type thorium reactor were also studied. From the results of performance analysis, we could ascertain advantages of RTR-type thorium fueled reactor in proliferation resistance, fuel cycle economics, and back-end fuel cycle. Also, we found that enhancement of proliferation resistance and safer operating conditions may be achieved by using the U-ZrH{sub 1.6} fuel in the seed region without additional penalties in comparison with the standard RTR`s U-Zr fuel. 6 refs., 2 figs., 6 tabs. (Author)

A design study of high breeding ratio sodium cooled metal fuel core without blanket fuels

International Nuclear Information System (INIS)

Kobayashi, Noboru; Ogawa, Takashi; Ohki, Shigeo; Mizuno, Tomoyasu; Ogata, Takanari

2009-01-01

The metal fuel core is superior to the mixed oxide fuel core because of its high breeding ratio and compact core size resulting from hard neutron spectrum and high heavy metal densities. Utilizing these characteristics, a conceptual design for a high breeding ratio was performed without blanket fuels. The design conditions were set so a sodium void worth of less than 8 $, a core height of less than 150 cm, the maximum cladding temperature of 650degC, and the maximum fuel pin bundle pressure drop of 0.4 MPa. The breeding ratio of the resultant core was 1.34 with 6wt% zirconium content fuel. Applying 3wt% zirconium content fuel enhanced the breeding ratio up to 1.40. (author)

Neutron detection of the Triga Mark III reactor, using nuclear track methodology

Energy Technology Data Exchange (ETDEWEB)

Espinosa, G., E-mail: espinosa@fisica.unam.mx; Golzarri, J. I. [Instituto de Física, Universidad Nacional Autónoma de México Circuito de la Investigación Científica, Ciudad Universitaria. México, DF (Mexico); Raya-Arredondo, R.; Cruz-Galindo, S. [Instituto Nacional de Investigaciones Nucleares (Mexico); Sajo-Bohus, L. [Universidad Simón Bolivar, Laboratorio de Física Nuclear, Caracas (Venezuela, Bolivarian Republic of)

2015-07-23

Nuclear Track Methodology (NTM), based on the neutron-proton interaction is one often employed alternative for neutron detection. In this paper we apply NTM to determine the Triga Mark III reactor operating power and neutron flux. The facility nuclear core, loaded with 85 Highly Enriched Uranium as fuel with control rods in a demineralized water pool, provide a neutron flux around 2 × 10{sup 12} n cm{sup −2} s{sup −1}, at the irradiation channel TO-2. The neutron field is measured at this channel, using Landauer{sup ®} PADC as neutron detection material, covered by 3 mm Plexiglas{sup ®} as converter. After exposure, plastic detectors were chemically etched to make observable the formed latent tracks induced by proton recoils. The track density was determined by a custom made Digital Image Analysis System. The resulting average nuclear track density shows a direct proportionality response for reactor power in the range 0.1-7 kW. We indicate several advantages of the technique including the possibility to calibrate the neutron flux density measured at low reactor power.

Capture programs, analysis, data graphication for the study of the thermometry of the TRIGA Mark III reactor core; Programas de captura, analisis y graficado de datos para el estudio de la termometria del nucleo del reactor TRIGA Mark III

Energy Technology Data Exchange (ETDEWEB)

Paredes G, L.C

1991-05-15

This document covers the explanation of the capture programs, analysis and graphs of the data obtained during the measurement of the temperatures of the instrumented fuel element of the TRIGA Mark III reactor and of the coolant one near to this fuel, using the conversion card from Analogic to Digital of 'Data Translation', and using a signal conditioner for five temperature measurers with the help of thermo par type K, developed by the Simulation and Control of the nuclear systems management department, which gives a signal from 0 to 10 Vcd for an interval of temperature of 0 to 1000 C. (Author)

Low NO subx heavy fuel combustor concept program. Phase 1A: Coal gas addendum

Science.gov (United States)

Rosfjord, T.; Sederquist, R.

1982-01-01

The performance and emissions from a rich-lean combustor fired on simulated coal gas fuels were investigated using a 12.7-cm diameter axially-staged burner originally designed for operation with high heating value liquid fuels. A simple, tubular fuel injector was substituted for the liquid fuel nozzle; no other combustor modifications were made. Four test fuels were studied including three chemically bound nitrogen-free gas mixtures with higher heating values of 88, 227, and 308 kj/mol (103, 258 and 349 Btu/scf), and a 227 kj/mol (258 Btu/scf) heating value doped with ammonia to produce a fuel nitrogen content of 0.5% (wt). Stable, ultra-low nitrogen oxide, smoke-free combustion was attained for the nitrogen-free fuels. Results with the doped fuel indicated that less than 5% conversion of NH3 to nitrogen oxide levels below Environmental Protection Agency limits could be achieved. In some instances, excessive CO levels were encountered. It is shown that use of a burner design employing a less fuel-rich primary zone than that found optimum for liquid fuels would yield more acceptable CO emissions.

Obtaining of total and thermal neutron flux in the carousel facility of the TRIGA MARK IPR-R1 reactor using the Monte Carlo transport method

International Nuclear Information System (INIS)

Guerra, Bruno Teixeira

2011-01-01

The IPR-R1 is a reactor type TRIGA, Mark-I model, manufactured by the General Atomic Company and installed at Nuclear Technology Development Centre (CDTN) of Brazilian Nuclear Energy Commission (CNEN), in Belo Horizonte, Brazil. It is a light water moderated and cooled, graphite-reflected, open-pool type research reactor. IPR-R1 works at 100 kW but it will be briefly licensed to operate at 250 kW. It presents low power, low pressure, for application in research, training and radioisotopes production. The fuel is an alloy of zirconium hydride and uranium enriched at 20% in 235 U. The goal this work is modelling of the IPR-R1 Research Reactor TRIGA using the codes MCNPX2.6.0 (Monte Carlo N-Particle Transport extend) and MCNP5 to the calculating the neutron flux in the carousel facility. In each simulation the sample was placed in a different position, totaling forty positions around of the reactor core. The comparison between the results obtained with experimental values from other work showing a relatively good agreement. Moreover, this methodology is a theoretical tool in validating of the experimental values and necessary for determining neutron flux which can not be accessible experimentally. (author)

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

International Nuclear Information System (INIS)

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

1978-01-01

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

Catalogue of fuel elements - 1. addendum October 1958

International Nuclear Information System (INIS)

Even, A.

1957-01-01

This document contains sheets presenting various characteristics of nuclear fuel elements which are distinguished with respect to their shape: cylinder bar, plate, tube. Each sheet comprises an indication of the atomic pile in which the fuel element is used, dimensions, cartridge data, data related to cooling, to combustion rate, and to fuel handling. A drawing of the fuel element is also given

7 CFR 319.37-10 - Marking and identity.

Science.gov (United States)

2010-01-01

... 7 Agriculture 5 2010-01-01 2010-01-01 false Marking and identity. 319.37-10 Section 319.37-10 Agriculture Regulations of the Department of Agriculture (Continued) ANIMAL AND PLANT HEALTH INSPECTION..., and Other Plant Products 1,2 § 319.37-10 Marking and identity. (a) Any restricted article for...

Fuel handling at Cernavoda 1 N.P.S. - commissioning and training philosophy

Energy Technology Data Exchange (ETDEWEB)

Standen, G W [AECL-Ansaldo Consortium, Cernavoda (Romania); Tiron, C; Marinescu, S [Regia Nationala de Electricitate (RENEL), Cernavoda (Romania); [Filiala Centrala Nuclearo Electrica (FCNE), Cernavoda (Romania)

1997-12-31

Efficient operation of a Candu nuclear power plant depends greatly on the reliable and safe operation of the fuel handling system. Successful commissioning of the system is obviously a key aspect of the reliability of the system and this coupled with a rigorous training programme for the fuel handling staff will ensure the system`s safe operation. This paper describes the philosophy used at Cernavoda 1 N.P.S. for the commissioning of the fuel handling systems and for the training of staff for operation and maintenance of these systems. The paper also reviews the commissioning programme, describing the milestones achieved and discussing some of the more interesting technical aspects which includes some unique Romanian input. In conclusion the paper looks at the organization of the mature fuel handling department from the operations, maintenance and technical support points of view and the long term plans for the future. (author). 1 fig.

Fuel handling at Cernavoda 1 N.P.S. - commissioning and training philosophy

International Nuclear Information System (INIS)

Standen, G.W.; Tiron, C.; Marinescu, S.

1996-01-01

Efficient operation of a Candu nuclear power plant depends greatly on the reliable and safe operation of the fuel handling system. Successful commissioning of the system is obviously a key aspect of the reliability of the system and this coupled with a rigorous training programme for the fuel handling staff will ensure the system's safe operation. This paper describes the philosophy used at Cernavoda 1 N.P.S. for the commissioning of the fuel handling systems and for the training of staff for operation and maintenance of these systems. The paper also reviews the commissioning programme, describing the milestones achieved and discussing some of the more interesting technical aspects which includes some unique Romanian input. In conclusion the paper looks at the organization of the mature fuel handling department from the operations, maintenance and technical support points of view and the long term plans for the future. (author). 1 fig

Protective Coatings for Wet Storage of Aluminium-Clad Spent Fuel

Energy Technology Data Exchange (ETDEWEB)

Fernandes, S.M.C.; Correa, O.V.; Souza, J.A. De; Ramanathan, L.V. [Materials science and Technology Center, Instituto de Pesquisas Energeticas e Nucleares - IPEN, Av. Prof. Lineu Prestes 2242, Cidade Universitaria, 05508-000 Sao Paulo (Brazil)

2011-07-01

Corrosion protection of spent RR fuel for long term wet storage was considered important, primarily from the safety standpoint and the use of conversion coatings was proposed in 2008. This paper presents the results of: (a) on-going field tests in which un-coated and lanthanide-based conversion coated Al alloy coupons were exposed to the IEA-R1 reactor spent fuel basin for durations of up to a year; (b) preparation of cerium modified hydrotalcite coatings and cerium sealed boehmite coatings on AA 6061 alloy; (c) corrosion resistance of coated specimens in NaCl solutions. The field studies indicated that the oxidized and cerium dioxide coated coupons were the most corrosion resistant. The cerium modified hydrotalcite and cerium sealed boehmite coated specimens showed marked increase in pitting corrosion resistance. (author)

Improved Cathode Structure for a Direct Methanol Fuel Cell

Science.gov (United States)

Valdez, Thomas; Narayanan, Sekharipuram

2005-01-01

An improved cathode structure on a membrane/electrode assembly has been developed for a direct methanol fuel cell, in a continuing effort to realize practical power systems containing such fuel cells. This cathode structure is intended particularly to afford better cell performance at a low airflow rate. A membrane/electrode assembly of the type for which the improved cathode structure was developed (see Figure 1) is fabricated in a process that includes brush painting and spray coating of catalyst layers onto a polymer-electrolyte membrane and onto gas-diffusion backings that also act as current collectors. The aforementioned layers are then dried and hot-pressed together. When completed, the membrane/electrode assembly contains (1) an anode containing a fine metal black of Pt/Ru alloy, (2) a membrane made of Nafion 117 or equivalent (a perfluorosulfonic acid-based hydrophilic, proton-conducting ion-exchange polymer), (3) a cathode structure (in the present case, the improved cathode structure described below), and (4) the electrically conductive gas-diffusion backing layers, which are made of Toray 060(TradeMark)(or equivalent) carbon paper containing between 5 and 6 weight percent of poly(tetrafluoroethylene). The need for an improved cathode structure arises for the following reasons: In the design and operation of a fuel-cell power system, the airflow rate is a critical parameter that determines the overall efficiency, cell voltage, and power density. It is desirable to operate at a low airflow rate in order to obtain thermal and water balance and to minimize the size and mass of the system. The performances of membrane/electrode assemblies of prior design are limited at low airflow rates. Methanol crossover increases the required airflow rate. Hence, one way to reduce the required airflow rate is to reduce the effect of methanol crossover. Improvement of the cathode structure - in particular, addition of hydrophobic particles to the cathode - has been

Nondestructive examination of Oconee 1 fuel assemblies after three cycles of irradiation

International Nuclear Information System (INIS)

Pyecha, T.D.; Davis, H.H.; Mayer, J.T.; Guthrie, B.A. III; Larson, J.G.

1979-09-01

The Babcock and Wilcox Company (B and W) in conjunction with Duke Power Company is participating in a Department of Energy sponsored research and development program to qualify current design pressurized water reactor (PWR) fuel assemblies for extended burnup (>40,000 MWd/mtU). The information obtained from this program will provide a basis for future design improvements in PWR fuel assemblies culminating in an extended burnup assembly having a nominal operating limit of approximately 50,000 MWd/mtU. An extension of the current assembly design to higher burnups will result in the following benefits: (1) lower uranium ore requirements, (2) greater fuel cycle efficiency, (3) reduction in spent fuel storage requirements, and (4) increased flexibility in tailoring fuel batch sizes to better accommodate the varying energy requirements of the utilities

Fuel cell power plants for automotive applications

Science.gov (United States)

McElroy, J. F.

1983-02-01

While the Solid Polymer Electrolyte (SPE) fuel cell has until recently not been considered competitive with such commercial and industrial energy systems as gas turbine generators and internal combustion engines, electrical current density improvements have markedly improved the capital cost/kW output rating performance of SPE systems. Recent studies of SPE fuel cell applicability to vehicular propulsion have indicated that with adequate development, a powerplant may be produced which will satisfy the performance, size and weight objectives required for viable electric vehicles, and that the cost for such a system would be competitive with alternative advanced power systems.

Theoretical evaluation of the production of the poisons Xe-135 and Sm-149 of the TRIGA Mark III reactor with mixed core; Evaluacion teorica de la produccion de los venenos Xe-135 y Sm-149 del reactor TRIGA Mark III con nucleo mixto

Energy Technology Data Exchange (ETDEWEB)

Paredes G, L.C

1991-11-15

It was theoretically determined the accumulation of the Xe{sup 135} and Sm {sup 149} in function, of the time during a stationary state of 72 h. continuous for the reactor TRIGA Mark III to 1 MW of thermal power with mixed core. The values of negative reactivity due to these isotopes are of 2.04 dollars and 0.694 dollars to the 72 h, quantities that will have to be compensated if wants that the reactor continues working to this power. Under the same conditions but considering a core with standard fuel, it was found a value of {rho} = 1.70 dollars, resulting a difference of 0.30 dollars of negative reactivity in function of the type of analyzed core. This difference is important for the calculations of fuel management of a reactor. The concentration in balance of the xenon was reaches after an operation to constant power of 1 MW by 50 h, contrary to the samarium that reaches it balance after 3 weeks of operation starting from the initial start up and it stays constant along the useful life of the reactor while a change of fuel doesn't exist. It was obtained that for operation times greater to 60 h. at 1 MW, a peak of negative reactivity of the Xe{sup 135} is generated between the 7 and 11 h after the instantaneous shut down, with a value of 2.43 dollars, that is to say 0.39 additional dollars to those taken place during the continuous irradiation. (Author)

Fuel cell power systems for remote applications. Phase 1 final report and business plan

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-02-01

The goal of the Fuel Cell Power Systems for Remote Applications project is to commercialize a 0.1--5 kW integrated fuel cell power system (FCPS). The project targets high value niche markets, including natural gas and oil pipelines, off-grid homes, yachts, telecommunication stations and recreational vehicles. Phase 1 includes the market research, technical and financial analysis of the fuel cell power system, technical and financial requirements to establish manufacturing capability, the business plan, and teaming arrangements. Phase 1 also includes project planning, scope of work, and budgets for Phases 2--4. The project is a cooperative effort of Teledyne Brown Engineering--Energy Systems, Schatz Energy Research Center, Hydrogen Burner Technology, and the City of Palm Desert. Phases 2 through 4 are designed to utilize the results of Phase 1, to further the commercial potential of the fuel cell power system. Phase 2 focuses on research and development of the reformer and fuel cell and is divided into three related, but potentially separate tasks. Budgets and timelines for Phase 2 can be found in section 4 of this report. Phase 2 includes: Task A--Develop a reformate tolerant fuel cell stack and 5 kW reformer; Task B--Assemble and deliver a fuel cell that operates on pure hydrogen to the University of Alaska or another site in Alaska; Task C--Provide support and training to the University of Alaska in the setting up and operating a fuel cell test lab. The Phase 1 research examined the market for power systems for off-grid homes, yachts, telecommunication stations and recreational vehicles. Also included in this report are summaries of the previously conducted market reports that examined power needs for remote locations along natural gas and oil pipelines. A list of highlights from the research can be found in the executive summary of the business plan.

Theoretical analysis of the temperature changes and resultant loss of fuel integrity in the IEA-R1 research reactor fuel elements following a loss of coalant accident

International Nuclear Information System (INIS)

Garone, J.G.M.

1983-01-01

The IEA-R1 core following a loss of coolant accident (LOCA) is analysed. THe AIRLOCA code was used to calculate fuel temperatures, heat generation due to fission product decay and convective and radiative heat transfer from the fuel elements to the surrounding air both during and following the loss of coolant. The influence of certain critical parameters, such as log time, specific power was studied in detail. Representative results are presented and suggestions made to ensure that fuel integrity is maintained following a LOCA. (Author) [pt

Neutron spectra in two beam ports of the TRIGA Mark III reactor

International Nuclear Information System (INIS)

Vega C, H. R.; Hernandez D, V. M.; Aguilar, F.; Paredes, L.; Rivera M, T.

2013-10-01

The neutron spectra have been measured in two beam ports, radial and tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research. Measurements were carried out with the core with mixed fuel (Leu 8.5/20 and Flip Heu 8.5/70). Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a 6 Lil(Eu) scintillator and 2, 3, 5, 8, 10 and 12 inches-diameter high density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code; from each spectrum the total neutron flux, the neutron mean energy and the neutron ambient dose equivalent dose were determined. Measured spectra show fission (E≥ 0.1 MeV), epithermal (from 0.4 eV up to 0.1 MeV) and thermal neutrons (E≤ 0.4 eV). For both reactor powers the spectra in the radial beam port have similar features which are different to the neutron spectrum characteristics in the tangential beam port. (Author)

Neutron spectra in two beam ports of the TRIGA Mark III reactor

Energy Technology Data Exchange (ETDEWEB)

Vega C, H. R.; Hernandez D, V. M. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98060 Zacatecas (Mexico); Aguilar, F.; Paredes, L. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Rivera M, T., E-mail: fermineutron@yahoo.com [IPN, Centro de Investigacion en Ciencia Aplicada y Tecnologia Avanzada, Unidad Legaria, Av. Legaria 694, 11500 Mexico D. F. (Mexico)

2013-10-15

The neutron spectra have been measured in two beam ports, radial and tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research. Measurements were carried out with the core with mixed fuel (Leu 8.5/20 and Flip Heu 8.5/70). Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a {sup 6}Lil(Eu) scintillator and 2, 3, 5, 8, 10 and 12 inches-diameter high density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code; from each spectrum the total neutron flux, the neutron mean energy and the neutron ambient dose equivalent dose were determined. Measured spectra show fission (E≥ 0.1 MeV), epithermal (from 0.4 eV up to 0.1 MeV) and thermal neutrons (E≤ 0.4 eV). For both reactor powers the spectra in the radial beam port have similar features which are different to the neutron spectrum characteristics in the tangential beam port. (Author)

Factors affecting actinide solubility in a repository for spent fuel, 1

International Nuclear Information System (INIS)

Snellman, Margit

1986-07-01

The main tasks in the study were to get information on the chemical conditions in a repository for spent fuel and information on factors affecting releases of actinides from spent fuel and solubility of actinides in a repository for spent fuel. The work in this field started at the Reactor Laboratory of the Technical Research Centre of Finland (VTT) in 1982. This is a report on the effects on the main parameters, Eh, pH, carbonate, organic compounds, colloids, microbes and radiation on the actinide solubility in the nearfield of the repository. Another task has been to identify available models and reported experience from actinide solubility calculations with different codes. 167 refs

Design of a new wet storage rack for spent fuels from IEA-R1 reactor

International Nuclear Information System (INIS)

Rodrigues, Antonio C.I.; Madi Filho, Tufic; Siqueira, Paulo T.D.; Ricci Filho, Walter

2015-01-01

The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks of the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating conditions, the storage will have capacity for about six years. Since the estimated useful life of the IEA-R1 is about another 20 years, it will be necessary to increase the storage capacity of spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. After an extensive literature review of material options given for this type of application we got to Boral® manufactured by 3M due to numerous advantages. This paper presents studies on the analysis of criticality using the computer code MCNP 5, demonstrating the possibility of doubling the storage capacity of current racks to attend the demand of the IEA-R1 reactor while attending the safety requirements the International Atomic Energy Agency. (author)

Design of a new wet storage rack for spent fuels from IEA-R1 reactor

Energy Technology Data Exchange (ETDEWEB)

Rodrigues, Antonio C.I.; Madi Filho, Tufic; Siqueira, Paulo T.D.; Ricci Filho, Walter, E-mail: acirodri@ipen.br, E-mail: tmfilho@ipen.br, E-mail: ptsiquei@ipen.br, E-mail: wricci@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks of the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating conditions, the storage will have capacity for about six years. Since the estimated useful life of the IEA-R1 is about another 20 years, it will be necessary to increase the storage capacity of spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. After an extensive literature review of material options given for this type of application we got to Boral® manufactured by 3M due to numerous advantages. This paper presents studies on the analysis of criticality using the computer code MCNP 5, demonstrating the possibility of doubling the storage capacity of current racks to attend the demand of the IEA-R1 reactor while attending the safety requirements the International Atomic Energy Agency. (author)

VHTR-fuel irradiation capsules for VT-1 hole of JRR-2

International Nuclear Information System (INIS)

Kikuchi, Teruo; Kikuchi, Akira; Tobita, Tsutomu; Kashimura, Satoru; Miyasaka, Yasuhiko

1977-02-01

Irradiations of VHTR fuels were made in the VT-1 irradiation hole of JRR-2. Three capsules, VP-1, VP-2 and VP-4, which contained fuel compacts, were irradiated for 300 hr at temperatures of 950 0 , 1370 0 and 1500 0 C up to the estimated burn-ups of 0.74, 0.87 and 0.80%FIMA, respectively. And, to study the amoeba effect of fuel particles, two capsules, VP-3 and VP-5, were irradiated for 300 hr at temperatures of 1650 0 and 1670 0 C up to the estimated burn-ups of 0.38 and 0.33%FIMA, respectively. (auth.)

Spent fuel pool cooling system upgrade for Kori Unit 1

International Nuclear Information System (INIS)

Sun Park, Jong; In Shin, Kyung

2014-01-01

Following Fukushima nuclear power plant accident, the needs for reliable performance of its own safety functions of Spent Fuel Pool Cooling System (SFPCS) has risen significantly to maintain the plant in a safe condition. Regulatory Guide 1.13 of United States Nuclear Regulatory Commission (USNRC) requires the SFPCS shall be designed safety related as Quality Group C and Seismic Category 1. However, the existing Spent Fuel Pool (SFP) of KORI Unit 1 was not designed as a safety system. In order to comply with the above licensing requirement for the extended operational life of KORI Unit 1, it has been decided to add a safety related Seismic Category 1 Makeup System to KORI Unit 1 and the existing SFPCS to be modified in dedicated channels with safety related equipment to enhance system's reliability as a means of providing diversity. This paper focuses on describing the relevant design requirements, applications, and supplemental facilities to the SFPCS of KORI Unit 1. (authors)

ANALYSIS OF GAMMA HEATING AT TRIGA MARK REACTOR CORE BANDUNG USING PLATE TYPE FUEL

Directory of Open Access Journals (Sweden)

Setiyanto Setiyanto

2016-10-01

Full Text Available ABSTRACT In accordance with the discontinuation of TRIGA fuel element production by its producer, the operation of all TRIGA type reactor of at all over the word will be disturbed, as well as TRIGA reactor in Bandung. In order to support the continuous operation of Bandung TRIGA reactor, a study on utilization of fuel plate mode, as used at RSG-GAS reactor, to replace the cylindrical model has been done. Various assessments have been done, including core design calculation and its safety aspects. Based on the neutronic calculation, utilization of fuel plate shows that Bandung TRIGA reactor can be operated by 20 fuel elements only. Compared with the original core, the new reactor core configuration is smaller and it results in some empty space that can be used for in-core irradiation facilities. Due to the existing of in-core irradiation facilities, the gamma heating value became a new factor that should be evaluated for safety analysis. For this reason, the gamma heating for TRIGA Bandung reactor using fuel plate was calculated by Gamset computer code. The calculations based on linear attenuation equations, line sources and gamma propagation on space. Calculations were also done for reflector positions (Lazy Susan irradiation facilities and central irradiation position (CIP, especially for any material samples. The calculation results show that gamma heating for CIP is significantly important (0,87 W/g, but very low value for Lazy Susan position (lest then 0,11 W/g. Based on this results, it can be concluded that the utilization of CIP as irradiation facilities need to consider of gamma heating as data for safety analysis report. Keywords: gamma heating, nuclear reactor, research reactor, reactor safety.   ABSTRAK Dengan dihentikannya produksi elemen bakar reaktor jenis Triga oleh produsen, maka semua reaktor TRIGA di dunia terganggu operasinya, termasuk juga reaktor TRIGA 2000 di Bandung. Untuk mendukung pengoperasian reaktor TRIGA Bandung

Fuel models and results from the TRAC-PF1/MIMAS TMI-2 accident calculation

International Nuclear Information System (INIS)

Schwegler, E.C.; Maudlin, P.J.

1983-01-01

A brief description of several fuel models used in the TRAC-PF1/MIMAS analysis of the TMI-2 accident is presented, and some of the significant fuel-rod behavior results from this analysis are given. Peak fuel-rod temperatures, oxidation heat production, and embrittlement and failure behavior calculated for the TMI-2 accident are discussed. Other aspects of fuel behavior, such as cladding ballooning and fuel-cladding eutectic formation, were found not to significantly affect the accident progression

Off-normal performance of EBR-II [Experimental Breeder Reactor] driver fuel

International Nuclear Information System (INIS)

Seidel, B.R.; Batte, G.L.; Lahm, C.E.; Fryer, R.M.; Koenig, J.F.; Hofman, G.L.

1986-09-01

The off-normal performance of EBR-II Mark-II driver fuel has been more than satisfactory as demonstrated by robust reliability under repeated transient overpower and undercooled loss-of-flow tests, by benign run-beyond-cladding-breach behavior, and by forgiving response to fabrication defects including lack of bond. Test results have verified that the metallic driver fuel is very tolerant of off-normal events. This behavior has allowed EBR-II to operate in a combined steady-state and transient mode to provide test capability without limitation from the metallic driver fuel

Neutron flux determination at the IPR-R1 Triga Mark I neutron beam extractor

International Nuclear Information System (INIS)

Zangirolami, Dante Marco; Maretti Junior, Fausto; Ferreira, Andrea Vidal

2009-01-01

The IPR-R1 Triga Mark I Reactor located at the CDTN/CNEN, Belo Horizonte, Brazil, has been operating since November of 1960. In this work, measurements of thermal and epithermal neutron flux along the IPR-R1 neutron beam extractor were performed by neutron activation of reference materials using the two foils method. The obtained results were compared with results from two previous works: an experimental measurement done in a previous reactor core configuration and a numerical work made by Monte Carlo simulation using the actual reactor core configuration. The main purpose of this work is to update the measured data to the actual reactor core configuration. (author)

Development and use of GREET 1.6 fuel-cycle model for transportation fuels and vehicle technologies

International Nuclear Information System (INIS)

Wang, M. Q.

2001-01-01

Since 1995, with funds from the U.S. Department of Energy's (DOE's) Office of Transportation Technologies (OTT), Argonne National Laboratory has been developing the Greenhouse gases, Regulated Emissions, and Energy use in Transportation (GREET) model. The model is intended to serve as an analytical tool for use by researchers and practitioners in estimating fuel-cycle energy use and emissions associated with alternative transportation fuels and advanced vehicle technologies. Argonne released the first version of the GREET mode--GREET 1.0--in June 1996. Since then, it has released a series of GREET versions with revisions, updates, and upgrades. In February 2000, the latest public version of the model--GREET 1.5a--was posted on Argonne's Transportation Technology Research and Development Center (TTRDC) Web site (www.transportation.anl.gov/ttrdc/greet)

Alternatives to traditional transportation fuels 1994. Volume 1

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-02-01

In this report, alternative and replacement fuels are defined in accordance with the EPACT. Section 301 of the EPACT defines alternative fuels as: methanol, denatured ethanol, and other alcohols; mixtures containing 85% or more (or such other percentage, but not less than 70%, as determined by the Secretary of Energy, by rule, to provide for requirements relating to cold start, safety, or vehicle functions) by volume of methanol, denatured ethanol, and other alcohols with gasoline or other fuels; natural gas; liquefied petroleum gas; hydrogen; coal-derived liquid fuels; fuels (other than alcohol) derived from biological materials; electricity (including electricity from solar energy); and any other fuel the Secretary determines, by rule, is substantially not petroleum and would yield substantial energy security benefits and substantial environmental benefits. The EPACT defines replacement fuels as the portion of any motor fuel that is methanol, ethanol, or other alcohols, natural gas, liquefied petroleum gas, hydrogen, coal-derived liquid fuels, fuels (other than alcohol) derived from biological materials, electricity (including electricity from solar energy), ethers, or any other fuel the Secretary of Energy determines, by rule, is substantially not petroleum and would yield substantial energy security benefits and substantial environmental benefits. This report covers only those alternative and replacement fuels cited in the EPACT that are currently commercially available or produced in significant quantities for vehicle demonstration purposes. Information about other fuels, such as hydrogen and biodiesel, will be included in later reports as those fuels become more widely used. Annual data are presented for 1992 to 1996. Data for 1996 are based on plans or projections for 1996.

Fuel cells science and engineering. Materials, processes, systems and technology. Vol. 1

Energy Technology Data Exchange (ETDEWEB)

Stolten, Detlef; Emonts, Bernd (eds.) [Forschungszentrum Juelich GmbH (DE). Inst. fuer Energieforschung (IEF), Brennstoffzellen (IEF-3)

2012-07-01

The first volume is divided in four parts and 22 chapters. It is structured as follows: PART I: Technology. Chapter 1: Technical Advancement of Fuel-Cell Research and Development (Dr. Bernd Emonts, Ludger Blum, Thomas Grube, Werner Lehnert, Juergen Mergel, Martin Mueller and Ralf Peters); 2: Single-Chamber Fuel Cells (Teko W. Napporn and Melanie Kuhn); 3: Technology and Applications of Molten Carbonate Fuel Cells (Barbara Bosio, Elisabetta Arato and Paolo Greppi); 4: Alkaline Fuel Cells (Erich Guelzow); 5: Micro Fuel Cells (Ulf Groos and Dietmar Gerteisen); 6: Principles and Technology of Microbial Fuel Cells (Jan B. A. Arends, Joachim Desloover, Sebastia Puig and Willy Verstraete); 7: Micro-Reactors for Fuel Processing (Gunther Kolb); 8: Regenerative Fuel Cells (Martin Mueller). PART II: Materials and Production Processes. Chapter 9: Advances in Solid Oxide Fuel Cell Development between 1995 and 2010 at Forschungszentrum Juelich GmbH, Germany (Vincent Haanappel); 10: Solid Oxide Fuel Cell Electrode Fabrication by Infiltration (Evren Gunen); 11: Sealing Technology for Solid Oxide Fuel Cells (K. Scott Weil); 12: Phosphoric Acid, an Electrolyte for Fuel Cells - Temperature and Composition Dependence of Vapor Pressure and Proton Conductivity (Carsten Korte); 13: Materials and Coatings for Metallic Bipolar Plates in Polymer Electrolyte Membrane Fuel Cells (Heli Wang and John A. Turner); 14: Nanostructured Materials for Fuel Cells (John F. Elter); 15: Catalysis in Low-Temperature Fuel Cells - An Overview (Sabine Schimpf and Michael Bron). PART III: Analytics and Diagnostics. Chapter 16: Impedance Spectroscopy for High-Temperature Fuel Cells (Ellen Ivers-Tiffee, Andre Leonide, Helge Schichlein, Volker Sonn and Andre Weber); 17: Post-Test Characterization of Solid Oxide Fuel-Cell Stacks (Norbert H. Menzler and Peter Batfalsky); 18: In Situ Imaging at Large-Scale Facilities (Christian Toetzke, Ingo Manke and Werner Lehnert); 19: Analytics of Physical Properties of Low

Occipital lobe seizures related to marked elevation of hemoglobin A1C: report of two cases.

Science.gov (United States)

Hung, Wan-Ling; Hsieh, Peiyuan F; Lee, Yi-Chung; Chang, Ming-Hong

2010-07-01

Occipital lobe seizures caused by nonketotic hyperglycemia (NKH) have been reported in only a few cases and are not fully characterized. We report two cases of NKH-related occipital lobe seizures with high hemoglobin A1C (HbA1C), epileptiform electroencephalograph (EEG) and MRI abnormalities. Both patients had moderate hyperglycemia (310-372 mg/dl) and mildly elevated serum osmolarity (295-304 mOsm/kg) but markedly elevated HbA1C (13.8-14.4%). One patient had a clinico-EEG seizure originating from the right occipital region during sleep. The other patient had an interictal epileptiform discharge consisting of unilateral occipital beta activity in sleep. None of the previously reported cases fulfilled the criteria of a nonketotic hyperglycemic hyperosmolar (NKHH) state, or showed any interictal beta paroxysms, spikes, sharp waves, or spike/sharp-slow wave complexes. We suggest that prolonged exposure to uncontrolled hyperglycemia, as indicated by HbA1C, rather than an acute NKHH state is crucial in the development of this peculiar seizure. We also suggest clinicians look for the presence of interictal focal beta paroxysms in addition to the usual epileptiform discharges while reading the EEG of these patients. 2010 British Epilepsy Association. Published by Elsevier Ltd. All rights reserved.

Design support document for the K Basins Vertical Fuel Handling Tools

International Nuclear Information System (INIS)

Bridges, A.E.

1995-01-01

The purpose of this document is to provide the design support information for the Vertical Fuel Handling Tools, developed for the removal of N Reactor fuel elements from their storage canisters in the K Basins storage pool and insertion into the Single Fuel Element Can for subsequent shipment to a Hot Cell for examination. Examination of these N Reactor fuel elements is part of the overall characterization effort. These new hand tools are required since previous fuel movement has involved grasping the fuel in a horizontal position. These tools are required to lift an element vertically from the storage canister. Additionally, a Mark II storage canister Lip Seal Protector was designed and fabricated for use during fuel retrieval. This device was required to prevent damage to the canister lip should a fuel element accidentally be dropped during its retrieval, using the handling tools. Supporting documentation for this device is included in this document

A numerical investigation on the influence of EGR in a supercharged SI engine fueled with gasoline and alternative fuels

International Nuclear Information System (INIS)

Mardi K, Mohsen; Khalilarya, Shahram; Nemati, Arash

2014-01-01

Highlights: • CFD modeling the combustion of different alternative fuels in SI engine. • 10% of EGR is the most desirable amount from the viewpoint of emissions and power. • EGR affects on methane fuel more than others. • Supercharging has the most noticeable effect on gasoline fuel and the least on hydrogen fuel. - Abstract: Alternative fuels are mostly extracted from renewable resources, and their emission levels can be lower than those of traditional fossil-based fuels. A computational fluid dynamics (CFD) method is utilized to investigate the effects of exhaust gas recirculation (EGR) and initial charge pressure on the emissions and performance of a SI engine. The engine is fueled separately by gasoline and some of potential alternative fuels including hydrogen, propane, methane, ethanol and methanol. The results of simulation are compared to the experimental data. In all validation cases, experimental and numerical results were observed to have good agreement with each other. The calculations are carried out for EGR ratios between 0% and 20% and four cases of initial pressure have been mentioned: P in = 1, 1.2, 1.4, 1.6 bar. The effect of EGR on NO x emission of methane is more than other fuels and its effect on IMEP of hydrogen is less than other fuels. From the viewpoints of emission and power, 10% of EGR seems to be the most desirable amount. The most noticeable effect of supercharging is on gasoline unlike hydrogen, which seems to be affected the least. The comparison of results shows that hydrogen due to its high heating value and burning without producing any carbon-based compounds such as HC, CO and CO 2 is an ideal alternative fuel compared to the other fuels

What light does Matthew's use of Mark in Matthew 1–4 throw on ...

African Journals Online (AJOL)

2016-06-30

Jun 30, 2016 ... use of Mark in these chapters is confined largely to Matthew 3–4, this article will ... that to add material is also a way of responding to Mark, so will give just as much attention to .... judgement (13:41–43; 25:31) with parallels in the Parables of ... Christology which would have gone far too far for most Jews.

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

International Nuclear Information System (INIS)

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

2017-01-01

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

PCR+ In Diesel Fuels and Emissions Research

Energy Technology Data Exchange (ETDEWEB)

McAdams, H.T.

2002-04-15

In past work for the U.S. Department of Energy (DOE) and Oak Ridge National Laboratory (ORNL), PCR+ was developed as an alternative methodology for building statistical models. PCR+ is an extension of Principal Components Regression (PCR), in which the eigenvectors resulting from Principal Components Analysis (PCA) are used as predictor variables in regression analysis. The work was motivated by the observation that most heavy-duty diesel (HDD) engine research was conducted with test fuels that had been ''concocted'' in the laboratory to vary selected fuel properties in isolation from each other. This approach departs markedly from the real world, where the reformulation of diesel fuels for almost any purpose leads to changes in a number of interrelated properties. In this work, we present new information regarding the problems encountered in the conventional approach to model-building and how the PCR+ method can be used to improve research on the relationship between fuel characteristics and engine emissions. We also discuss how PCR+ can be applied to a variety of other research problems related to diesel fuels.

TRAC-PF1/MOD3 calculations of Savannah River Laboratory Rig FA single-annulus heated experiments

International Nuclear Information System (INIS)

Fischer, S.R.; McDaniel, C.K.

1992-01-01

This paper presents the results of TRAC-PF1/MOD3 benchmarks of the Rig FA experiments performed at the Savannah River Laboratory to simulate prototypic reactor fuel assembly behavior over a range of fluid conditions typical of the emergency cooling system (ECS) phase of a loss-of-coolant accident (LOCA). The primary purpose of this work was to use the SRL Rig FA tests to qualify the TRAC-PF1/MOD3 computer code and models for computing Mark-22 fuel assembly LOCA/ECS power limits. This qualification effort was part of a larger effort undertaken by the Los Alamos National Laboratory for the US Department of Energy to independently confirm power limits for the Savannah River Site K Reactor. The results of this benchmark effort as discussed in this paper demonstrate that TRAC/PF1/MOD3 coupled with proper modeling is capable of simulating thermal-hydraulic phenomena typical of that encountered in Mark-22 fuel assembly during LOCA/ECS conditions

Computer code for the thermal-hydraulic analysis of ITU TRIGA Mark-II reactor

International Nuclear Information System (INIS)

Ustun, G.; Durmayaz, A.

2002-01-01

Istanbul Technical University (ITU) TRIGA Mark-II reactor core consists of ninety vertical cylindrical elements located in five rings. Sixty-nine of them are fuel elements. The reactor is operated and cooled with natural convection by pool water, which is also cooled and purified in external coolant circuits by forced convection. This characteristic leads to consider both the natural and forced convection heat transfer in a 'porous-medium analysis'. The safety analysis of the reactor requires a thermal-hydraulic model of the reactor to determine the thermal-hydraulic parameters in each mode of operation. In this study, a computer code cooled TRIGA-PM (TRIGA - Porous Medium) for the thermal-hydraulic analysis of ITU is considered. TRIGA Mark-II reactor code has been developed to obtain velocity, pressure and temperature distributions in the reactor pool as a function of core design parameters and pool configuration. The code is a transient, thermal-hydraulic code and requires geometric and physical modelling parameters. In the model, although the reactor is considered as only porous medium, the other part of the reactor pool is considered partly as continuum and partly as porous medium. COMMIX-1C code is used for the benchmark purpose of TRIGA-PM code. For the normal operating conditions of the reactor, estimations of TRIGA-PM are in good agreement with those of COMMIX-1C. After some more improvements, this code will be employed for the estimation of LOCA scenario, which can not be analyses by COMMIX-1C and the other multi-purpose codes, considering a break at one of the beam tubes of the reactor

Exhaust emissions from an indirect injection dual-fuel engine

International Nuclear Information System (INIS)

Abd Alla, G.H.; Badr, O.A.; Soliman, H.A.; Abd Rabbo, M.F.

2000-01-01

Diesel engines operating on gaseous fuels are commonly known as dual-fuel engines. In the present work, a single-cylinder, compression ignition, indirect injection research (Ricardo E6) engine has been installed at United Arab Emirates University for investigation of the exhaust emissions when the engine is operating as a dual-fuel engine. The influence of changes in major operating and design parameters, such as the concentration of gaseous fuel in the cylinder charge, pilot fuel quantity, injection timing and intake temperature, on the production of exhaust emissions was investigated. Diesel fuel was used as the pilot fuel, while methane or propane was used as the main fuel which was inducted in the intake manifold and mixed with the intake air. The experimental investigations showed that the poor emissions at light loads can be improved significantly by increasing the concentration of gaseous fuel (total equivalence ratio), employing a large pilot fuel quantity, advancing the injection timing of the pilot fuel and increasing the intake temperature. It is demonstrated that, in general, any measure that tends to increase the size of the combustion regions within the overly lean cylinder charge will reduce markedly the concentrations of unburned hydrocarbons and carbon monoxide in the exhaust gases. (Author)