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Sample records for chalk river pool test reactor

  1. Reactor loops at Chalk River

    International Nuclear Information System (INIS)

    This report describes broadly the nine in-reactor loops, and their components, located in and around the NRX and NRU reactors at Chalk River. First an introduction and general description is given of the loops and their function, supplemented with a table outlining some loop specifications and nine simplified flow sheets, one for each individual loop. The report then proceeds to classify each loop into two categories, the 'main loop circuit' and the 'auxiliary circuit', and descriptions are given of each circuit's components in turn. These components, in part, are comprised of the main loop pumps, the test section, loop heaters, loop coolers, delayed-neutron monitors, surge tank, Dowtherm coolers, loop piping. Here again photographs, drawings and tables are included to provide a clearer understanding of the descriptive literature and to include, in tables, some specifications of the more important components in each loop. (author)

  2. Mortality study of Canadian military personnel exposed to radiation: atomic test blasts and Chalk River nuclear reactor clean-ups, 1950's

    International Nuclear Information System (INIS)

    This report describes a historical cohort study of the group of Canadian military personnel exposed to radiation in the 1950s at atomic bomb test blasts in the U.S. and Australia, and at clean-up operations at the Chalk River Nuclear Laboratories. Overall and cause-specific mortality in the exposed group was compared to that of the control cohort of unexposed military personnel, matched on age, service, rank and trade. Analyses indicated no elevation in the exposed cohort, in overall or cause-specific mortality due to diseases associated with radiation. Since this study was restricted to an investigation of mortality, we must stress that we cannot generalize these results or conclusions to current morbidity experienced by the exposed cohort

  3. Ecologically acceptable flows in Chalk rivers

    OpenAIRE

    Acreman, Mike; Dunbar, Michael

    2010-01-01

    The term ‘Chalk rivers’ is used to describe all those water courses dominated by groundwater discharge from Chalk geology. Natural conditions and historical modification have generated an ecosystem, with rich and unique assemblages and with high value to society (e.g. SACs, SSSIs, visual amenity and fisheries. Chalk rivers are considered to be sensitive to hydrological and morphological change and there is concern that flood defence and land drainage schemes, catchment agriculture, urbanisati...

  4. The results from the second high-pressure melt ejection test completed in the Molten Fuel Moderator Interaction Facility at Chalk River Laboratories

    Energy Technology Data Exchange (ETDEWEB)

    Nitheanandan, T.; Kyle, G.; O' Connor, R. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2007-09-15

    The Canadian nuclear power generation industry, represented by the CANDU Owners Group (COG), is funding an experimental program at Chalk River Laboratories to study the interaction between the molten material ejected from the fuel channel and the moderator. These experiments are designed to address one of the very low probability postulated accident events considered for CANDU Pressurized Heavy Water Reactors (PHWRs), where an array of fuel channels contain the nuclear fuel and high-temperature, high-pressure coolant. Under severely restricted flow blockage conditions postulated in a fuel channel, the temperature excursion could result in fuel melting, consequential failure of the fuel channel, and ejection of the molten fuel at high pressures into the heavy water moderator at near atmospheric pressure. The objective of the experimental program is to demonstrate that a highly energetic Molten Fuel Moderator Interaction (MFMI) and associated high-pressure pulse can be ruled out. The second high-pressure melt ejection test using 22 kg of prototypical corium was completed recently at Chalk River Laboratories. The second test consisted of heating a thermite mixture of U, U{sub 3}O{sub 8}, Zr, and CrO{sub 3}, simulating the molten material expected in a fuel channel, inside a 1 m length of insulated pressure tube. Once the molten material reached the desired temperature of {approx}2400{sup o}C, the molten material was ejected into the surrounding tank of 63{sup o}C water. At the time of melt ejection, the static pressure in the test section was 3.35 MPa. The confinement vessel pressure reached a peak volume of 201 kPa following the rupture of the test section. The peak dynamic pressure measured on the inner vessel walls ranged between 0.7 MPa and 1 MPa. The dynamic pressure history, debris size, and the effects of the material interacting with tubes representing neighbouring fuel channels were investigated. (author)

  5. The results from the second high-pressure melt ejection test completed in the Molten Fuel Moderator Interaction Facility at Chalk River Laboratories

    International Nuclear Information System (INIS)

    The Canadian nuclear power generation industry, represented by the CANDU Owners Group (COG), is funding an experimental program at Chalk River Laboratories to study the interaction between the molten material ejected from the fuel channel and the moderator. These experiments are designed to address one of the very low probability postulated accident events considered for CANDU Pressurized Heavy Water Reactors (PHWRs), where an array of fuel channels contain the nuclear fuel and high-temperature, high-pressure coolant. Under severely restricted flow blockage conditions postulated in a fuel channel, the temperature excursion could result in fuel melting, consequential failure of the fuel channel, and ejection of the molten fuel at high pressures into the heavy water moderator at near atmospheric pressure. The objective of the experimental program is to demonstrate that a highly energetic Molten Fuel Moderator Interaction (MFMI) and associated high-pressure pulse can be ruled out. The second high-pressure melt ejection test using 22 kg of prototypical corium was completed recently at Chalk River Laboratories. The second test consisted of heating a thermite mixture of U, U3O8, Zr, and CrO3, simulating the molten material expected in a fuel channel, inside a 1 m length of insulated pressure tube. Once the molten material reached the desired temperature of ∼2400oC, the molten material was ejected into the surrounding tank of 63oC water. At the time of melt ejection, the static pressure in the test section was 3.35 MPa. The confinement vessel pressure reached a peak volume of 201 kPa following the rupture of the test section. The peak dynamic pressure measured on the inner vessel walls ranged between 0.7 MPa and 1 MPa. The dynamic pressure history, debris size, and the effects of the material interacting with tubes representing neighbouring fuel channels were investigated. (author)

  6. Molten fuel moderator interaction program at Chalk River Laboratories

    Energy Technology Data Exchange (ETDEWEB)

    Nitheanandan, T.; Kyle, G.; O' Connor, R.; Sanderson, D.B. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2006-12-15

    The Canadian nuclear power generation industry, represented by the CANDU Owners Group (COG), has been funding an experimental program at Chalk River Laboratories (CRL) to study the interaction between molten material ejected from a fuel channel and the moderator. These experiments were designed to address one of the very low probability postulated accident events considered for CANDU Pressurized Heavy Water Reactors. The reactor consists of an array of horizontal fuel channels that contain the UO{sub 2}, nuclear fuel and high-temperature, high-pressure heavy water coolant. Under severely restricted flow blockage conditions, approaching 100% reduction of the flow area, postulated in a fuel channel, the temperature excursion could result in fuel melting, consequential failure of the fuel channel, and ejection of the molten fuel at high pressures into the heavy water moderator at near atmospheric pressure. In preparation for these tests, a chemical mixture called a thermite, that could produce a simulated molten fuel when ignited, was developed in partnership with Argonne National Laboratory (USA). Following this thermite development, two base-case reference tests were completed. The two base-case reference tests, with no molten material present, were performed in the Molten-Fuel Moderator-Interaction (MFMI) facility at CRL. Following the base-case reference tests, a high-pressure melt ejection test using prototypical corium was conducted. The objectives of this paper are to provide an overview of the MFMI program and present the results obtained from thermite development, base-case and melt ejection experiments. (author)

  7. Molten fuel moderator interaction program at Chalk River Laboratories

    International Nuclear Information System (INIS)

    The Canadian nuclear power generation industry, represented by the CANDU Owners Group (COG), has been funding an experimental program at Chalk River Laboratories (CRL) to study the interaction between molten material ejected from a fuel channel and the moderator. These experiments were designed to address one of the very low probability postulated accident events considered for CANDU Pressurized Heavy Water Reactors. The reactor consists of an array of horizontal fuel channels that contain the UO2, nuclear fuel and high-temperature, high-pressure heavy water coolant. Under severely restricted flow blockage conditions, approaching 100% reduction of the flow area, postulated in a fuel channel, the temperature excursion could result in fuel melting, consequential failure of the fuel channel, and ejection of the molten fuel at high pressures into the heavy water moderator at near atmospheric pressure. In preparation for these tests, a chemical mixture called a thermite, that could produce a simulated molten fuel when ignited, was developed in partnership with Argonne National Laboratory (USA). Following this thermite development, two base-case reference tests were completed. The two base-case reference tests, with no molten material present, were performed in the Molten-Fuel Moderator-Interaction (MFMI) facility at CRL. Following the base-case reference tests, a high-pressure melt ejection test using prototypical corium was conducted. The objectives of this paper are to provide an overview of the MFMI program and present the results obtained from thermite development, base-case and melt ejection experiments. (author)

  8. The results from the second high-pressure melt ejection test completed in the molten fuel moderator interaction facility at Chalk River Laboratories

    International Nuclear Information System (INIS)

    For a Candu reactor and under severely restricted flow blockage conditions postulated in a fuel channel, the temperature excursion could result in fuel melting, consequential failure of the fuel channel, and ejection of the molten fuel at high pressures into the heavy water moderator at near atmospheric pressure. The objective of the experimental program is to demonstrate that a highly energetic Molten Fuel Moderator Interaction (MFMI) and associated high-pressure pulse can be ruled out for a Candu reactor. The second high-pressure melt ejection test using 22 kg of prototypical corium was completed recently at Chalk River Laboratories. The second test consisted in heating a thermite mixture of U, U3O8, Zr, and CrO3, simulating the molten material expected in a fuel channel, inside a 1 m length of insulated pressure tube. Once the molten material reached the desired temperature of about 2400 C degrees, the molten material was ejected into the surrounding tank of 63 C water. At the time of melt ejection, the static pressure in the test section was 3.35 MPa. The confinement vessel pressure reached a peak value of 201 kPa following the rupture of the test section. The peak dynamic pressure measured on the inner vessel walls ranged between 0.7 MPa and 1 MPa. The total debris collected inside the tank was 22.65 kg. The debris inside the inner tank consists of corium, quartz and Zircar. The majority of the corium particles were less than 1 mm in diameter and the calculated value of the mean size of the debris appears to be 0.581 mm. An analysis of the confinement vessel gas inventory indicated 17.6% hydrogen

  9. The results from the second high-pressure melt ejection test completed in the molten fuel moderator interaction facility at Chalk River Laboratories

    Energy Technology Data Exchange (ETDEWEB)

    Nitheanandan, T.; Kyle, G.; O' Connor, R. [Chalk River Laboratories, Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2007-07-01

    For a Candu reactor and under severely restricted flow blockage conditions postulated in a fuel channel, the temperature excursion could result in fuel melting, consequential failure of the fuel channel, and ejection of the molten fuel at high pressures into the heavy water moderator at near atmospheric pressure. The objective of the experimental program is to demonstrate that a highly energetic Molten Fuel Moderator Interaction (MFMI) and associated high-pressure pulse can be ruled out for a Candu reactor. The second high-pressure melt ejection test using 22 kg of prototypical corium was completed recently at Chalk River Laboratories. The second test consisted in heating a thermite mixture of U, U{sub 3}O{sub 8}, Zr, and CrO{sub 3}, simulating the molten material expected in a fuel channel, inside a 1 m length of insulated pressure tube. Once the molten material reached the desired temperature of about 2400 C degrees, the molten material was ejected into the surrounding tank of 63 C water. At the time of melt ejection, the static pressure in the test section was 3.35 MPa. The confinement vessel pressure reached a peak value of 201 kPa following the rupture of the test section. The peak dynamic pressure measured on the inner vessel walls ranged between 0.7 MPa and 1 MPa. The total debris collected inside the tank was 22.65 kg. The debris inside the inner tank consists of corium, quartz and Zircar. The majority of the corium particles were less than 1 mm in diameter and the calculated value of the mean size of the debris appears to be 0.581 mm. An analysis of the confinement vessel gas inventory indicated 17.6% hydrogen.

  10. An Investigation into the Transportation of Irradiated Uranium/Aluminum Targets from a Foreign Nuclear Reactor to the Chalk River Laboratories Site in Ontario, Canada - 12249

    International Nuclear Information System (INIS)

    This investigation required the selection of a suitable cask and development of a device to hold and transport irradiated targets from a foreign nuclear reactor to the Chalk River Laboratories in Ontario, Canada. The main challenge was to design and validate a target holder to protect the irradiated HEU-Al target pencils during transit. Each of the targets was estimated to have an initial decay heat of 118 W prior to transit. As the targets have little thermal mass the potential for high temperature damage and possibly melting was high. Thus, the primary design objective was to conceive a target holder to dissipate heat from the targets. Other design requirements included securing the targets during transportation and providing a simple means to load and unload the targets while submerged five metres under water. A unique target holder (patent pending) was designed and manufactured together with special purpose experimental apparatus including a representative cask. Aluminum dummy targets were fabricated to accept cartridge heaters, to simulate decay heat. Thermocouples were used to measure the temperature of the test targets and selected areas within the target holder and test cask. After obtaining test results, calculations were performed to compensate for differences between experimental and real life conditions. Taking compensation into consideration the maximum target temperature reached was 231 deg. C which was below the designated maximum of 250 deg. C. The design of the aluminum target holder also allowed generous clearance to insert and unload the targets. This clearance was designed to close up as the target holder is placed into the cavity of the transport cask. Springs served to retain and restrain the targets from movement during transportation as well as to facilitate conductive heat transfer. The target holder met the design requirements and as such provided data supporting the feasibility of transporting targets over a relatively long period of time

  11. Anthropogenic radionuclides in Ottawa River sediment near Chalk River Laboratories

    International Nuclear Information System (INIS)

    The Ottawa River has received nuclear reactor effluent from Chalk River Laboratories (CRL) for more than 60 years, including releases from a NRX accident in 1952. Recent interest in the potential impact of these historical releases and the possible need for remediation of a small region immediately downstream from the release point has led to comprehensive studies to assess risk to people and wildlife. In this paper, the results of an extensive survey of gamma-emitting anthropogenic radionuclides in Ottawa River sediment in the vicinity of CRL are presented. Anthropogenic radionuclides detected in Ottawa River sediment include 60Co, 94Nb, 137Cs, 152Eu, 154Eu, 155Eu and 241Am. Concentrations of all anthropogenic radionuclides decline rapidly with distance downstream of the process outfall, reaching stable concentrations about 2 km downstream. All of these radionuclides are found at some sites within 2 km upstream of the process outfall suggesting limited upstream transport and sedimentation. Comparison of anthropogenic radionuclides with several representative primordial radionuclides shows that with the exception of sites at the process outfall and within 2 km downstream of the process outfall, primordial radionuclide concentrations greatly exceed CRL derived anthropogenic radionuclide concentrations. Thus, over 60 years of radionuclide releases from operations at CRL have had little impact on radionuclide concentrations in Ottawa River sediment, except at a few sites immediately adjacent to the process outfall. (author)

  12. Development and irradiation testing of Al-U3Si2 at Chalk River Laboratories

    International Nuclear Information System (INIS)

    Mini-elements containing Al-64 wt% U3Si2 (3.15 gU/cm3), with three discrete U3Si2 particle-size distributions, have been irradiated up to 93 at% burnup in the NRU reactor. The uranium silicide (U-7.0Si) was used in the as-cast condition, and contained up to 4 wt% free uranium in the U3Si2 matrix. Post-irradiation examinations (PIE) of the high-burnup elements have been recently completed. PIE included underwater and hot-cell examinations, immersion density measurements, neutron radiography, optical and scanning electron microscopy (SEM) with wavelength dispersion X-ray (WDX) analysis, and computerized image analysis of the fission-gas bubble-size distributions. The results show that the Al-U3Si2 swelled less than Al-U3Si fuel previously irradiated under similar conditions in NRU, and no significant swelling dependence on particle-size distribution was observed. Al-U3Si2 core volume increases ranged from 4.2 to 4.7 vol%, compared to 5.8 to 6.8 vol% for Al-U3Si fuel with identical uranium loadings. SEM examinations revealed that the U3Si2 (U-7.0Si) particles contained regions with relatively ordered, very dense populations of sub-micron fission-gas bubbles. In contrast, the gas bubbles are randomly distributed within U3Si (U-3.96Si) particles, vary widely in size, and small bubbles coalesce to form larger bubbles. The capability of U3Si2 to retain fission gas in small bubbles accounts for the lower swelling. (author)

  13. Results from the fourth high-pressure melt ejection test completed in the molten fuel moderator interaction facility at Chalk River Laboratories

    International Nuclear Information System (INIS)

    The fourth high-pressure melt ejection test using prototypical corium was completed at Chalk River Laboratories. This test was one of four tests planned by Atomic Energy of Canada Limited to study the potential for energetic interaction between molten fuel and water. The experiments were designed to address one of the very low probability postulated accident events considered for Candu Pressurized Heavy Water Reactors (PHWRs). The accident event considered is the severe reduction in the coolant flow to a single channel. This reduction could result from a blockage in the flow or a break in the inlet piping to a fuel channel. If the reduction in the flow is severe (approaching complete cessation of the flow), the fuel channel will overheat and fail. Such a failure is not predicted to propagate to other fuel channels; the scenario is terminated with the emergency coolant injection. Under severely restricted flow blockage conditions, the temperature excursion could result in fuel melting. Conservative safety analysis assessments consider the implications of the worst-case scenario, which can involve the ejection of the molten material from the fuel channel into the heavy-water moderator. The predictions are that the melt will be finely fragmented and will transfer energy to the moderator as it is dispersed, creating a modest pressure pulse in the calandria vessel. The high-pressure melt ejection experiments funded by the Candu Owners Group have been performed to confirm these predictions and to show that a highly energetic 'steam explosion, ' and associated high-pressure pulse, is not possible. The high-pressure melt ejection test described here consisted of heating 12.5 kg of a thermite mixture U, UO2, Zr, and CrO3, representing the molten material in a fuel channel, inside an insulated pressure tube. When the molten material reached the desired temperature of ∼2400 deg.C, the pressure inside the tube was raised to about 10.5 MP a, and the pressure tube failed due

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

    International Nuclear Information System (INIS)

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

  15. Experience at Chalk River with a cw electron accelerator

    International Nuclear Information System (INIS)

    For several years a group at Chalk River has been studying the behaviour of structures operated in the cw mode under heavy beam loading. Three side-coupled structures, modelled on the LAMPF design, have been built and tests up to 50% beam loading have been performed on two of them. Control systems have been developed to regulate the disturbances arising from high average power in a multi-tank accelerator and procedures worked out to handle beam currents up to 20 mA at 4 MeV. A pancake-coupled structure has been designed for high power operation and results of low power tests on an aluminum model are presented. Tests at high power with a 50 mA electron beam are planned. (author)

  16. Model description of CHERPAC (Chalk River Environmental Research Pathways Analysis Code); results of testing with post-Chernobyl data from Finland

    International Nuclear Information System (INIS)

    CHERPAC (Chalk River Environmental Research Pathways Analysis Code), a time-dependent code for assessing doses from accidental and routine releases of radionuclides, has been under development since 1987. A complete model description is provide here with equations, parameter values, assumptions and information on parameter distributions for uncertainty analysis. Concurrently, CHERPAC has been used to participate in the two internal model validation exercises BIOMOVS (BIOspheric MOdel Validation Study) and VAMP (VAlidation of Assessment Model Predictions, a co-ordinated research program of the International Atomic Energy Agency). CHERPAC has been tested for predictions of concentrations of 137Cs in foodstuffs, body burden and dose over time using data collected after the Chernobyl accident of 1986 April. CHERPAC's results for the recent VAMP scenario for southern Finland are particularly accurate and should represent what the code can do under Canadian conditions. CHERPAC's predictions are compared with the observations from Finland for four and one-half years after the accident as well as with the results of the other participating models from nine countries. (author). 18 refs., 23 figs., 2 appendices

  17. Backfitting swimming pool reactors

    International Nuclear Information System (INIS)

    Calculations based on measurements in a critical assembly, and experiments to disclose fuel element surface temperatures in case of accidents like stopping of primary coolant flow during full power operation, have shown that the power of the swimming pool type research reactor FRG-2 (15 MW, operating since 1967) might be raised to 21 MW within the present rules of science and technology, without major alterations of the pool buildings and the cooling systems. A backfitting program is carried through to adjust the reactor control systems of FRG-2 and FRG-1 (5 MW, housed in the same reactor hall) to the present safety rules and recommendations, to ensure FRG-2 operation at 21 MW for the next decade. (author)

  18. Widespread methanotrophic primary production in lowland chalk rivers

    OpenAIRE

    Shelley, Felicity; Grey, Jonathan; Trimmer, Mark

    2014-01-01

    Methane is oversaturated relative to the atmosphere in many rivers, yet its cycling and fate is poorly understood. While photosynthesis is the dominant source of autotrophic carbon to rivers, chemosynthesis and particularly methane oxidation could provide alternative sources of primary production where the riverbed is heavily shaded or at depth beneath the sediment surface. Here, we highlight geographically widespread methanotrophic carbon fixation within the gravel riverbeds of over 30 chalk...

  19. Thermal hydraulics in the hot pool of Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Sodium cooled Fast Breeder Test Reactor (FBTR) of 40 MWt/13 MWe capacity is in operation at Kalpakkam, near Chennai. Presently it is operating with a core of 10.5 MWt. Knowledge of temperatures and flow pattern in the hot pool of FBTR is essential to assess the thermal stresses in the hot pool. While theoretical analysis of the hot pool has been conducted by a three-dimensional code to access the temperature profile, it involves tuning due to complex geometry, thermal stresses and vibration. With this in view, an experimental model was fabricated in 1/4 scale using acrylic material and tests were conducted in water. Initially hydraulic studies were conducted with ambient water maintaining Froude number similarity. After that thermal studies were conducted using hot and cold water maintaining Richardson similitude. In both cases Euler similarity was also maintained. Studies were conducted simulating both low and full power operating conditions. This paper discusses the model simulation, similarity criteria, the various thermal hydraulic studies that were carried out, the results obtained and the comparison with the prototype measurements.

  20. Contaminated groundwater characterization at the Chalk River Laboratories, Ontario, Canada

    Energy Technology Data Exchange (ETDEWEB)

    Schilk, A.J.; Robertson, D.E.; Thomas, C.W.; Lepel, E.A. [Pacific Northwest National Lab., Richland, WA (United States); Champ, D.R.; Killey, R.W.D.; Young, J.L.; Cooper, E.L. [Chalk River Labs., Chalk River, Ontario (Canada)

    1993-03-01

    The licensing requirements for the disposal of low-level radioactive waste (10 CFR 61) specify the performance objectives and technical requisites for federal and commercial land disposal facilities, the ultimate goal of which is to contain the buried wastes so that the general population is adequately protected from harmful exposure to any released radioactive materials. A major concern in the operation of existing and projected waste disposal sites is subterranean radionuclide transport by saturated or unsaturated flow, which could lead to the contamination of groundwater systems as well as uptake by the surrounding biosphere, thereby directly exposing the general public to such materials. Radionuclide transport in groundwater has been observed at numerous commercial and federal waste disposal sites [including several locations within the waste management area of Chalk River Laboratories (CRL)], yet the physico-chemical processes that lead to such migration are still not completely understood. In an attempt to assist in the characterization of these processes, an intensive study was initiated at CRL to identify and quantify the mobile radionuclide species originating from three separate disposal sites: (a) the Chemical Pit, which has received aqueous wastes containing various radioisotopes, acids, alkalis, complexing agents and salts since 1956, (b) the Reactor Pit, which has received low-level aqueous wastes from a reactor rod storage bay since 1956, and (c) the Waste Management Area C, a thirty-year-old series of trenches that contains contaminated solid wastes from CRL and various regional medical facilities. Water samples were drawn downgradient from each of the above sites and passed through a series of filters and ion-exchange resins to retain any particulate and dissolved or colloidal radionuclide species, which were subsequently identified and quantified via radiochemical separations and gamma spectroscopy. These groundwaters were also analyzed for anions

  1. Inventory of radioactivity in Ottawa River-bed sediments near the Chalk River Laboratories

    International Nuclear Information System (INIS)

    AECL's Chalk River Laboratories (CRL) are situated on the Ontario side of the Ottawa River about 200 km NW of the City of Ottawa. Since 1947, water for cooling CRL's research reactors has been piped from and returned to the Ottawa River. From 1952 to the present time, cooling water has been discharged through the Process Sewer at a rate of 1.5 to 2 m3/s. The Outfall, which is the discharge from the Process Sewer, is in 18 m of water, 65 m offshore. Flow is directed toward the river surface through three 'diffuser vents,' creating a turbulent swirl at the surface and maintaining a patch of open water in winter. In addition to cooling water, the Outfall has, over the years, included small additional effluents from a heavy water recovery plant, a decontamination centre and a waste treatment centre. Although the effluent has been monitored and has met applicable regulatory requirements, investigations of the riverbed near the Outfall revealed radioactivity. In 2001, a riverbed reconnaissance and a detailed coring program were initiated for the purpose of determining the inventory of residual radioactivity. (author)

  2. The role of alluvial valley deposits in groundwater–surface water exchange in a Chalk river

    OpenAIRE

    Abesser, Corinna; Shand, Paul; Gooddy, Daren; Peach, Denis

    2008-01-01

    To understand the processes of surface water–groundwater exchange in Chalk catchments, a detailed hydrogeochemical study was carried out in the Lambourn catchment in southeast England. Monthly monitoring of river flow and groundwater levels and water chemistry has highlighted a large degree of heterogeneity at the river-corridor scale. The data suggest an irregular connection between the river, the alluvial deposits, and the Chalk aquifer at the study site. The groundwaters in the alluvial gr...

  3. Computational simulation of the natural circulation occurring in an experimental test section of a pool type research reactor

    International Nuclear Information System (INIS)

    The present work presents a computational simulation of the natural circulation phenomenon developing in an experimental test section of a pool type research reactor. The test section has been designed using a reduced scale in height 1:4.7 in relation to a pool type 30 MW research reactor prototype. It comprises a cylindrical vessel, which is opened to atmosphere, and representing the reactor pool; a natural circulation pipe, a lower plenum, and a heater containing electrical resistors in rectangular plate format, which represents the fuel elements, with a chimney positioned on the top of the resistor assembly. In the computational simulation, it was used a commercial CFD software, without any turbulence model. Besides, in the presence of the natural circulation, a laminar flow has been assumed and the equations of the mass conservation, momentum and energy were solved by the finite element method. In addition, the results of the simulation are presented in terms of velocities and temperatures differences, respectively: at inlet and outlet of the heater and of the natural circulation pipe. (author)

  4. Isotope hydrology of the Chalk River Laboratories site, Ontario, Canada

    Science.gov (United States)

    Peterman, Zell; Neymark, Leonid; King-Sharp, K.J.; Gascoyne, Mel

    2016-01-01

    This paper presents results of hydrochemical and isotopic analyses of groundwater (fracture water) and porewater, and physical property and water content measurements of bedrock core at the Chalk River Laboratories (CRL) site in Ontario. Density and water contents were determined and water-loss porosity values were calculated for core samples. Average and standard deviations of density and water-loss porosity of 50 core samples from four boreholes are 2.73 ± 12 g/cc and 1.32 ± 1.24 percent. Respective median values are 2.68 and 0.83 indicating a positive skewness in the distributions. Groundwater samples from four deep boreholes were analyzed for strontium (87Sr/86Sr) and uranium (234U/238U) isotope ratios. Oxygen and hydrogen isotope analyses and selected solute concentrations determined by CRL are included for comparison. Groundwater from borehole CRG-1 in a zone between approximately +60 and −240 m elevation is relatively depleted in δ18O and δ2H perhaps reflecting a slug of water recharged during colder climatic conditions. Porewater was extracted from core samples by centrifugation and analyzed for major dissolved ions and for strontium and uranium isotopes. On average, the extracted water contains 15 times larger concentration of solutes than the groundwater. 234U/238U and correlation of 87Sr/86Sr with Rb/Sr values indicate that the porewater may be substantially older than the groundwater. Results of this study show that the Precambrian gneisses at Chalk River are similar in physical properties and hydrochemical aspects to crystalline rocks being considered for the construction of nuclear waste repositories in other regions.

  5. Clinch River Breeder Reactor Plant steam generator: FEW tube test model post test examination

    International Nuclear Information System (INIS)

    The Steam Generator Few Tube Test (FTT) is part of an extensive testing program being carried out in support of the Clinch River Breeder Reactor Plant (CRBRP) steam generator design. The testing of full-length seven-tube evaporator and three-tube superheater models of the CRBRP design was conducted to provide steady-state thermal/hydraulic performance data to full power per tube and to verify the absence of multi-year endurance problems. The problems encountered with the mechanical features of the FTT model design which led to premature test termination and the results of the post-test examination are described

  6. Edibility of sport fishes in the Ottawa River near Chalk River Laboratories

    International Nuclear Information System (INIS)

    To address the question of edibility of fish in the Ottawa River near Chalk River Laboratories (CRL), 123 game fish were collected for analysis from four locations: Mackey and Rolphton (45 km and 35 km upstream of Chalk River Laboratories (CRL), respectively), the Sandspit (Pointe au Bapteme) and Cotnam Island (1.6 km and 45 km downstream of CRL, respectively). Twenty-six to thirty-six game fish were collected at each location in 2007 and samples of flesh or bone were analyzed. Trap nets were used to collect only the fish required, allowing release of management-sensitive species. The focus was on walleye (Sander vitreus) because they are abundant and popular among anglers. A few northern pike (Esox lucius) and a smaller number of smallmouth bass (Micropterus dolomieui) were also collected at three of the four sites. Samples of the fish were analyzed for cesium-137 (137Cs), strontium-90 (90Sr), mercury (Hg), and selected organo-chlorine compounds. Concentrations of 137Cs in the flesh and 90Sr in the bones of sport fish were low and similar at all four locations and appear to reflect the global residuals from nuclear weapons testing (primarily in the 1960's) as opposed to releases from CRL. Possible explanations are: 1) Reductions in radionuclide releases from CRL in recent decades and 2) Relatively large foraging ranges of sport fish. Mercury concentrations were elevated in fishes in the Ottawa River and were significantly higher at the Sandspit and Rolphton than at Mackey and Cotnam Island (p<0.001). Mercury concentrations from the four sites are comparable to concentrations in other Ontario and Quebec lakes. It is advisable therefore, that consumers follow the fish consumption guidelines issued by provincial authorities when eating fish from the Ottawa River. Organo-chlorine compounds were not detected in walleye; however, they were detected in all eight of the pike collected at Cotnam Island. The highest organo-chlorine concentrations were measured in two of the

  7. Determination of neutron energy spectrum at a pneumatic rabbit station of a typical swimming pool type material test research reactor

    International Nuclear Information System (INIS)

    The method of multiple foil activation was used to measure the neutron energy spectrum, experimentally, at a rabbit station of Pakistan Research Reactor-1 (PARR-1), which is a typical swimming pool type material test research reactor. The computer codes MSITER and SANDBP were used to adjust the spectrum. The pre-information required by the adjustment codes was obtained by modelling the core and its surroundings in three-dimensions by using the one dimensional transport theory code WIMS-D/4 and the multidimensional finite difference diffusion theory code CITATION. The input spectrum covariance information required by MSITER code was also calculated from the CITATION output. A comparison between calculated and adjusted spectra shows a good agreement

  8. Use of borehole-geophysical logs and hydrologic tests to characterize crystalline rock for nuclear-waste storage, Whiteshell Nuclear Research Establishment, Manitoba, and Chalk River Nuclear Laboratory, Ontario, Canada

    International Nuclear Information System (INIS)

    A number of borehole methods were used in the investigation of crystalline rocks at Whiteshell Nuclear Research Establishment and Chalk River Nuclear Laboratory in Canada. The selection of a crystalline-rock mass for the storage of nuclear waste likely will require the drilling and testing of a number of deep investigative boreholes in the rock mass. Although coring of at least one hole in each new area is essential, methods for making in-situ geophysical and hydrologic measurements can substitute for widespread coring and result in significant savings in time and money. Borehole-geophysical logging techniques permit the lateral extrapolation of data from a core hole. Log response is related to rock type, alteration, and the location and character of fractures. The geophysical logs that particularly are useful for these purposes are the acoustic televiewer and acoustic waveform, neutron and gamma, resistivity, temperature, and caliper. The acoustic-televiewer log of the borehole wall can provide high resolution data on the orientation and apparent width of fractures. In situ hydraulic tests of single fractures or fracture zones isolated by packers provide quantitative information on permeability, extent, and interconnection. The computer analysis of digitized acoustic waveforms has identified a part of the waveform that has amplitude variations related to permeabilities measured in the boreholes by packer tests. 38 refs., 37 figs., 4 tabs

  9. Clinch River Breeder Reactor Plant Steam Generator Few Tube Test model post-test examination

    International Nuclear Information System (INIS)

    The Steam Generator Few Tube Test (FTT) was part of an extensive testing program carried out in support of the Clinch River Breeder Reactor Plant (CRBRP) steam generator design. The testing of full-length seven-tube evaporator and three-tube superheater models of the CRBRP design was conducted to provide steady-state thermal/hydraulic performance data to full power per tube and to verify the absence of multi-year endurance problems. This paper describes the problems encountered with the mechanical features of the FTT model design which led to premature test termination, and the results of the post-test examination. Conditions of tube bowing and significant tube and tube support gouging was observed. An interpretation of the visual and metallurgical observations is also presented. The CRBRP steam generator has undergone design evaluations to resolve observed deficiences found in the FFTM

  10. Proceedings of a workshop on geophysical and related geoscientific research at Chalk River, Ontario

    International Nuclear Information System (INIS)

    A large part of the Canadian Nuclear Fuel Waste Management Program is geoscience research and development aimed at obtaining information to quantify the transport of radionuclides through the geosphere and at determining the geotechnical properties required for disposal vault design. The geosphere at potential disposal sites is characterized in part by the use of remote sensing (geophysical) methods. In 1977 public concern about the disposal of radioactive waste resulted in field work being restricted to the site of Chalk River Nuclear Laboratories, which was used to develop, evaluate and compare various techniques in order to optimize the methods for obtaining geoscience information. Methods tested at Chalk River are to be applied at other research sites. Most investigations have been carried out around Maskinonge Lake, using about thirty boreholes sink into bedrock. The boreholes provide subsurface geological information that can be used as a reference to compare the responses of various geophysical methods and equipment. Regional studies, including airborne geophysical surveys, have also been conducted. The 25 papers presented at this workshop provide comprehensive documentation of the most significant results of geophysical studies. The workshop also provided an evaluation of geophysical techniques and their utility to the Nuclear Fuel Waste Management Program

  11. The Canadian HT dispersion experiment at Chalk River - June 1987

    International Nuclear Information System (INIS)

    A trace amount (3.54TBq) of tritiated hydrogen, HT, was released to the atmosphere at an experimental field at the Chalk River Laboratories on June 10, 1987 in order to study the environmental behaviour of HT. Experimental results showed that direct oxidation of HT in the atmosphere was small and confirmed that surface soils convert atmospheric HT to HTO. The HTO formed in the soil was slowly emitted to the atmosphere giving rise to the small concentrations of HTO observed in the air during the release and for a period of several weeks thereafter. HTO/HT ratios in air during the plume passage increased with downwind distance from a value of order 10-5 at 5 m to values between 4 x 10-4 and 8 x 10-4 at 400 m. Deposition velocities for HT to soil were in the range 10-4 to 10-3 m s-1. Rates of reemission of tritium from the soil to the atmosphere were typically a few percent per hour within one to two days of the release, declining to less than one percent per hour over two weeks. Tritium deposition velocities and reemission rates determined for soils in the field agreed well with laboratory measurements on field samples, and were similar in range to previous exposure chamber experiments carried out in various countries in the laboratory and field under non-winter conditions. Direct uptake of HT by vegetation was not detected. The time history of vegetation tritium was consistent with uptake of HTO from soil and atmosphere and with incorporation of tritium into the organically bound form through photosynthesis. The experiment provides an extensive data base suitable for the detailed evaluation of mathematical models describing the short range dispersion of tritium. The results indicate that the short range dose from a release of HT would be much less than the dose from an equivalent release of HTO

  12. Management of legacy spent nuclear fuel wastes at the Chalk River Laboratories: the challenges and innovative solutions implemented

    International Nuclear Information System (INIS)

    AECL has operated research reactors at the Chalk River Laboratories (CRL) site since 1947, for the purpose of nuclear energy and scientific research and for the production of radioisotopes. During the 1950s and 60s, a variety of spent nuclear fuel wastes were produced by irradiating metallic uranium and other prototype fuels. These legacy waste fuels were initially stored in water-filled fuel storage bays for a period of several years before being placed in storage containers and transferred to the CRL Waste Management Areas (WMAs), where they have been stored in below-grade, vertical cylindrical steel and concrete structures called 'tile holes'. (author)

  13. Review of Savannah River Site K Reactor inservice inspection and testing restart program

    International Nuclear Information System (INIS)

    Inservice inspection (ISI) and inservice testing (IST) programs are used at commercial nuclear power plants to monitor the pressure boundary integrity and operability of components in important safety-related systems. The Department of Energy (DOE) - Office of Defense Programs (DP) operates a Category A (> 20 MW thermal) production reactor at the Savannah River Site (SRS). This report represents an evaluation of the ISI and IST practices proposed for restart of SRS K Reactor as compared, where applicable, to current ISI/IST activities of commercial nuclear power facilities

  14. Groundwater monitoring and plume discharge zone characterization for the NRX radiostrontium plume at Chalk River Laboratories

    Energy Technology Data Exchange (ETDEWEB)

    Olfert, J.M.; Audet, M.; Killey, D., E-mail: olfertjm@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-12-15

    Groundwater is the principal pathway for the migration of most radiological and non-radiological compounds from past and present operating areas at Atomic Energy of Canada Limited's Chalk River Laboratories (CRL). The CRL Groundwater Monitoring Program (GWMP) was established to measure the groundwater quality around the perimeters of areas affected, or potentially affected, by groundwater plumes. One of these is the NRX Rod Bays plume, a legacy plume that originated from the fuel storage bays of the National Research Experimental (NRX) reactor. This plume contains primarily {sup 90}Sr migrating along the groundwater flow system to the Ottawa River. A characterization study of the shoreline region was completed recently to map the plume discharge zone by collecting samples from mini-piezometers and groundwater seeps (springs) during a period of low river level. Analysis of discharging groundwaters determined that the {sup 90}Sr concentrations were very similar to those sampled from nearby (upgradient) GWMP monitoring wells. With this favorable correlation, the high density of seep and mini-piezometer sampling along the shoreline allowed refinements to be made in defining the northerly and southerly boundaries of the radiostrontium plume. The seep and mini-piezometer sampling also provided evidence that the monitoring wells sampled routinely within the CRL GWMP are positioned appropriately for providing representative sampling of the plume. Shoreline seep and mini-piezometer sampling can lead to refinements in the conceptual site model for plumes with limited effort and cost. The supplemental characterization work can also potentially identify other targets for routine groundwater monitoring. (author)

  15. Nondestructive testing of PWR type fuel rods by eddy currents and metrology in the OSIRIS reactor pool

    International Nuclear Information System (INIS)

    The Saclay Reactor Department has developed a nondestructive test bench, now installed above channel 1 of the OSIRIS reactor. As part of investigations into the dynamics of PWR fuel degradation, a number of fuel rods underwent metrological and eddy current inspection, after irradiation

  16. Canada, Atomic Energy of Canada Limited (AECL), Chalk River Labs: Reuse and Licence Termination of a Number of Facilities at the Chalk River Labs to Allow for Refurbishment of the Site. Annex A. I-1

    International Nuclear Information System (INIS)

    Chalk River Labs is located along the Ottawa River in Ontario, Canada, approximately 200 km north-west of Ottawa. The site began construction in 1944 following the expropriation of approximately 1 500 ha of land. A number of research reactors were constructed at the site along with numerous nuclear labs, hot cells and administrative facilities in support of the research and development work planned for the site. The principal occupants of the Chalk River site are AECL employees with a strong presence from National Resources Canada (NRC) and other small research groups. The site is undergoing substantial changes with an emphasis on minimizing the impact of increasing the builtup area footprint in conjunction with site upgrades and new build projects. To accomplish this task, a number of refurbishment and decommissioning projects were planned. Decommissioning projects were initiated to make room for new development through a number of initiatives. The decommissioning mandate includes the removal of a select group of original deteriorating facilities to make room for new construction and to decommission other facilities to facilitate redevelopment and reuse of the available space. In Canada, the Canadian Nuclear Safety Commission (CNSC) issues nuclear licences. The licensees must demonstrate that it is safe to continue operations of the nuclear site and request a renewal of their licence. CNSC will issue a new operating licence for a specific period of time at which the licensee must demonstrate that it is safe to proceed with a licence renewal. A request to terminate a licensable activity must be submitted to the CNSC. Upon approval to proceed, it must be demonstrated that the licensable activities have ceased and the facility has been appropriately decommissioned. Licence termination requires a demonstration that the land or previous activities presents a low risk and that the process can be used to support redevelopment because it results in a scrutinized

  17. Current status of the waste identification program at AECL's Chalk River Laboratories

    International Nuclear Information System (INIS)

    The management of routine operating waste by Waste Management and Decommissioning (WM and D) at Atomic Energy of Canada Limited's (AECL) Chalk River Laboratories (CRL) is supported by the Waste Identification (WI) Program. The principal purpose of the WI Program is to minimize the cost and the effort associated with waste characterization and waste tracking, which are needed to optimize waste handling, storage and disposal. The major steps in the WI Program are: (1) identify and characterize the processes that generate the routine radioactive wastes accepted by WM and D - radioisotope production, radioisotope use, reactor operation, fuel fabrication, et cetera (2) identify and characterize the routine blocks of waste generated by each process or activity - the initial characterization is based on inference (process knowledge) (3) prepare customized, template data sheets for each routine waste block - templates contain information such as package type, waste material, waste type, solidifying agent, the average non-radiological contaminant inventory, the average radiological contaminant inventory, and the waste class (4) ensure generators 'use the right piece of paper with the right waste' when they transfer waste to WM and D - that is they use the correct template data sheets to transfer routine wastes, by: identifying and marking waste collection points in the generator's facility; ensuring that generators implement effective waste collection/segregation procedures; implementing standard procedures to transfer waste to WM and D; and, auditing waste collection and segregation within a generator's facility (5) determine any additional waste block characterization requirements (is anything needed beyond the original characterization by process knowledge?) This paper describes the WI Program, it provides an example of its implementation, and it summarizes the current status of its implementation for both CRL and non-CRL waste generators. (author)

  18. Response of invertebrates from the hyporheic zone of chalk rivers to eutrophication and land use.

    Science.gov (United States)

    Pacioglu, Octavian; Moldovan, Oana Teodora

    2016-03-01

    Whereas the response of lotic benthic macroinvertebrates to different environmental stressors is a widespread practice nowadays in assessing the water and habitat quality, the use of hyporheic zone invertebrates is still in its infancy. In this study, classification and regression trees analysis were employed in order to assess the ecological requirements and the potential as bioindicators for the hyporheic zone invertebrates inhabiting four lowland chalk rivers (south England) with contrasting eutrophication levels (based on surface nitrate concentrations) and magnitude of land use (based on percentage of fine sediments load and median interstitial space). Samples of fauna, water and sediment were sampled twice, during low (summer) and high (winter) groundwater level, at depths of 20 and 35 cm. Certain groups of invertebrates (Glossosomatidae and Psychomyiidae caddisflies, and riffle beetles) proved to be good indicators of rural catchments, moderately eutrophic and with high fine sediment load. A diverse community dominated by microcrustaceans (copepods and ostracods) were found as good indicators of highly eutrophic urban streams, with moderate-high fine sediment load. However, the use of other taxonomic groups (e.g. chironomids, oligochaetes, nematodes, water mites and the amphipod Gammarus pulex), very widespread in the hyporheic zone of all sampled rivers, is of limited use because of their high tolerance to the analysed stressors. We recommend the use of certain taxonomic groups (comprising both meiofauna and macroinvertebrates) dwelling in the chalk hyporheic zone as indicators of eutrophication and colmation and, along with routine benthic sampling protocols, for a more comprehensive water and habitat quality assessment of chalk rivers.

  19. Analytical simulation of boiling water reactor pressure suppression pool swell

    International Nuclear Information System (INIS)

    In a loss-of-coolant accident, the pressure suppression pool of a boiling water reactor swells as a steam/air mixture is expelled from the drywell into the pool and large gas bubbles are formed beneath the surface. Many tests have been performed to quantify pool swell loads, but analytical methods have been limited in their ability to provide accurate loading estimates. With advancement of numerical methods, it is now feasible to numerically simulate the pool swell process. A finite difference solution algorithm is used to solve the transient imcompressible equations for the liquid flow field. Boundary conditions at the fluid-gas interface are determined using a simplified gas flow model. The program is used to simulate several pool swell tests: comparison of the simulation with test data shows good agreement

  20. Analytical simulation of boiling water reactor pressure suppression pool swell

    Energy Technology Data Exchange (ETDEWEB)

    Widener, S.K.

    1986-01-01

    In a loss-of-coolant accident, the pressure suppression pool of a boiling water reactor swells as a steam/air mixture is expelled from the drywell into the pool and large gas bubbles are formed beneath the surface. Many tests have been performed to quantify pool swell loads, but analytical methods have been limited in their ability to provide accurate loading estimates. With advancement of numerical methods, it is now feasible to numerically simulate the pool swell process. A finite difference solution algorithm is used to solve the transient imcompressible equations for the liquid flow field. Boundary conditions at the fluid-gas interface are determined using a simplified gas flow model. The program is used to simulate several pool swell tests: comparison of the simulation with test data shows good agreement.

  1. Microflora of nuclear research reactor pool water

    International Nuclear Information System (INIS)

    The circulation of pool water through the nuclear reactor core produces a bactericidal effect on the microflora due to the influence of various kinds of radiation. The microbe contents return to their initial level in 2 to 4 months after the circulation has stopped. The microflora comprises mainly cocci in large numbers, G-positive rods and fungi, and lower amounts of G-negative rods as compared with the water with which the reactor pool was initially filled. Increased amounts are present of radiation-resistant forms exhibiting intense production of catalase and nuclease. Supposedly, the presence of these enzymes is in some way beneficial to the microbes in their survival in the high-radiation zones. (author). 1 fig., 2 tabs., 12 refs

  2. Drivers of abundance and community composition of benthic macroinvertebrates in Ottawa River sediment near Chalk River Laboratories

    Energy Technology Data Exchange (ETDEWEB)

    Bond, M.J.; Rowan, D.; Silke, R.; Carr, J., E-mail: bondm@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-12-15

    The Ottawa River has received effluent from Chalk River Laboratories (CRL) for more than 60 years. Some radionuclides and contaminants released in effluents are bound rapidly to particles and deposited in bottom sediments where they may be biologically available to benthic invertebrates and other aquatic biota. As part of a larger ecological assessment, we assess the potential impact of contaminated sediments in the vicinity of CRL on local benthic community structure. Using bivariate and multivariate approaches, we demonstrate that CRL operations have had little impact on the local benthic community. Despite elevated anthropogenic radionuclide activity concentrations in sediment near CRL's process outfall, the benthic community is no less abundant or diverse than what is observed upstream at background levels. The Ottawa River benthic invertebrate community is structured predominantly by natural physical and biological conditions in the sediment, specifically sediment water content and organic content. These natural habitat conditions have a stronger influence on macroinvertebrate communities than sediment contamination. (author)

  3. Using environmental tracers to assess the extent of river-groundwater interaction in a quarried area of the English Chalk

    Energy Technology Data Exchange (ETDEWEB)

    Darling, W.G., E-mail: wgd@bgs.ac.uk [British Geological Survey, Maclean Building, Wallingford OX10 8BB (United Kingdom); Gooddy, D.C. [British Geological Survey, Maclean Building, Wallingford OX10 8BB (United Kingdom); Riches, J. [Thames Water Utilities Limited, Rose Kiln Court, Rose Kiln Lane, Reading RG2 0BY (United Kingdom); Wallis, I. [British Geological Survey, Maclean Building, Wallingford OX10 8BB (United Kingdom)

    2010-07-15

    The Swanscombe area of Kent, SE England represents a typical example of a heavily quarried Chalk area currently undergoing re-development. Because the Chalk is also an important aquifer, a good understanding of groundwater movement is required if environmental impacts are to be minimised and the water resource maximised. In particular, the nature of the relationship between the River Darent and groundwater in the Swanscombe Chalk Block requires better characterisation. Here, 'environmental tracers' in the form of ambient concentrations of stable isotopes, chlorofluorocarbons (CFCs), sulphur hexafluoride (SF{sub 6}) and tritium ({sup 3}H) are used to investigate this and other aspects of groundwater movement in the vicinity of the quarries. Stable isotopic contrasts indicate little evidence for widespread river infiltration to the regional Chalk aquifer, although stable isotope and {sup 3}H data suggest that 20-35% of the abstraction by river-valley public water supply boreholes may be derived from the river. The CFCs, while present at above-modern concentrations in almost all groundwaters, can be used as tracers, indicating basically S-N flowpaths in the area south of the quarries, though sub-karstic conduits associated with areas of Palaeogene cover add a level of uncertainty at the local scale. Simple piston flow residence times based on SF{sub 6} range from 1 to 17 a, but the data are probably better interpreted in terms of mixing between varying amounts of modern recharge derived from the south and deeper stored groundwater. The information gained from environmental tracers can therefore contribute to effective resource management.

  4. Seasonal nutrient dynamics in a chalk stream: the River Frome, Dorset, UK.

    Science.gov (United States)

    Bowes, M J; Leach, D V; House, W A

    2005-01-01

    Chalk streams provide unique, environmentally important habitats, but are particularly susceptible to human activities, such as water abstraction, fish farming and intensive agricultural activity on their fertile flood-meadows, resulting in increased nutrient concentrations. Weekly phosphorus, nitrate, dissolved silicon, chloride and flow measurements were made at nine sites along a 32 km stretch of the River Frome and its tributaries, over a 15 month period. The stretch was divided into two sections (termed the middle and lower reach) and mass balances were calculated for each determinand by totalling the inputs from upstream, tributaries, sewage treatment works and an estimate of groundwater input, and subtracting this from the load exported from each reach. Phosphorus and nitrate were retained within the river channel during the summer months, due to bioaccumulation into river biota and adsorption of phosphorus to bed sediments. During the autumn to spring periods, there was a net export, attributed to increased diffuse inputs from the catchment during storms, decomposition of channel biomass and remobilisation of phosphorus from the bed sediment. This seasonality of retention and remobilisation was higher in the lower reach than the middle reach, which was attributed to downstream changes in land use and fine sediment availability. Silicon showed much less seasonality, but did have periods of rapid retention in spring, due to diatom uptake within the river channel, and a subsequent release from the bed sediments during storm events. Chloride did not produce a seasonal pattern, indicating that the observed phosphorus and nitrate seasonality was a product of annual variation in diffuse inputs and internal riverine processes, rather than an artefact of sampling, flow gauging and analytical errors.

  5. Simulating Heterogeneous Infiltration and Contaminant leaching Processes at Chalk River, Ontario

    Science.gov (United States)

    Ali, M. A.; Ireson, A. M.; Keim, D.

    2015-12-01

    A study is conducted at a waste management area in Chalk River, Ontario to characterize flow and contaminant transport with the aim of contributing to improved hydrogeological risk assessment in the context of waste management. Field monitoring has been performed to gain insights into the unsaturated zone characteristics, moisture dynamics, and contaminant transport rates. The objective is to provide quantitative estimates of surface fluxes (quantification of infiltration and evaporation) and investigations of unsaturated zone processes controlling water infiltration and spatial variability in head distributions and flow rates. One particular issue is to examine the effectiveness of the clayey soil cap installed to prevent infiltration of water into the waste repository and the top sand soil cover above the clayey layer to divert the infiltrated water laterally. The spatial variability in the unsaturated zone properties and associated effects on water flow and contaminant transport observed at the site, have led to a concerted effort to develop improved model of flow and transport based on stochastic concepts. Results obtained through the unsaturated zone model investigations are combined with the hydrogeological and geochemical components and develop predictive tools to assess the long term fate of the contaminants at the waste management site.

  6. Post-irradiation examination of the 37M fuel bundle at Chalk River Laboratories (AECL)

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Daniels, T. [Ontario Power Generation, Pickering, Ontario (Canada); Montin, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2014-03-15

    The modified (-element (37M) fuel bundle was designed by Ontario Power Generation (OPG) to improve Critical Heat Flux (CHF) performance in ageing pressure tubes. A modification of the conventional 37-element fuel bundle design, the 37M fuel bundle allows more coolant flow through the interior sub-channels by way of a smaller central element. A demonstration irradiation (DI) of thirty-two fuel bundles was completed in 2011 at OPG's Darlington Nuclear Generating Station to confirm the suitability of the 37M fuel bundles for full core implementation. In support of the DI, fuel elements were examined in the Chalk River Laboratories Hot Cells. Inspection activities included: Bundle and element visual examination; Bundle and element dimensional measurements; Verification of bundle and element integrity; and Internal Gas Volume Measurements. The inspection results for 37M were comparable to that of conventional 37-element CANDU fuel. Fuel performance parameters of the 37M DI fuel bundle and fuel elements were within the range observed for similarly operated conventional 37-element CANDU fuel. Based on these Post Irradiation Examination (PIE) results, 37M fuel performed satisfactorily. (author)

  7. Suppression Pool Mixing and Condensation Tests in PUMA Facility

    International Nuclear Information System (INIS)

    Condensation of steam with non-condensable in the form of jet flow or bubbly flow inside the suppression pool is an important phenomenon on determining the containment pressure of a passively safe boiling water reactor. 32 cases of pool mixing and condensation test have been performed in Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility under the sponsor of the U.S. Nuclear Regulatory Commission to investigate thermal stratification and pool mixing inside the suppression pool during the reactor blowdown period. The test boundary conditions, such as the steam flow rate, the noncondensable gas flow rate, the initial water temperature, the pool initial pressure and the vent opening submergence depth, which covers a wide range of prototype (SBWR-600) conditions during Loss of Coolant Accident (LOCA) were obtained from the RELAP5 calculation. The test results show that steam is quickly condensed at the exit of the vent opening. For pure steam injection or low noncondensable injection cases, only the portion above the vent opening in the suppression pool is heated up by buoyant plumes. The water below the vent opening can be heated up slowly through conduction. The test results also show that the degree of thermal stratification in suppression pool is affected by the vent opening submergence depth, the pool initial pressure and the steam injection rate. And it is slightly affected by the initial water temperature. From these tests it is concluded that the pool mixing is strongly affected by the noncondensable gas flow rate. (authors)

  8. Residual radioactivity guidelines for the heavy water components test reactor at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Owen, M.B. Smith, R.; McNeil, J.

    1997-04-01

    Guidelines were developed for acceptable levels of residual radioactivity in the Heavy Water Components Test Reactor (HWCTR) facility at the conclusion of its decommissioning. Using source terms developed from data generated in a detailed characterization study, the RESRAD and RASRAD-BUILD computer codes were used to calculate derived concentration guideline levels (DCGLs) for the radionuclides that will remain in the facility. The calculated DCGLs, when compared to existing concentrations of radionuclides measured during a 1996 characterization program, indicate that no decontamination of concrete surfaces will be necessary. Also, based on the results of the calculations, activated concrete in the reactor biological shield does not have to be removed, and imbedded radioactive piping in the facility can remain in place. Viewed in another way, the results of the calculations showed that the present inventory of residual radioactivity in the facility (not including that associated with the reactor vessel and steam generators) would produce less than one millirem per year above background to a hypothetical individual on the property. The residual radioactivity is estimated to be approximately 0.04 percent of the total inventory in the facility as of March, 1997. According to the results, the only radionuclides that would produce greater than 0.0.1-millirem per year are Am-241 (0.013 mrem/yr at 300 years), C-14 (0.022 mrem/yr at 1000 years) and U-238 (0.034 mrem/yr at 6000 years). Human exposure would occur only through the groundwater pathways, that is, from water drawn from, a well on the property. The maximum exposure would be approximately one percent of the 4 millirem per year ground water exposure limit established by the U.S. Environmental Protection Agency. 11 refs., 13 figs., 15 tabs.

  9. TRIGA-III research reactor pool inspection using an underwater vehicle

    Energy Technology Data Exchange (ETDEWEB)

    Song, T. K.; Lee, J. R.; Kim, S. H.; Yoon, J. S.; Lee, B. J. [KAERI, Taejon (Korea, Republic of)

    1999-10-01

    For the inspection of radioactivity at the nuclear reactor and spent fuel storage pool, an underwater vehicle system has been developed. This underwater vehicle is navigated freely by five thruster which are controlled by developed control system and has a faculty of radiation detection at the inner wall and special point in pool using the radiation detector which is attached to the bottom of the vehicle. In this paper, the developed underwater vehicle and its components are described in detail. Also, the field test result in TRIGA-III research reactor pool is described.

  10. Convective cooling in a pool-type research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sipaun, Susan, E-mail: susan@nm.gov.my [Malaysian Nuclear Agency, Industrial Technology Division, Blok 29T, Bangi 43200, Selangor (Malaysia); Usman, Shoaib, E-mail: usmans@mst.edu [Missouri University of Science and Technology, Nuclear Engineering, 222 Fulton Hall 301 W.14th St., Rolla 64509 MO (United States)

    2016-01-22

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U{sub 3}Si{sub 2}Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm{sup −3}. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.

  11. Convective cooling in a pool-type research reactor

    Science.gov (United States)

    Sipaun, Susan; Usman, Shoaib

    2016-01-01

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

  12. Elk River Reactor dismantling

    International Nuclear Information System (INIS)

    The dismantling program was carried out in three overlapping phases: the planning phase which included the preliminary planning and selection of the dismantling approach, the dismantling phase which included all work performed to remove the reactor facility and restore the site to its pre-reactor condition, and the closeout phase which included the final site survey and efforts necessary to terminate the AEC license and contract. Of particular interest was the use of a remotely operated plasma cutting torch to section the pressure vessel internals, the pressure vessel and the outer thermal shield, the use of explosives in removal of the biological shield and the method of establishment of the criteria for material disposal

  13. Report on fuel pool water loss tests

    Energy Technology Data Exchange (ETDEWEB)

    Zalenski, R.F. [West Valley Nuclear Services Co., West Valley, NY (United States)

    1995-12-31

    To resolve potential concerns on the integrity of the fuel storage pool at the West Valley Demonstration Project (WVDP), a highly accurate testing technique was developed to quantify water losses from the pool. The fuel pool is an unlined, single wall, reinforced concrete structure containing approximately 818,000 gallons of water. Since an initial test indicated that water losses could possibly be attributed solely to evaporation, a cover was suspended and sealed over the pool to block evaporation losses. High accuracy water level and temperature instrumentation was procured and installed. The conclusions of this report indicate that unaccounted-for water losses from the pool are insignificant and there is no detectable leakage within the range of test accuracy.

  14. Vernal Pool Study 2005 Wallkill River National Wildlife Refuge

    Data.gov (United States)

    US Fish and Wildlife Service, Department of the Interior — These are data sheets from Wallkill River National Wildlife Refuge that will be part of a larger study to estimate the amphibian occupancy of vernal pool habitat at...

  15. Core neutronics of a swimming pool research reactor

    International Nuclear Information System (INIS)

    The initial cores of the 5 MW swimming pool research reactor of the Nuclear Research Centre, Tehran have been analyzed using the computer codes METHUSELAH and EQUIPOISE. The effective multiplication factor, critical mass, moderator temperature and void coefficients of the core have been calculated and compared with vendor's values. Calculated values agree reasonably well with the vendor's results. (author)

  16. Test reactor technology

    International Nuclear Information System (INIS)

    The Reactor Development Program created a need for engineering testing of fuels and materials. The Engineering Test Reactors were developed around the world in response to this demand. The design of the test reactors proved to be different from that of power reactors, carrying the fuel elements closer to the threshold of failure, requiring more responsive instrumentation, more rapid control element action, and inherent self-limiting behavior under accident conditions. The design of the experimental facilities to exploit these reactors evolved a new, specialized, branch of engineering, requiring a very high-lvel scientific and engineering team, established a meticulous concern with reliability, the provision for recovery from their own failures, and detailed attention to possible interactions with the test reactors. This paper presents this technology commencing with the Materials Testing Reactor (MTR) through the Fast Flux Test Facility, some of the unique experimental facilities developed to exploit them, but discusses only cursorily the experiments performed, since sample preparation and sample analyses were, and to some extent still are, either classified or proprietary. The Nuclear Engineering literature is filled with this information

  17. Test Pool Questions, Area III.

    Science.gov (United States)

    Sloan, Jamee Reid

    This manual contains multiple choice questions to be used in testing students on nurse training objectives. Each test includes several questions covering each concept. The concepts in section A, medical surgical nursing, are diseases of the following systems: musculoskeletal; central nervous; cardiovascular; gastrointestinal; urinary and male…

  18. Test Area North Pool Stabilization Project: Environmental assessment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-05-01

    The Test Area North (TAN) Pool is located within the fenced TAN facility boundaries on the Idaho National Engineering Laboratory (INEL). The TAN pool stores 344 canisters of core debris from the March, 1979, Three Mile Island (TMI) Unit 2 reactor accident; fuel assemblies from Loss-of-Fluid Tests (LOFT); and Government-owned commercial fuel rods and assemblies. The LOFT and government owned commercial fuel rods and assemblies are hereafter referred to collectively as {open_quotes}commercial fuels{close_quotes} except where distinction between the two is important to the analysis. DOE proposes to remove the canisters of TMI core debris and commercial fuels from the TAN Pool and transfer them to the Idaho Chemical Processing Plant (ICPP) for interim dry storage until an alternate storage location other than at the INEL, or a permanent federal spent nuclear fuel (SNF) repository is available. The TAN Pool would be drained and placed in an industrially and radiologically safe condition for refurbishment or eventual decommissioning. This environmental assessment (EA) identifies and evaluates environmental impacts associated with (1) constructing an Interim Storage System (ISS) at ICPP; (2) removing the TMI and commercial fuels from the pool and transporting them to ICPP for placement in an ISS, and (3) draining and stabilizing the TAN Pool. Miscellaneous hardware would be removed and decontaminated or disposed of in the INEL Radioactive Waste Management Complex (RWMC). This EA also describes the environmental consequences of the no action alternative.

  19. US team measurements during the June 1987 experimental HT release at the Chalk River Nuclear Laboratories, Ontario, Canada

    International Nuclear Information System (INIS)

    In June 1987, an experiment was performed at the Chalk River Nuclear Laboratories in Ontario, Canada, to study the oxidation of HT in the environment. The experiment involved a 30-minute release of 100 Ci of HT to the atmosphere at an elevation of one meter. The HTOHT ratios were shown to slowly increase downwind (/approximately/4 /times/ 10/sup /minus/5/ at 50 meters to almost 10/sup /minus/3 at 400 meters) as conversion of HT takes place. For several days after the release, HTO concentrations in the atmosphere remained elevated. Freeze-dried water from vegetation samples was found to be very low in HTO immediately after the release suggesting a very low direct uptake of HTO in air by vegetation. The tritiated water concentration increased during the first day, peaking during the second day (about 400 to 600 pCiml of water at 50 meters from the source) and decreasing by the end of the second day. The organically bound tritium continued to accumulate during the period following exposure (about 10 pCigm dry weight at 50 meters after two days). 4 refs., 6 figs., 2 tabs

  20. Savannah River Site production reactor technical specifications. K Production Reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-01

    These technical specifications are explicit restrictions on the operation of the Savannah River Site K Production Reactor. They are designed to preserve the validity of the plant safety analysis by ensuring that the plant is operated within the required conditions bounded by the analysis, and with the operable equipment that is assumed to mitigate the consequences of an accident. Technical specifications preserve the primary success path relied upon to detect and respond to accidents. This report describes requirements on thermal-hydraulic limits; limiting conditions for operation and surveillance for the reactor, power distribution control, instrumentation, process water system, emergency cooling and emergency shutdown systems, confinement systems, plant systems, electrical systems, components handling, and special test exceptions; design features; and administrative controls.

  1. Conceptual design of a pool type molten salt breeder reactor

    International Nuclear Information System (INIS)

    The renewed interest in molten salt coolant technology is backed by the 50 years history of molten salt nuclear technology development, mainly in Oak Ridge National Laboratory (ORNL). In Indian context MSBR is found to be one of the options for sustainable nuclear energy generation, especially in the third stage of the nuclear programme. The system can be operated at high temperature which makes high efficiency power conversion and efficient hydrogen generation through thermo-chemical reactions possible. At present development is in progress in BARC on two molten salt reactor concepts, one is pool type and the other is loop type. Here the design of pool type concept with 850MWe power is described. The core is designed to operate in the fast spectrum region so the conversion of 233U breeding is possible from thorium. Preliminary thermal hydraulic analysis is carried out with LiF-ThF4-UF4 as the primary fuel and coolant. The blanket material is also a molten salt, LiF-ThF4. Reactor physics calculations are also carried out for the feasibility studies of the core design of the reactor. FLiNaK is used as the secondary coolant for the calculations. Both forced circulation and natural circulation options are evaluated. (author)

  2. Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Owen, M.B.

    1997-04-01

    This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future.

  3. Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1

    International Nuclear Information System (INIS)

    This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future

  4. Bronx River bed sediments phosphorus pool and phosphorus compound identification

    Science.gov (United States)

    Wang, J.; Pant, H. K.

    2008-12-01

    Phosphorus (P) transport in the Bronx River degraded water quality, decreased oxygen levels, and resulted in bioaccumulation in sediment potentially resulting in eutrophication, algal blooms and oxygen depletion under certain temperature and pH conditions. The anthropogenic P sources are storm water runoff, raw sewage discharge, fertilizer application in lawn, golf course and New York Botanical Garden; manure from the Bronx zoo; combined sewoverflows (CSO's) from parkway and Hunts Point sewage plant; pollutants from East River. This research was conducted in the urban river system in New York City area, in order to control P source, figure out P transport temporal and spatial variations and the impact on water quality; aimed to regulate P application, sharing data with Bronx River Alliance, EPA, DEP and DEC. The sediment characteristics influence the distribution and bioavailbility of P in the Bronx River. The P sequential extraction gave the quantitative analysis of the P pool, quantifying the inorganic and organic P from the sediments. There were different P pool patterns at the 15 sites, and the substantial amount of inorganic P pool indicated that a large amount P is bioavailable. The 31P- NMR (Nuclear Magnetic Resonance Spectroscopy) technology had been used to identify P species in the 15 sites of the Bronx River, which gave a qualitative analysis on phosphorus transport in the river. The P compounds in the Bronx River bed sediments are mostly glycerophophate (GlyP), nucleoside monophosphates (NMP), polynucleotides (PolyN), and few sites showed the small amount of glucose-6-phosphate (G6P), glycerophosphoethanoamine (GPEA), phosphoenopyruvates (PEP), and inosine monophosphate (IMP). The land use spatial and temporal variations influence local water P levels, P distributions, and P compositions.

  5. UMRS LTRMP 2010/11 LCU Mapping -- Illinois River Starved Rock Pool

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — Aerial photographs for Pools 1-13 Upper Mississippi River System and Pools, Alton-Marseilles, Illinois River were collected in color infrared (CIR) in August of...

  6. UMRS LTRMP 2010/11 LCU Mapping -- Illinois River LaGrange Pool

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — Aerial photographs for Pools 1-13 Upper Mississippi River System and Pools, Alton-Marseilles, Illinois River were collected in color infrared (CIR) in August of...

  7. UMRS LTRMP 2010/11 LCU Mapping -- Illinois River Marseillies Pool

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — Aerial photographs for Pools 1-13 Upper Mississippi River System and Pools, Alton-Marseilles, Illinois River were collected in color infrared (CIR) in August of...

  8. UMRS LTRMP 2010/11 LCU Mapping -- Illinois River Peoria Pool

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — Aerial photographs for Pools 1-13 Upper Mississippi River System and Pools, Alton-Marseilles, Illinois River were collected in color infrared (CIR) in August of...

  9. UMRS LTRMP 2010/11 LCU Mapping -- Illinois River Alton Pool

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — Aerial photographs for Pools 1-13 Upper Mississippi River System and Pools, Alton-Marseilles, Illinois River were collected in color infrared (CIR) in August of...

  10. Design of neutron radiography facility in pool for the reactor RA-10

    International Nuclear Information System (INIS)

    RA-10 project consists in the design and construction of a multipurpose reactor for multiple applications, including radioisotopes production, material testing and an in pool facility for neutron imaging. Neutron imaging is a powerful tool for studies of materials and offer several advantages among other attenuation-based techniques. In this study mechanical and neutronic requirements for the RA-10 in pool neutron imaging facility are described. The MCNP neutronic model and the mechanical design satisfying these requirements in a first engineering stage are described. (author)

  11. Natural and mixed convection in the cylindrical pool of TRIGA reactor

    Science.gov (United States)

    Henry, R.; Tiselj, I.; Matkovič, M.

    2016-05-01

    Temperature fields within the pool of the JSI TRIGA MARK II nuclear research reactor were measured to collect data for validation of the thermal hydraulics computational model of the reactor tank. In this context temperature of the coolant was measured simultaneously at sixty different positions within the pool during steady state operation and two transients. The obtained data revealed local peculiarities of the cooling water dynamics inside the pool and were used to estimate the coolant bulk velocity above the reactor core. Mixed natural and forced convection in the pool were simulated with a Computational Fluid Dynamics code. A relatively simple CFD model based on Unsteady RANS turbulence model was found to be sufficient for accurate prediction of the temperature fields in the pool during the reactor operation. Our results show that the simple geometry of the TRIGA pool reactor makes it a suitable candidate for a simple natural circulation benchmark in cylindrical geometry.

  12. Reactor Simulator Testing

    Science.gov (United States)

    Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise J.

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  13. Draft environmental assessment -- Test Area North pool stabilization project update

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-06-01

    The purpose of this Environmental Assessment (EA) is to update the ``Test Area North Pool Stabilization Project`` EA (DOE/EA-1050) and finding of no significant impact (FONSI) issued May 6, 1996. This update analyzes the environmental and health impacts of a drying process for the Three Mile Island (TMI) nuclear reactor core debris canisters now stored underwater in a facility on the Idaho National Engineering and Environmental Laboratory (INEEL). A drying process was analyzed in the predecision versions of the EA released in 1995 but that particular process was determined to be ineffective and dropped form the Ea/FONSI issued May 6, 1996. The origin and nature of the TMI core debris and the proposed drying process are described and analyzed in detail in this EA. As did the 1996 EA, this update analyzes the environmental and health impacts of removing various radioactive materials from underwater storage, dewatering these materials, constructing a new interim dry storage facility, and transporting and placing the materials into the new facility. Also, as did the 1996 EA, this EA analyzes the removal, treatment and disposal of water from the pool, and placement of the facility into a safe, standby condition. The entire action would take place within the boundaries of the INEEL. The materials are currently stored underwater in the Test Area North (TAN) building 607 pool, the new interim dry storage facility would be constructed at the Idaho Chemical Processing Plant (ICPP) which is about 25 miles south of TAN.

  14. Environmental Assessment -- Test Area North pool stabilization project update

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-08-01

    The purpose of this Environmental Assessment (EA) is to update the ``Test Area North Pool Stabilization Project`` EA (DOE/EA-1050) and finding of no significant impact (FONSI) issued May 6, 1996. This update analyzes the environmental and health impacts of a drying process for the Three Mile Island (TMI) nuclear reactor core debris canisters now stored underwater in a facility on the Idaho National Engineering and Environmental Laboratory (INEEL). A drying process was analyzed in the predecision versions of the EA released in 1995 but that particular process was determined to be ineffective and dropped from the EA/FONSI issued May 6, 1996. A new drying process was subsequently developed and is analyzed in Section 2.1.2 of this document. As did the 1996 EA, this update analyzes the environmental and health impacts of removing various radioactive materials from underwater storage, dewatering these materials, constructing a new interim dry storage facility, and transporting and placing the materials into the new facility. Also, as did the 1996 EA, this EA analyzes the removal, treatment and disposal of water from the pool, and placement of the facility into a safe, standby condition. The entire action would take place within the boundaries of the INEEL. The materials are currently stored underwater in the Test Area North (TAN) building 607 pool, the new interim dry storage facility would be constructed at the Idaho Chemical Processing Plant (ICPP) which is about 25 miles south of TAN.

  15. Reactor Simulator Testing Overview

    Science.gov (United States)

    Schoenfeld, Michael P.

    2013-01-01

    Test Objectives Summary: a) Verify operation of the core simulator, the instrumentation & control system, and the ground support gas and vacuum test equipment. b) Examine cooling & heat regeneration performance of the cold trap purification. c) Test the ALIP pump at voltages beyond 120V to see if the targeted mass flow rate of 1.75 kg/s can be obtained in the RxSim. Testing Highlights: a) Gas and vacuum ground support test equipment performed effectively for operations (NaK fill, loop pressurization, and NaK drain). b) Instrumentation & Control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings and ramped within prescribed constraints. It effectively interacted with reactor simulator control model and defaulted back to temperature control mode if the transient fluctuations didn't dampen. c) Cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the minimum temperature indicating the design provided some heat regeneration. d) ALIP produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  16. Equipment and methods for examinations of fuel rods in the MIR reactor storage pool

    International Nuclear Information System (INIS)

    A wide range of tests of fuel rods and structural materials of water-cooled power reactors is performed in the loop facilities (LFs) of the MIR reactor. Depending on the objectives and tasks of different experiments, the performance of periodical interim examinations of irradiated items is required. However, as a result of some circumstances, it is not always possible to conduct them in hot cells. In this context, JSC 'SSC RIAR' has developed the equipment for interim examinations of fuel rods and design components of the experimental fuel assemblies (EFAs) in the MIR reactor storage pool (SP). Besides, this equipment can be used for cleaning of the examined items from surface deposits prior to the measurements. The paper describes the main characteristics and capabilities of the developed equipment, methodical aspects of the performed interim examinations, as well as some experimental results obtained using this equipment. In future, its upgrade is planned. (author)

  17. Remote maintenance considerations for swimming pool tokamak reactor

    International Nuclear Information System (INIS)

    Swimming Pool Tokamak Reactor (SPTR) is one of the candidate devices which are expected to demonstrate physical and engineering feasibility for fusion power reactors. In SPTR, water shield is adopted instead of solid shield structures. Among the advantages of SPTR are, from viewpoint of remote maintenance, small handling weight and high space availability between TF coils and a vacuum vessel. On the other hand, high dose rate during reactor repair and adverse effects on remote maintenance equipment by the shielding water might be the disadvantage of SPTR, where it is assumed that the shielding water is drained during reactor repair. Since the design of SPTR is still at the preliminary stage, for remote maintenance, much effort has been directed to clarification of design conditions such as environment and handling weight. As for the remote maintenance system concepts, studies have been focussed on those for a vacuum vessel and its internal structure (blanket, divertor and protection walls) expected to be repaired more frequently. The vacuum vessel assembly is divided into 21 sectors and number of TF coils is 14. A pair of TF coils are connected with each other by antitorque beams on the whole side surface. Vacuum vessel cassettes and associated blanket, divertor and protection walls are replaced through seven windows between TF coils pairs. Therefore each vacuum vessel cassette is required moving mechanisms in toroidal and radial directions. Options for slide mechanisms are wheels, balls, rollers and water bearings. Options for driving the cassette are self-driving by hydraulic motors and external driving by rack-pinion, wires or specific vehicles. As a result of studies, the moving mechanism with wheels and hydraulic motors has been selected for the reference design, and the system with water bearings and rack-pinion as an alternative. Furthermore typical concepts have been obtained for remote maintenance equipment such as wall-mounted manipulators, tools for

  18. Evaluation of pool swell velocity during large break loss of coolant accident in boiling water reactor Mark III containment design

    Energy Technology Data Exchange (ETDEWEB)

    Yan Jin, E-mail: jinyan10@gmail.co [GE-Hitachi Nuclear Energy, 3901 Castle Hayne Road, Wilmington, M/L-30, NC 28402 (United States); Bolger, Francis [GE-Hitachi Nuclear Energy, 3901 Castle Hayne Road, Wilmington, M/L-30, NC 28402 (United States)

    2010-07-15

    In boiling water reactor (BWR) design, safety scenarios such as main steam line break need to be evaluated. After the main steam line break, the steam will fill the upper dry well of the containment. It will then enter the vertical vent and eventually flow into the suppression pool via horizontal vents. The steam will create large bubbles in the suppression pool and cause the pool to swell. The impact of the pool swell on the equipment inside the pool and containment structure needed to be evaluated for licensing. GE has conducted a series of one-third scale three-vent air tests in supporting the horizontal vent pressure suppression system used in Mark III containment design for General Electric BWR plants. During the test, the air-water interface locations were tracked by conductivity probes. The pressure was measured at many locations inside the test rig as well. The purpose of the test was to provide a basis for the pool swell load definition for the Mark III containment. In this paper, a transient three-dimensional CFD model to simulate the one-third scale Mark III suppression pool swell process is illustrated. The Volume of Fluid (VOF) multiphase model is used to explicitly track the interface between the water liquid and the air. The CFD results such as flow velocity, pressure, interface locations are compared to the data from the test. Through comparisons, a technical approach to numerically model the pool swell phenomenon is established and benchmarked.

  19. Evaluation of pool swell velocity during large break loss of coolant accident in boiling water reactor Mark III containment design

    International Nuclear Information System (INIS)

    In boiling water reactor (BWR) design, safety scenarios such as main steam line break need to be evaluated. After the main steam line break, the steam will fill the upper dry well of the containment. It will then enter the vertical vent and eventually flow into the suppression pool via horizontal vents. The steam will create large bubbles in the suppression pool and cause the pool to swell. The impact of the pool swell on the equipment inside the pool and containment structure needed to be evaluated for licensing. GE has conducted a series of one-third scale three-vent air tests in supporting the horizontal vent pressure suppression system used in Mark III containment design for General Electric BWR plants. During the test, the air-water interface locations were tracked by conductivity probes. The pressure was measured at many locations inside the test rig as well. The purpose of the test was to provide a basis for the pool swell load definition for the Mark III containment. In this paper, a transient three-dimensional CFD model to simulate the one-third scale Mark III suppression pool swell process is illustrated. The Volume of Fluid (VOF) multiphase model is used to explicitly track the interface between the water liquid and the air. The CFD results such as flow velocity, pressure, interface locations are compared to the data from the test. Through comparisons, a technical approach to numerically model the pool swell phenomenon is established and benchmarked.

  20. Reactor physics calculations and their experimental validation for conversion and upgrading of a typical swimming pool type research reactor

    International Nuclear Information System (INIS)

    Detailed neutronic analysis of a typical swimming pool type research reactor, Pakistan Research Reactor-1 (PARR-1), was carried out for conversion of its core from 93% highly enriched uranium to 20% low enriched uranium fuel with power upgrading from 5 to 10 MW. Standard computer codes WIMS-D/4 and CITATION were employed to calculate core excess reactivity, power defect, reactivity effect of xenon and samarium, reactivity worth of fuel element, worth of control rods, shutdown margin, reactivity feedback coefficients, neutron flux and power peaking factors. A series of low and high power tests were performed on the newly converted core to determine its performance. A comparison between the calculated and measured results is presented in this article. The agreement is generally good

  1. Temperature coefficient of reactivity of a typical swimming pool type research reactor using low enriched uranium fuel

    International Nuclear Information System (INIS)

    The temperature coefficients of reactivity of a swimming pool type material test research reactor have been calculated using standard computer codes. It is observed that the core reactivity loss due to increase in water temperature and void formation is sensitive to control rod position at criticality. The reactivity decreases more rapidly when the core volume is small. (author)

  2. A novel representation of chalk hydrology in a land surface model

    Science.gov (United States)

    Rahman, Mostaquimur; Rosolem, Rafael

    2016-04-01

    Unconfined chalk aquifers contain a significant portion of water in the United Kingdom. In order to optimize the assessment and management practices of water resources in the region, modelling and monitoring of soil moisture in the unsaturated zone of the chalk aquifers are of utmost importance. However, efficient simulation of soil moisture in such aquifers is difficult mainly due to the fractured nature of chalk, which creates high-velocity preferential flow paths in the unsaturated zone. In this study, the Joint UK Land Environment Simulator (JULES) is applied on a study area encompassing the Kennet catchment in Southern England. The fluxes and states of the coupled water and energy cycles are simulated for 10 consecutive years (2001-2010). We hypothesize that explicit representation for the soil-chalk layers and the inclusion of preferential flow in the fractured chalk aquifers improves the reproduction of the hydrological processes in JULES. In order to test this hypothesis, we propose a new parametrization for preferential flow in JULES. This parametrization explicitly describes the flow of water in soil matrices and preferential flow paths using a simplified approach which can be beneficial for large-scale hydrometeorological applications. We also define the overlaying soil properties obtained from the Harmonized World Soil Database (HWSD) in the model. Our simulation results are compared across spatial scales with measured soil moisture and river discharge, indicating the importance of accounting for the physical properties of the medium while simulating hydrological processes in the chalk aquifers.

  3. Simulator for materials testing reactors

    International Nuclear Information System (INIS)

    A real-time simulator for both reactor and irradiation facilities of a materials testing reactor, “Simulator of Materials Testing Reactors”, was developed for understanding reactor behavior and operational training in order to utilize it for nuclear human resource development and to promote partnership with developing countries which have a plan to introduce nuclear power plant. The simulator is designed based on the JMTR (Japan Materials Testing Reactor), and it simulates operation, irradiation tests and various kinds of anticipated operational transients and accident conditions caused by the reactor and irradiation facilities. The development of the simulator was sponsored by the Japanese government as one of the specialized projects of advanced research infrastructure in order to promote basic as well as applied researches. This report summarizes the simulation components, hardware specification and operation procedure of the simulator. (author)

  4. E-SCAPE: A scale facility for liquid-metal, pool-type reactor thermal hydraulic investigations

    Energy Technology Data Exchange (ETDEWEB)

    Van Tichelen, Katrien, E-mail: kvtichel@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Mirelli, Fabio, E-mail: fmirelli@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Greco, Matteo, E-mail: mgreco@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Viviani, Giorgia, E-mail: giorgiaviviani@gmail.com [University of Pisa, Lungarno Pacinotti 43, 56126 Pisa (Italy)

    2015-08-15

    Highlights: • The E-SCAPE facility is a thermal hydraulic scale model of the MYRRHA fast reactor. • The focus is on mixing and stratification in liquid-metal pool-type reactors. • Forced convection, natural convection and the transition are investigated. • Extensive instrumentation allows validation of computational models. • System thermal hydraulic and CFD models have been used for facility design. - Abstract: MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a flexible fast-spectrum research reactor under design at SCK·CEN. MYRRHA is a pool-type reactor with lead bismuth eutectic (LBE) as primary coolant. The proper understanding of the thermal hydraulic phenomena occurring in the reactor pool is an important issue in the design and licensing of the MYRRHA system and liquid-metal cooled reactors by extension. Model experiments are necessary for understanding the physics, for validating experimental tools and to qualify the design for the licensing. The E-SCAPE (European SCAled Pool Experiment) facility at SCK·CEN is a thermal hydraulic 1/6-scale model of the MYRRHA reactor, with an electrical core simulator, cooled by LBE. It provides experimental feedback to the designers on the forced and natural circulation flow patterns. Moreover, it enables to validate the computational methods for their use with LBE. The paper will elaborate on the design of the E-SCAPE facility and its main parameters. Also the experimental matrix and the pre-test analysis using computational fluid dynamics (CFD) and system thermal hydraulics codes will be described.

  5. Decay heat removal in pool type fast reactor using passive systems

    Energy Technology Data Exchange (ETDEWEB)

    Parthasarathy, U. [Thermal Hydraulics Section, Nuclear and Safety Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Sundararajan, T. [Department of Mechanical Engineering, IIT-Madras, Chennai 600 036 (India); Balaji, C., E-mail: balaji@iitm.ac.in [Department of Mechanical Engineering, IIT-Madras, Chennai 600 036 (India); Velusamy, K.; Chellapandi, P.; Chetal, S.C. [Thermal Hydraulics Section, Nuclear and Safety Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Three dimensional thermal hydraulic analysis of decay heat system in a fast reactor model predictions compared with experimental results from PHENIX. Black-Right-Pointing-Pointer Calculations confirm adequacy of natural convection in decay heat removal. Black-Right-Pointing-Pointer Inter-wrapper flow found to reduce peak temperatures by 50 K in the blanket zone. - Abstract: Post shutdown decay heat removal in a fast reactor is one of the most important safety functions which must be accomplished with a very high reliability. To achieve high reliability, the fast breeder reactor design has emphasized on passive or near passive decay heat removal systems utilizing the natural convection in the heat removal path. A typical passive decay heat removal system used in recent designs of fast breeder reactors consists of a sodium to sodium heat exchanger and sodium to air heat exchanger which together remove heat directly from the hot pool to the final heat sink, which is air. Since these are safety systems, it is necessary to confirm the design with detailed numerical analysis. The numerical studies include pool hydraulics, natural convection phenomena in closed loops, flow through narrow gaps between SA, multi-scale modeling, etc. Toward understanding the evolution of thermal hydraulic parameters during natural convection decay heat removal, a three-dimensional CFD model for the primary system coupled with an appropriate one-dimensional model for the secondary system is proposed. The model has been validated against the results of natural convection test conducted in PHENIX reactor. Adopting the model for the Indian PFBR, six different decay heat removal cases have been studied which bring out the effect of safety grade decay heat removal system (SGDHRS) capacity, secondary sodium inventory and inter-wrapper flow heat transfer on the subassembly outlet temperatures that are important for safety evaluation of the reactor. From the

  6. Decay heat removal in pool type fast reactor using passive systems

    International Nuclear Information System (INIS)

    Highlights: ► Three dimensional thermal hydraulic analysis of decay heat system in a fast reactor model predictions compared with experimental results from PHENIX. ► Calculations confirm adequacy of natural convection in decay heat removal. ► Inter-wrapper flow found to reduce peak temperatures by 50 K in the blanket zone. - Abstract: Post shutdown decay heat removal in a fast reactor is one of the most important safety functions which must be accomplished with a very high reliability. To achieve high reliability, the fast breeder reactor design has emphasized on passive or near passive decay heat removal systems utilizing the natural convection in the heat removal path. A typical passive decay heat removal system used in recent designs of fast breeder reactors consists of a sodium to sodium heat exchanger and sodium to air heat exchanger which together remove heat directly from the hot pool to the final heat sink, which is air. Since these are safety systems, it is necessary to confirm the design with detailed numerical analysis. The numerical studies include pool hydraulics, natural convection phenomena in closed loops, flow through narrow gaps between SA, multi-scale modeling, etc. Toward understanding the evolution of thermal hydraulic parameters during natural convection decay heat removal, a three-dimensional CFD model for the primary system coupled with an appropriate one-dimensional model for the secondary system is proposed. The model has been validated against the results of natural convection test conducted in PHENIX reactor. Adopting the model for the Indian PFBR, six different decay heat removal cases have been studied which bring out the effect of safety grade decay heat removal system (SGDHRS) capacity, secondary sodium inventory and inter-wrapper flow heat transfer on the subassembly outlet temperatures that are important for safety evaluation of the reactor. From the results, it is concluded that the delay in initiation of SGDHRS, replacement

  7. An analysis of postulated accident for 49-2 Swimming Pool Reactor

    International Nuclear Information System (INIS)

    The thermal hydrodynamic code RETRAN-02 is used for safety analysis of Swimming Pool Reactor. Accident of partial-loss of flow, loss of offsite electric power and unexpected reactivity insertion are analysed and discussed. These results will be helpful for operation safety of the reactor

  8. Quality control of pool water from IEA-R1 reactor

    International Nuclear Information System (INIS)

    This paper presents the results of the pool water monitoring program of the IEA-R1 reactor of IPEN/CNEN-SP in normal operation. The considered period was previous to the systems upgrade and modernization for the new reactor operation condition: a power of 5 MW and operation time of 100 hours weekly. (author)

  9. Reliability modeling of Clinch River breeder reactor electrical shutdown systems

    International Nuclear Information System (INIS)

    The initial simulation of the probabilistic properties of the Clinch River Breeder Reactor Plant (CRBRP) electrical shutdown systems is described. A model of the reliability (and availability) of the systems is presented utilizing Success State and continuous-time, discrete state Markov modeling techniques as significant elements of an overall reliability assessment process capable of demonstrating the achievement of program goals. This model is examined for its sensitivity to safe/unsafe failure rates, sybsystem redundant configurations, test and repair intervals, monitoring by reactor operators; and the control exercised over system reliability by design modifications and the selection of system operating characteristics. (U.S.)

  10. Criticality safety calculations of the Soreq research reactor storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Caner, M.; Hirshfeld, H.; Nagler, A.; Silverman, I.; Bettan, M. [Soreq Nuclear Research Center, Yavne 81800 (Israel); Levine, S.H. [Penn State University, University Park 16802 (United States)

    2001-07-01

    The IRR-l spent fuel is to be relocated in a storage pool. The present paper describes the actual facility and summarizes the Monte Carlo criticality safety calculations. The fuel elements are to be placed inside cadmium boxes to reduce their reactivity. The fuel element is 7.6 cm by 8.0 cm in the horizontal plane. The cadmium box is effectively 9.7 cm by 9.7 cm, providing significant water between the cadmium and the fuel element. The present calculations show that the spent fuel storage pool is criticality safe even for fresh fuel elements. (author)

  11. Technical outline of a high temperature pool reactor with inherent passive safety features

    International Nuclear Information System (INIS)

    Many reactor designers world wide have successfully established technologies for very small reactors (less than 10 MWTH), and technologies for large power reactors (greater than 1000 MWTH), but have not developed small reactors (between 10 MWTH and 1000 MWth) which are safe, economic, and capable of meeting user technical, economic, and safety requirements. This is largely because the very small reactor technologies and the power reactor technologies are not amiable to safe and economic upsizing/downsizing. This paper postulates that new technologies, or novel combinations of existing technologies are necessary to the design of safe and economic small reactors. The paper then suggest a set of requirements that must be satisfied by a small reactor design, and defines a pool reactor that utilizes lead coolant and TRISO fuel which has the potential for meeting these requirements. This reactor, named LEADIR-PS, (an acronym for LEAD-cooled Integral Reactor, Passively Safe) incorporates the inherent safety features of the Modular High Temperature Gas Cooled Reactor (MWGR), while avoiding the cost of reactor and steam generator pressure vessels, and the safety concerns regarding pressure vessel rupture. This paper includes the description of a standard 200MW thermal reactor module based on this concept, called LEADIR-PS 200. (author)

  12. Digital modeling of radioactive and chemical waste transport in the aquifer underlying the Snake River Plain at the National Reactor Testing Station, Idaho

    Science.gov (United States)

    Robertson, J.B.

    1974-01-01

    Industrial and low-level radioactive liquid wastes at the National Reactor Testing Station (NRTS) in Idaho have been disposed to the Snake River Plain aquifer since 1952. Monitoring studies have indicated that tritium and chloride have dispersed over a 15-square mile (39-square kilometer) area of the aquifer in low but detectable concentrations and have only migrated as far as 5 miles (8 kilometers) downgradient from discharge points. The movement of cationic waste solutes, particularly 90Sr and 137Cs, has been significantly retarded due to sorption phenomena, principally ion exchange. 137Cs has shown no detectable migration in the aquifer and 90Sr has migrated only about 1.5 miles (2 kilometers) from the Idaho Chemical Processing Plant (ICPP) discharge well, and is detectable over an area of only 1.5 square miles ( 4 square kilometers) of the aquifer. Digital modeling techniques have been applied successfully to the analysis of the complex waste-transport system by utilizing numerical solution of the coupled equations of groundwater motion and mass transport. The model includes the effects of convective transport, flow divergence, two-dimensional hydraulic dispersion, radioactive decay, and reversible linear sorption. The hydraulic phase of the model uses the iterative, alternating direction, implicit finite-difference scheme to solve the groundwater flow equations, while the waste-transport phase uses a modified method of characteristics to solve the solute transport equations simulated by the model. The modeling results indicate that hydraulic dispersion (especially transverse) is a much more significant influence than previously suggested by earlier studies. The model has been used to estimate future waste migration patterns for varied assumed hydrological and waste conditions up through the year 2000. The hydraulic effects of recharge from the Big Lost River have an important (but not predominant) influence on the simulated future migration patterns. For the

  13. Chalk as a reservoir

    DEFF Research Database (Denmark)

    Fabricius, Ida Lykke

    , or be partly or fully cemented where diagenesis is more pronounced. It is a chalk characteristic that permea bility is controlled by the porosity and internal surface of the mud matrix, whereas the larger pores play an insignificant role. Cemented microfossils may take up a significant volume in a wackestone......Reservoir properties of chalk depend on the primary sediment composition as well as on subsequent diagenesis and tectonic events. Chalks of the North Sea almost exclusively have mudstone or wackestone texture. Microfossils may have retained their porosity where degree of diagenesis is low...... reduces porosity at the same time as it increases specific surface and thus cause permeability to be low. In the Central North Sea the silica is quartzitic. Silica rich chalk intervals are typically found in the Ekofisk and Hod formations. In addition to silica, Upper Cretaceous and Palæogene chalks...

  14. Reduced scale simulations of boiling water reactor pool swell: some limitations to the scaling laws

    International Nuclear Information System (INIS)

    Several potential sources of misscaling in reduced scale experimental tests have been systematically investigated. Increases in the enthalpy in-flux during pool swell increase resultant uploads; slight boundary flexibility due to small air bubbles attached to the pool walls or true fluid structure interaction can increase peak pool boundary loads; the presence of water vapor in the wetwell airspace can either increase or decrease pool swell uploads, depending on the vapor fraction initially present. 14 refs

  15. Simulation of CANDU Fuel Behaviour into In-Reactor LOCA Tests

    International Nuclear Information System (INIS)

    The purpose of this work is to simulate the behaviour of an instrumented, unirradiated, zircaloy sheathed UO2 fuel element assembly of CANDU type, subjected to a coolant depressurization transient in the X-2 pressurized water loop of the NRX reactor at the Chalk River Nuclear Laboratories in 1983. The high-temperature transient conditions are such as those associated with the onset of a loss of coolant accident (LOCA). The data and the information related to the experiment are those included in the OECD/NEA-IFPE Database (IFPE/CANDU-FIO-131 NEA-1783/01). As tool for this simulation is used the TRANSURANUS fuel performance code, developed at ITU, Germany, along with the corresponding fabrication and in-reactor operating conditions specific of the CANDU PHWR fuel. The results, analyzed versus the experimental ones, are encouraging and perfectible. (author)

  16. Advanced test reactor. Testing capabilities and plans

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. For future research, some ATR modifications and enhancements are currently planned. In 2007 the US Department of Energy designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR for material testing research by a broader user community. This paper provides more details on some of the ATR capabilities, key design features, experiments, and plants for the NSUF. (author)

  17. Thermal stratification experiments with the condensation pool test rig

    Energy Technology Data Exchange (ETDEWEB)

    Laine, J.; Puustinen, M.

    2006-01-15

    This report summarizes the results of the thermal stratification experiments with the condensation pool test rig. One experiment was carried out in March and another one in May 2005 with a scaled down test facility designed and constructed at Lappeenranta University of Technology. The main purpose of the experiments was to study thermal stratification phenomenon in the condensation pool during steam discharge and to produce data for the validation of the stratification model of the APROS code. (au)

  18. Production and release of 14C from a swimming pool reactor

    International Nuclear Information System (INIS)

    The annual production rate of 14C in the Apsara swimming pool reactor works out to be about 2.94 mCi. The concentration distribution of 14C in different compartments viz. pool water, reactor hall air and ion-exchange resin ranged from 200 to 440 pCi/l, 0.09 to 0.38 pCi/l, an average concentration of 8.16 pCi/g respectively. The mean residence time of 14C in pool water is evaluated to be about 7 days taking into account various sinks. The study revealed atmospheric exchange at the air-water interface as the dominant process responsible for the loss of 14C from the pool water. (author). 7 refs., 2 figs., 4 tabs

  19. Development, Implementation and Experimental Validations of Activation Products Models for Water Pool Reactors

    International Nuclear Information System (INIS)

    Some parameters were obtained both calculations and experiments in order to determined the source of the meaning activation products in water pool reactors. In this case, the study was done in RA-6 reactor (Centro Atomico Bariloche - Argentina).In normal operation, neutron flux on core activates aluminium plates.The activity on coolant water came from its impurities activation and meanly from some quantity of aluminium that, once activated, leave the cladding and is transported by water cooling system.This quantity depends of the 'recoil range' of each activation reaction.The 'staying time' on pool (the time that nuclides are circulating on the reactor pool) is another characteristic parameter of the system.Stationary state activity of some nuclides depends of this time.Also, several theoretical models of activation on coolant water system are showed, and their experimental validations

  20. PITR: Princeton Ignition Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1978-12-01

    The principal objectives of the PITR - Princeton Ignition Test Reactor - are to demonstrate the attainment of thermonuclear ignition in deuterium-tritium, and to develop optimal start-up techniques for plasma heating and current induction, in order to determine the most favorable means of reducing the size and cost of tokamak power reactors. This report describes the status of the plasma and engineering design features of the PITR. The PITR geometry is chosen to provide the highest MHD-stable values of beta in a D-shaped plasma, as well as ease of access for remote handling and neutral-beam injection.

  1. Risk-based Prioritization of Facility Decommissioning and Environmental Restoration Projects in the National Nuclear Legacy Liabilities Program at the Chalk River Laboratory - 13564

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, Jerel G.; Kruzic, Michael [WorleyParsons, Mississauga, ON, L4W 4H2 (United States); Castillo, Carlos [WorleyParsons, Las Vegas, NV 89128 (United States); Pavey, Todd [WorleyParsons, Idaho Falls, ID 83402 (United States); Alexan, Tamer [WorleyParsons, Burnaby, BC, V5C 6S7 (United States); Bainbridge, Ian [Atomic Energy Canada Limited, Chalk River Laboratories, Chalk River, ON, K0J1J0 (Canada)

    2013-07-01

    Chalk River Laboratory (CRL), located in Ontario Canada, has a large number of remediation projects currently in the Nuclear Legacy Liabilities Program (NLLP), including hundreds of facility decommissioning projects and over one hundred environmental remediation projects, all to be executed over the next 70 years. Atomic Energy of Canada Limited (AECL) utilized WorleyParsons to prioritize the NLLP projects at the CRL through a risk-based prioritization and ranking process, using the WorleyParsons Sequencing Unit Prioritization and Estimating Risk Model (SUPERmodel). The prioritization project made use of the SUPERmodel which has been previously used for other large-scale site prioritization and sequencing of facilities at nuclear laboratories in the United States. The process included development and vetting of risk parameter matrices as well as confirmation/validation of project risks. Detailed sensitivity studies were also conducted to understand the impacts that risk parameter weighting and scoring had on prioritization. The repeatable prioritization process yielded an objective, risk-based and technically defendable process for prioritization that gained concurrence from all stakeholders, including Natural Resources Canada (NRCan) who is responsible for the oversight of the NLLP. (authors)

  2. Risk-based Prioritization of Facility Decommissioning and Environmental Restoration Projects in the National Nuclear Legacy Liabilities Program at the Chalk River Laboratory - 13564

    International Nuclear Information System (INIS)

    Chalk River Laboratory (CRL), located in Ontario Canada, has a large number of remediation projects currently in the Nuclear Legacy Liabilities Program (NLLP), including hundreds of facility decommissioning projects and over one hundred environmental remediation projects, all to be executed over the next 70 years. Atomic Energy of Canada Limited (AECL) utilized WorleyParsons to prioritize the NLLP projects at the CRL through a risk-based prioritization and ranking process, using the WorleyParsons Sequencing Unit Prioritization and Estimating Risk Model (SUPERmodel). The prioritization project made use of the SUPERmodel which has been previously used for other large-scale site prioritization and sequencing of facilities at nuclear laboratories in the United States. The process included development and vetting of risk parameter matrices as well as confirmation/validation of project risks. Detailed sensitivity studies were also conducted to understand the impacts that risk parameter weighting and scoring had on prioritization. The repeatable prioritization process yielded an objective, risk-based and technically defendable process for prioritization that gained concurrence from all stakeholders, including Natural Resources Canada (NRCan) who is responsible for the oversight of the NLLP. (authors)

  3. The effect of encroachments on structure impact loads during a pool swell transient based on small-scale testing

    International Nuclear Information System (INIS)

    Experiments were conducted to investigate suppression pool dynamics in boiling water reactor (BWR) containments which have large overhanging structures attached to the drywell wall. Several 1/10 linear scale air blowdown tests utilizing Froude scaling (balance of gravity and inertia forces) were performed in this tests series. The drywall pressure was measured and high speed movies were made of the pool response. The resultant pool response was a function of encroachment size. Small encroachments did not significantly alter the response obtained for he unobstructed pool. For the large radial and circumferential encroachment, however, the increased inertia of the extra water lifted by the rising bubble delayed the transient, resulting in much lower pool swell velocities. This led to a stable liquid surface at higher elevations, but the surface curvature coupled with the relatively low pool surface velocities significantly mitigates structure impact loadings

  4. Garigliano Nuclear Power Plant, Italy: Decontamination and Rearranging of Reactor Canal and Spent Fuel Pool

    International Nuclear Information System (INIS)

    Garigliano nuclear power plant was a 506 MW(th), first generation, dual cycle BWR. It started operation in 1964 and finally shut down in 1978, following the discovery of serious damage to a secondary steam generator. This section describes decontamination activities carried out in 1991–1993 in preparation for safe enclosure of Garigliano reactor building.1 Activities were carried out after completion of spent fuel transport off-site (1985–1987). A schematic of the spent fuel pool and adjacent areas is provided. Decontamination activities included the following: (a) Agitation and resuspension of pool sediments using water jets and water filtration. (b) Lowering of water level and parallel decontamination of pool walls with high pressure water jets of approximately 700 kg/cm2. (c) Removal, decontamination and interim storage on gangways of equipment located on the pool south-east wall. (d) Removal, decontamination and storage of the fuel transport container platform. (e) Removal of four fuel racks to their pool wall bearings, decontamination and transfer to the fresh fuel room. (f) Decontamination of the vessel head platform, removal from the reactor canal, brushing and coating to allow preservation and fixing of loose contamination. Eventually, this component was placed back in the reactor canal. (g) Construction in the reactor canal of an interim structure supporting fuel racks. At the completion of the work, this structure was dismounted, decontaminated and removed. (h) Removal of fuel racks (five at a time) to their pool wall bearings, decontamination and interim storage in the reactor canal. (i) Gradual lowering of the pool water level to some 50 cm from the pool floor and parallel decontamination of fixed structures and walls. (j) Discovery by visual inspection and radiological checks, of activated components on the floor of the pool. Retrieval of all this material, segmentation as needed, temporary storage in containers and later transfer to the high

  5. TRIGA reactor spent fuel pool under severe earthquake conditions

    International Nuclear Information System (INIS)

    Supplemental criticality safety analysis of a pool type storage for TRIGA spent fuel at 'Jozef Stefan' Institute in Ljubljana, Slovenia, is presented. Previous results (Ravnik, M, Glumac, B., 1996) have shown that subcriticality is not guaranteed for some postulated accidents. To mitigate this deficiency, a study was made about replacing a certain number of fuel elements in the rack with absorber rods (Glumac, B., Ravnik, M., Logar, M., 1997) to lower the supercriticality probability, when the pitch is decreased to contact (as a consequence of a severe earthquake) in a square arrangement. The criticality analysis for the hexagonal contact pitch is presented in this paper, following the same scenario as outlined above. The Monte Carlo computer code MCNP4B with ENDF-B/VI library and detailed three dimensional geometry was used. First, the analysis about the influence of the number of triangular fuel piles on the bottom that could appear, if the fuel rack, made of three segments, disintegrates, is presented. Next, the number of uniformly mixed absorber rods in the lattice needed to sustain the subcriticality of the storage for hexagonal contact pitch is studied. Because of supercriticality possibility due to random mixing of the absorber rods in the case of lattice compaction, a probabilistic study was made in order to sample the probability density functions for random lattice loadings of the absorber rods. The results show that reasonably low probabilities for supercriticality can be achieved even when fresh 12 wt.% standard TRIGA fuel is stored in the spent fuel pool. (orig.)

  6. Multi-dimensional pool analysis of Phenix end-of-life natural circulation test with MARS-LMR code

    International Nuclear Information System (INIS)

    Highlights: • The natural circulation test performed in Phenix reactor has been analyzed with MARS-LMR code. • A multi-dimensional approach for the hot pool and the cold pool has been adopted in the analysis. • A detailed comparison between the test data and the simulation results has been performed. - Abstract: The MARS-LMR code is a key system analysis tool for the development of a sodium-cooled fast reactor in Korea. The code has been successfully applied for the transient analysis of conceptual designs of SFR since 2007 mainly based on a one-dimensional approach. In recent studies, it was identified that one-dimensional modeling of a pool-type SFR has limitations on describing complicated thermal–hydraulic phenomena in pool regions at natural circulation conditions. In the present study, the natural circulation test performed in Phenix reactor by CEA has been analyzed with a multi-dimensional approach of MARS-LMR. Only the hot pool and the cold pool regions are modeled multi-dimensionally and other parts of the plant are described one-dimensionally in the analysis. Even though a very careful treatment of initial flow condition is required, this multi-dimensional modeling of pool regions results in quite accurate prediction of the temperature distributions measured at several points during the test when it is compared to the results with one-dimensional pool nodalization. It is suggested that a detailed modeling of pool regions is essential for the future analysis of pool-type SFRs. The multi-dimensional modeling capability can be enhanced through the improvement of the existing system code or by the combination of system code and CFD code

  7. Storage of water reactor spent fuel in water pools. Survey of world experience

    International Nuclear Information System (INIS)

    Following discharge from a nuclear reactor, spent fuel has to be stored in water pools at the reactor site to allow for radioactive decay and cooling. After this initial storage period, the future treatment of spent fuel depends on the fuel cycle concept chosen. Spent fuel can either be treated by chemical processing or conditioning for final disposal at the relevant fuel cycle facilities, or be held in interim storage - at the reactor site or at a central storage facility. Recent forecasts predict that, by the year 2000, more than 150,000 tonnes of heavy metal from spent LWR fuel will have been accumulated. Because of postponed commitments regarding spent fuel treatment, a significant amount of spent fuel will still be held in storage at that time. Although very positive experience with wet storage has been gained over the past 40 years, making wet storage a proven technology, it appears desirable to summarize all available data for the benefit of designers, storage pool operators, licensing agenices and the general public. Such data will be essential for assessing the viability of extended water pool storage of spent nuclear fuel. In 1979, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD jointly issued a questionnaire dealing with all aspects of water pool storage. This report summarizes the information received from storage pool operators

  8. Quantitative analysis of gamma ray emitting radionuclide in reactor pool water of HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myong Seop; Kim, Hee Gon; Ahn, Guk Hoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-11-15

    The species and concentrations of the radionuclide in the primary coolant of HANARO were analyzed by using the gamma ray spectroscopy. The full energy peak efficiency for the volume source was calibrated as function of the photon energy for an HRGs detector. The primary coolant of HANARO was picked at the primary coolant purification system, and the water at the upper part of the reactor pool was taken at about 20cm under the pool surface. In the primary coolant, the concentrations of Na-24, Mg-27 and Al-28 were much higher than those of other nuclide, and they were in 1{approx}6x10'6'Bq/liter. Their origins were radiative reactions of aluminium used as the structure material and cladding of the nuclear fuel. The concentrations of Xe-138 and Xe-133 were relatively higher than those of other fission fragments. The source of the fission fragments in the coolant was the surface contamination of the nuclear fuel by uranium. Ar-41, Ce-141, Na-24 and Xe-133 were detected in the water at the upper part of the reactor pool. Na-24 was the main source of the pool top radiation level, and Xe-133 and Ar-41 were the main gaseous radionuclide released through the reactor pool surface.

  9. Quantitative analysis of gamma-ray emitting radionuclide in reactor pool water of HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myong-Seop; Kim, Hee-Gon; Ahn, Guk-Hoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-11-15

    The species and concentrations of the radionuclide in the primary coolant of HANARO were analyzed by using the gamma-ray spectroscopy. The full-energy peak efficiency for the volume source was calibrated as a function of the photon energy for an HPGe detector. The primary coolant of HANARO was picked at the primary coolant purification system, and the water at the upper part of the reactor pool was taken at about 20 cm under the pool surface. In the primary coolant, the concentrations of Na-24, Mg-27 and Al-28 were much higher than those of other nuclide, and they were in 1-6 x 10{sup 6} Bq/liter. Their origins were radiative reactions of aluminium used as the structure material and cladding of the nuclear fuel. The concentrations of Xe-138 and Xe-133 were relatively higher than those of other fission fragments. The source of the fission fragments in the coolant was the surface contamination of the nuclear fuel by uranium. Ar-41, Ce-141, Na-24 and Xe-133 were detected in the water at the upper part of the reactor pool. Na-24 was the main source of the pool top radiation level, and Xe-133 and Ar-41 were the main gaseous radionuclide released through the reactor pool surface.

  10. 3-dimensional thermohydraulic analysis of KALIMER reactor pool during unprotected accidents

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Bum; Hahn Do Hee

    2003-01-01

    During a normal reactor scram, the heat generation is reduced almost instantaneously while the coolant flow rate follows the pump coastdown. This mismatch between power and flow results in a situation where the core flow entering the hot pool is at a lower temperature than the temperature of the bulk pool sodium. This temperature difference leads to thermal stratification. Thermal stratification can occur in the hot pool region if the entering coolant is colder than the existing hot pool coolant and the flow momentum is not large enough to overcome the negative buoyancy force. Since the fluid of hot pool enters IHXs, the temperature distribution of hot pool can alter the overall system response. Hence, it is necessary to predict the pool coolant temperature distribution with sufficient accuracy to determine the inlet temperature conditions for the IHXs and its contribution to the net buoyancy head. Therefore, two-dimensional hot pool thermohydraulic model named HP2D has been developed. In this report code-to-code comparison analysis between HP2D and COMMIX-1AR/P has been performed in the case of steady-state and UTOP.

  11. TRIGA reactor spent fuel pool under severe earthquake conditions

    Energy Technology Data Exchange (ETDEWEB)

    Logar, M. [Univ. of Maribor (Slovenia). Fac. of Elec. Eng.; Glumac, B.; Maucec, M. [`Jozef Stefan` Institute, Jamova 39, POB 100, 1111 Ljubljana (Slovenia)

    1998-07-01

    Supplemental criticality safety analysis of a pool type storage for TRIGA spent fuel at `Jozef Stefan` Institute in Ljubljana, Slovenia, is presented. Previous results (Ravnik, M, Glumac, B., 1996) have shown that subcriticality is not guaranteed for some postulated accidents. To mitigate this deficiency, a study was made about replacing a certain number of fuel elements in the rack with absorber rods (Glumac, B., Ravnik, M., Logar, M., 1997) to lower the supercriticality probability, when the pitch is decreased to contact (as a consequence of a severe earthquake) in a square arrangement. The criticality analysis for the hexagonal contact pitch is presented in this paper, following the same scenario as outlined above. The Monte Carlo computer code MCNP4B with ENDF-B/VI library and detailed three dimensional geometry was used. First, the analysis about the influence of the number of triangular fuel piles on the bottom that could appear, if the fuel rack, made of three segments, disintegrates, is presented. Next, the number of uniformly mixed absorber rods in the lattice needed to sustain the subcriticality of the storage for hexagonal contact pitch is studied. Because of supercriticality possibility due to random mixing of the absorber rods in the case of lattice compaction, a probabilistic study was made in order to sample the probability density functions for random lattice loadings of the absorber rods. The results show that reasonably low probabilities for supercriticality can be achieved even when fresh 12 wt.% standard TRIGA fuel is stored in the spent fuel pool. (orig.) 7 refs.

  12. Cost estimates of operating onsite spent fuel pools after final reactor shutdown

    International Nuclear Information System (INIS)

    This report presents estimates of the annual costs of operating spent fuel pools at nuclear power stations after the final shutdown of one or more onsite reactors. Its purpose is to provide basic spent fuel storage cost information for use in evaluating DOE's reference nuclear waste management system, as well as alternate systems. The basic model of an independent spent fuel storage installation (ISFSI) used in this study was based on General Electric Corporation's Morris Operation and was modified to reflect mean storage capabilities at an unspecified, or ''generic,'' US reactor site. Cost data for the model came from several sources, including both operating and shutdown nuclear power stations and existing ISFSIs. Duke Power Company has estimated ISFSI costs based on existing spent fuel storage costs at its nuclear power stations. Similarly, nuclear material handling facilities such as the Morris Operation, the West Valley Demonstration Project, and the retired Humbolt Bay nuclear power station have compiled spent fuel storage cost data based on years of operating experience. Consideration was given to the following factors that would cause operating costs to vary among pools: (1) The number of spent fuel pools at a given reactor site; (2) the number of operating and shutdown reactors onsite; (3) geographic location; and (4) pool storage capacity. 10 ref., 6 figs., 7 tabs

  13. CFD aided analysis of a scaled down model of the Brazilian Multipurpose Reactor (RMB) pool

    International Nuclear Information System (INIS)

    Research reactors are commonly built inside deep pools that provide radiological and thermal protection and easy access to its core. Reactors with thermal power in the order of MW usually use an auxiliary thermal-hydraulic circuit at the top of its pool to create a purified hot water layer (HWL). Thermal-hydraulic analysis of the flow configuration in the pool and HWL is paramount to insure radiological protection. A useful tool for these analyses is the application of CFD (Computational Fluid Dynamics). To obtain satisfactory results using CFD it is necessary the verification and validation of the CFD numerical model. Verification is divided in code and solution verifications. In the first one establishes the correctness of the CFD code implementation and in the former estimates the numerical accuracy of a particular calculation. Validation is performed through comparison of numerical and experimental results. This paper presents a dimensional analysis of the RMB (Brazilian Multipurpose Reactor) pool to determine a scaled down experimental installation able to aid in the HWL numerical investigation. Two CFD models were created one with the same dimensions and boundary conditions of the reactor prototype and the other with 1/10 proportion size and boundary conditions set to achieve the same inertial and buoyant forces proportions represented by Froude Number between the two models. Results comparing the HWL thickness show consistence between the prototype and the scaled down model behavior. (author)

  14. CFD aided analysis of a scaled down model of the Brazilian Multipurpose Reactor (RMB) pool

    Energy Technology Data Exchange (ETDEWEB)

    Schweizer, Fernando L.A.; Lima, Claubia P.B.; Costa, Antonella L.; Veloso, Maria A.F., E-mail: ando.schweizer@gmail.com, E-mail: claubia@nuclear.ufmg.br, E-mail: antonella@nuclear.ufmg.br, E-mail: mdora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (DEN/UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Santos, Andre A.C.; Costa, Antonio C.L., E-mail: aacs@cdtn.br, E-mail: aclc@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN/-MG), Belo Horizonte, MG (Brazil)

    2013-07-01

    Research reactors are commonly built inside deep pools that provide radiological and thermal protection and easy access to its core. Reactors with thermal power in the order of MW usually use an auxiliary thermal-hydraulic circuit at the top of its pool to create a purified hot water layer (HWL). Thermal-hydraulic analysis of the flow configuration in the pool and HWL is paramount to insure radiological protection. A useful tool for these analyses is the application of CFD (Computational Fluid Dynamics). To obtain satisfactory results using CFD it is necessary the verification and validation of the CFD numerical model. Verification is divided in code and solution verifications. In the first one establishes the correctness of the CFD code implementation and in the former estimates the numerical accuracy of a particular calculation. Validation is performed through comparison of numerical and experimental results. This paper presents a dimensional analysis of the RMB (Brazilian Multipurpose Reactor) pool to determine a scaled down experimental installation able to aid in the HWL numerical investigation. Two CFD models were created one with the same dimensions and boundary conditions of the reactor prototype and the other with 1/10 proportion size and boundary conditions set to achieve the same inertial and buoyant forces proportions represented by Froude Number between the two models. Results comparing the HWL thickness show consistence between the prototype and the scaled down model behavior. (author)

  15. Hyporheic Exchange in Gravel-Bed Rivers with Pool-Riffle Morphology: A 3D Model

    Science.gov (United States)

    Tonina, D.; Buffington, J. M.

    2004-12-01

    The hyporheic zone is a saturated band of sediment that surrounds river flow and forms a linkage between the river and the aquifer. It is a rich ecotone where benthic, hyporheic, and groundwater species temporarily or permanently reside. Head gradients along the streambed draw river water into the hyporheic zone and expel pore water into the stream. This process, known as hyporheic exchange, is important for delivering nutrients, oxygen and other solutes to the sediment, and for washing away waste products to support this ecotone. It is an essential component of the carbon and nitrogen cycles, and it controls in-stream contaminant transport. Although hyporheic exchange has been studied in sand-bed rivers with two-dimensional dune morphology, few studies have been conducted for gravel-bed rivers with three-dimensional pool-riffle geometry. The hyporheic zone of gravel-bed rivers is particularly important for salmonids, many of which are currently at risk world wide. Salmon and trout lay their eggs within the hyporheic zone for incubation. After hatching, the alevins live in the gravel before emerging into the stream. The upwelling and downwelling hyporheic fluxes are intense in these streams due to the highly permeable sediment and strong head variations forced by shallow flow over high-amplitude bed forms. Moreover, gravel-bed rivers show a wide range of flow regimes that change seasonally and have strong effects on hyporheic exchange. To study this exchange, we used four sets of pool-riffle geometries in twelve recirculating flume experiments. We kept a constant bed-form wavelength, but changed the bed-form amplitude and imposed three discharges, covering a wide range of hydraulic and geometric characteristics. Hyporheic exchange was predicted from a three-dimensional model based on bedform-induced pumping transport, where the boundary head profile is the pressure head distribution at the sediment interface, measured with an array of mini-piezometers buried within

  16. Combined effects experiments with the condensation pool test facility

    Energy Technology Data Exchange (ETDEWEB)

    Puustinen, M. [Lappeenranta Univ. of Technology, Nuclear Safety Research Unit (Finland)

    2007-01-15

    This report summarizes the results of the condensation pool experiments in spring 2006, where steam and steam/air mixture was blown into the pool through a DN200 blowdown pipe. Altogether three experiments, each consisting of several blows, were carried out with a scaled down test facility designed and constructed at Lappeenranta University of Technology. The main purpose of the experiments was to study the effects of non-condensable gas present in the discharge flow. Particularly pressure pulses inside the blowdown pipe and at the pool bottom caused by chugging were of interest. The test pool was an open stainless steel tank with a wall thickness of 4 mm and a bottom thickness of 5 mm containing 15 m3 of water. The nearby PACTEL test facility was used as a steam source. During the experiments the initial pressure of the steam source was 0.5 MPa and the pool water bulk temperature ranged from 40 C to 70 C. The test facility was equipped with high frequency instrumentation for capturing different aspects of the investigated phenomena. The data acquisition program recorded data with the frequency of 10 kHz. A digital high-speed video camera was used for visual observation of the pool interior. Air, in quantities even less than 1 %, reduced the condensation rate considerably. The high pressure pulses registered inside the blowdown pipe due to water hammer propagation during chugging almost disappeared when the combined discharge period of steam and air started. With noncondensable gas fractions above 3 % the damping of pressure oscillations inside the blowdown pipe was practically complete. Air quantities in the vicinity of 2 % started to have an effect also on the oscillations measured by the pressure sensor at the pool bottom. Both the amplitude and frequency of the pressure pulses decreased considerably. The experiments demonstrated that even small quantities of noncondensable gas can have a strong diminishing effect on pressure oscillations and structural loads

  17. FASTER test reactor preconceptual design report summary

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, Steven [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-29

    The FASTER reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  18. FASTER test reactor preconceptual design report summary

    International Nuclear Information System (INIS)

    The FASTER reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  19. Reactor Simulator Integration and Testing

    Science.gov (United States)

    Schoenfield, M. P.; Webster, K. L.; Pearson, J. B.

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator (RxSim) test loop was designed and built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing were to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V because the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This Technical Memorandum summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained, which was lower than the predicted 750 K but 156 K higher than the cold temperature, indicating the design provided some heat regeneration. The annular linear induction pump tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  20. Real time simulator for material testing reactor

    International Nuclear Information System (INIS)

    Japan Atomic Energy Agency (JAEA) is now developing a real time simulator for a material testing reactor based on Japan Materials Testing Reactor (JMTR). The simulator treats reactor core system, primary and secondary cooling system, electricity system and irradiation facility systems. Possible simulations are normal reactor operation, unusual transient operation and accidental operation. The developed simulator also contains tool to revise/add facility in it for the future development. (author)

  1. Pool swell sub-scale testing and code comparison

    International Nuclear Information System (INIS)

    The main objective of the experiment was to investigate the pool swell dynamics in general and the forces on the lowered central part of the diaphragm between drywell and wetwell in particular. Apart from the high speed camera pressure transducers and strain gauges were used to monitor the transient. Data was recorded on a 14 channel FM recorder and then digitalised and plotted. In total more than one hundred tests were performed including parametric variations of for example geometry, break flow, initial drywell pressure and initial water level. In parallel to this experiment pool swell calculations have been performed with the computer codes COPTA and STEALTH. COPTA which is a lumped mass code for pressure suppression containment analysis has a slug pool swell mode. STEALTH which is a general purpose lagrangian hydrodynamics code has been used in a 2-D axisymmetric version. The STEALTH code has been used to calculate the radial variations in the vertical displacement and velocity of the pool surface and to predict the load on the lowered central part of the diaphragm. A comparison between the calculations and the experimental data indicates that both codes are sufficiently correct in their description of the pool swell transient. (orig.)

  2. Experimental study of prompt neutron decay constant α for 300# pool reactor under mixed core

    International Nuclear Information System (INIS)

    The experimental study of prompt neutron decay constant α for 300# pool reactor under mixed core was carried out through a suit of reactor power spectral density measurement system. The two channel continuous current signals of neutron in the reactor were acquired by ionization chamber DL129 which was symmetrically putted in reactor core. The power spectral density, for two channel signals, was computed using the application program of data acquirement and data process analysis. Finally, by using the non-linear least squares method, the prompt neutron decay constant α was fitted. By comparison, the experimental results well accord to the theory calculation within the error range. The deviation can meet the actual need of project. (authors)

  3. Mark III confirmatory test program: one-third scale pool swell impact tests, Test Series 5805

    International Nuclear Information System (INIS)

    A series of 51 blowdown tests was performed in support of the Mark III pressure suppression concept with particular emphasis on the effect of pool swell impact on structures located above the suppression pool. The integrated steam generator and drywell of the Pressure Suppression Test Facility was used to accelerate the water mass in the one-third scale suppression pool to velocities typical of Mark III containments, and the impact of this water on I-beams, pipes, and gratings was investigated. The loading mechanism was found to be high velocity pressure waves which traveled along the surface of impacted structures, with a wave velocity defined by the movement of the points of intersection between the horizontal target structures and the rising curved pool surface. The impulse associated with this loading was found to correlate as a function of pool approach velocity, target geometry, and water ligament thickness, the last variable being important only when the ligament thickness approached target dimensions. For pool surface velocities expected to occur in Mark III, the maximum measured impulses for all targets were 35 percent or less of those being used for Mark III design specifications. For targets of circular cross section, loads were one-half or less than the values for comparable flat surfaces. Both the factor of three and the pipe shape factor must be considered when evaluating the conservatism in the Mark III design specifications

  4. Maintenance operation by divers on a swimming-pool type reactor (Osiris, CEN Saclay). Technical and medical prevention: an example of multidisciplinary ergonomic step

    International Nuclear Information System (INIS)

    Maintenance works in a swimming-pool reactor was performed by a team of divers. A multidisciplinary ergonomic study had previously defined the working procedure. The ergonomic approach is analysed. The divers' working techniques are described. After work, medical tests showed that previsions were verified and proved the methods as safe. This technique by divers' interventions should open new possibilities in nuclear industry

  5. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 16

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  6. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 26

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  7. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 8

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  8. Vernal Pool Study 2001 Wallkill River National Wildlife Refuge

    Data.gov (United States)

    US Fish and Wildlife Service, Department of the Interior — This is a series of data sheets from 2001 at Wallkill River National Wildlife Refuge that track and monitor the eggs and larvae of amphibians. The surveys also...

  9. Vernal Pool Study 2003 Wallkill River National Wildlife Refuge

    Data.gov (United States)

    US Fish and Wildlife Service, Department of the Interior — This is a series of data sheets from 2003 at Wallkill River National Wildlife Refuge that track and monitor the eggs and larvae of amphibians. The surveys also...

  10. Vernal Pool Study 2004 Wallkill River National Wildlife Refuge

    Data.gov (United States)

    US Fish and Wildlife Service, Department of the Interior — This is a series of data sheets from 2004 at Wallkill River National Wildlife Refuge that track and monitor the eggs and larvae of amphibians. The surveys also...

  11. Vernal Pool Study 2002 Wallkill River National Wildlife Refuge

    Data.gov (United States)

    US Fish and Wildlife Service, Department of the Interior — This is a series of data sheets from 2002 at Wallkill River National Wildlife Refuge that track and monitor the eggs and larvae of amphibians. The surveys also...

  12. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 9

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  13. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 2

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  14. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 15

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  15. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 20

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  16. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 12

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  17. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 13

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  18. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 10

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  19. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 21

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  20. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 18

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  1. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 24

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  2. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 5

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  3. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 14

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  4. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 7

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  5. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 6

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  6. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 4

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  7. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 1

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  8. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 19

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  9. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 22

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  10. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 3

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  11. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 17

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  12. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 25

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  13. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 11

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  14. Overview of pool hydraulic design of Indian prototype fast breeder reactor

    Indian Academy of Sciences (India)

    K Velusamy; P Chellapandi; S C Chetal; Baldev Raj

    2010-04-01

    Thermal hydraulics plays an important role in the design of liquid metal cooled fast breeder reactor components, where thermal loads are dominant. Detailed thermal hydraulic investigations of reactor components considering multi-physics heat transfer are essential for choosing optimum designs among the various possibilities. Pool hydraulics is multi-dimensional in nature and simple one-dimensional treatment for the same is often inadequate. Computational Fluid Dynamics (CFD) plays a critical role in the design of pool type reactors and becomes an increasingly popular tool, thanks to the advancements in computing technology. In this paper, thermal hydraulic characteristics of a fast breeder reactor, design limits and challenging thermal hydraulic investigations carried out towards successful design of Indian Prototype Fast Breeder Reactor (PFBR) that is under construction, are highlighted. Special attention is paid to phenomena like thermal stratification, thermal stripping, gas entrainment, inter-wrapper flow in decay heat removal and multiphysics cellular convection. The issues in these phenomena and the design solutions to address them satisfactorily are elaborated. Experiments performed for special phenomena, which are not amenable for CFD treatment and experiments carried out for validation of the computer codes have also been described.

  15. Assessing sediments from Upper Mississippi River navigational pools using a benthic invertebrate community evaluation and the sediment quality triad approach

    Science.gov (United States)

    Canfield, T.J.; Brunson, E.L.; Dwyer, F.J.; Ingersoll, C.G.; Kemble, N.E.

    1998-01-01

    Benthic invertebrate samples were collected from 23 pools in the Upper Mississippi River (UMR) and from one station in the Saint Croix River (SCR) as part of a study to assess the effects of the extensive flooding of 1993 on sediment contamination in the UMR system. Sediment contaminants of concern included both organic and inorganic compounds. Oligochaetes and chironomids constituted over 80% of the total abundance in samples from 14 of 23 pools in the UMR and SCR samples. Fingernail clams comprised a large portion of the community in three of 23 UMR pools and exceeded abundances of 1,000/m2 in five of 23 pools. Total abundance ranged from 250/m2 in samples from pool 1 to 22,389/m2 in samples from pool 19. Abundance values are comparable with levels previously reported in the literature for the UMR. Overall frequency of chironomid mouthpart deformities was 3% (range 0-13%), which is comparable to reported incidence of deformities in uncontaminated sediments previously evaluated. Sediment contamination was generally low in the UMR pools and the SCR site. Correlations between benthic measures and sediment chemistry and other abiotic parameters exhibited few significant or strong correlations. The sediment quality triad (Triad) approach was used to evaluate data from laboratory toxicity tests, sediment chemistry, and benthic community analyses; it showed that 88% of the samples were not scored as impacted based on sediment toxicity, chemistry, and benthic measures. Benthic invertebrate distributions and community structure within the UMR in the samples evaluated in the present study were most likely controlled by factors independent of contaminant concentrations in the sediments.

  16. Engineering, safety, and economic evaluations of ASPIRE [Advanced Safe Pool Immersed REactor

    International Nuclear Information System (INIS)

    A preconceptual design of a tokamak fusion reactor concept called ASPIRE (Advanced Safe Pool Immersed REactor) has been developed. This concept provides many of the attractive features that are needed to enhance the capability of fusion to become the power generation technology for the 21st century. Specifically, these features are: inherent safety, low pressure, environmental compatibility, moderate unit size, high availability, high thermal efficiency, simplicity, low radioactive inventory, Class C radioactive waste disposal, and low cost of electricity. We have based ASPIRE on a second stability tokamak. However, the concept is equally applicable to a first stability tokamak or to most other magnetic fusion systems

  17. Probabilistic analysis of some safety aspects of a swimming pool reactor

    International Nuclear Information System (INIS)

    A probabilistic risk analysis of some safety aspects without the investigation of radioactivity release has been performed for the 10 MW (thermal) swimming-pool research reactor SAPHIR. Our presentation is focused on the 7 internal initiating events found to be relevant with respect to accident sequences that could result with core melt due to loss of coolant or overcriticality. The results are given by the core melt frequencies for the investigated accident sequences. It could be demonstrated by our investigation that the core melt hazard of the reactor is extremely low. (author)

  18. Effect of reactivity insertion rate on peak power and temperatures in swimming pool type research reactor

    International Nuclear Information System (INIS)

    It is essential to study the reactor behavior under different accidental conditions and take proper measures for its safe operation. We have studied the effect of reactivity insertion, with and without scram conditions, on peak power and temperatures of fuel, cladding and coolant in typical swimming pool type research reactor. The reactivity ranging from 1 $ to 2 $ and insertion times from 0.25 to 1 second have been considered. The computer code PARET has been used and results are presented in this article. (author)

  19. Justify of implementation of a hot water layer system in swimming pool research reactor IEA-R1m

    International Nuclear Information System (INIS)

    The IPEN/CNEN-SP has a swimming pool research reactor (IEA-R1m) in operation since 1957 at 2 MW. In 1998, after some modifications, its nominal power increased to 5 MW. Among these modifications some adaptations had to be accomplished in the radiological protection and operational procedure. The present work aim to study the need of implementation of a hot water layer in order to reduce the dose in the workers in the vicinity of the reactor swimming pool. Applying the principles of radioprotection optimization, it was concluded that the decision of the construction of one hot water layer system in the reactor swimming pool, is not necessary. (author)

  20. SAVANNAH RIVER SITE R REACTOR DISASSEMBLY BASIN IN SITU DECOMMISSIONING

    Energy Technology Data Exchange (ETDEWEB)

    Langton, C.; Blankenship, J.; Griffin, W.; Serrato, M.

    2009-12-03

    The US DOE concept for facility in-situ decommissioning (ISD) is to physically stabilize and isolate in tact, structurally sound facilities that are no longer needed for their original purpose of, i.e., generating (reactor facilities), processing(isotope separation facilities) or storing radioactive materials. The 105-R Disassembly Basin is the first SRS reactor facility to undergo the in-situ decommissioning (ISD) process. This ISD process complies with the105-R Disassembly Basin project strategy as outlined in the Engineering Evaluation/Cost Analysis for the Grouting of the R-Reactor Disassembly Basin at the Savannah River Site and includes: (1) Managing residual water by solidification in-place or evaporation at another facility; (2) Filling the below grade portion of the basin with cementitious materials to physically stabilize the basin and prevent collapse of the final cap - Sludge and debris in the bottom few feet of the basin will be encapsulated between the basin floor and overlying fill material to isolate if from the environment; (3) Demolishing the above grade portion of the structure and relocating the resulting debris to another location or disposing of the debris in-place; and (4) Capping the basin area with a concrete slab which is part of an engineered cap to prevent inadvertent intrusion. The estimated total grout volume to fill the 105-R Reactor Disassembly Basin is 24,424 cubic meters or 31,945 cubic yards. Portland cement-based structural fill materials were design and tested for the reactor ISD project and a placement strategy for stabilizing the basin was developed. Based on structural engineering analyses and work flow considerations, the recommended maximum lift height is 5 feet with 24 hours between lifts. Pertinent data and information related to the SRS 105-R-Reactor Disassembly Basin in-situ decommissioning include: regulatory documentation, residual water management, area preparation activities, technology needs, fill material designs

  1. Determination of 16N and 19O activities in loop water of swimming pool reactor

    International Nuclear Information System (INIS)

    Measurements of activities for 16N and 19O nuclei in the loop water of swimming pool reactor at China Institute of Atomic Energy were carried out. In order to verify the experiment results, a calculation for same purpose was also performed. The results show their coincidence is well in uncertainty range. The evaluated recommendation data for 18O(n, γ)19O reaction cross sections are also given in the paper. (authors)

  2. Conceptual design of swimming pool type tokamak power reactor (SPTR-P)

    International Nuclear Information System (INIS)

    A preliminary design study of a tokamak power reactor utilizing the deuterium/tritium/lithium fuel cycle based on a swimming pool type reactor (SPTR) concept is presented. Its primary aim is to investigate the characteristics of the swimming-pool concept in which water replaces much of the steel normally required for shielding. The major design features are: steady state operation, RF wave for plasma heating and current drive, solid tritium breeder material (Li2O), modified austenitic stainless steel as first wall and blanket structural material, pumped limiter for ash exhaust, unified assembling of blanket and vacuum vessel and pressurized water cooling. The huge and heavy solid shield structure protecting superconducting magnets which brings about great difficulties in repair and maintenance is eliminated by submerging the reactor in a water pool. The water plays a role of shielding. In addition the water shield concept reduces radioactive waste disposal and to ease radiation streaming shielding. Key design parameters are: net electric power of 1000 MW, fusion power of 3200 MW, neutron wall loading of 3.3 MW/m2, major radius of 6.9 m, plasma radius of 2.0 m, plasma elongation of 1.6, plasma current of 16 MA, total beta of 7 %, toroidal field on axis of 5.2 T. (author)

  3. Design guide for Category III reactors: pool type reactors. [US DOE

    Energy Technology Data Exchange (ETDEWEB)

    Brynda, W J; Lobner, P R; Powell, R W; Straker, E A

    1978-11-01

    The Department of Energy (DOE) in the ERDA Manual requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirement of Category III reactor structures, components, and systems.

  4. Assembling a computerized adaptive testing item pool as a set of linear tests

    NARCIS (Netherlands)

    Linden, van der Wim J.; Ariel, Adelaide; Veldkamp, Bernard P.

    2006-01-01

    Test-item writing efforts typically results in item pools with an undesirable correlational structure between the content attributes of the items and their statistical information. If such pools are used in computerized adaptive testing (CAT), the algorithm may be forced to select items with less th

  5. Sipping test update device for fuel elements cladding inspections in IPR-r1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, R.R.; Mesquita, A.Z.; Andrade, E.P.D.; Gual, Maritza R., E-mail: rrr@cdtn.br, E-mail: amir@cdtn.br, E-mail: edson@cdtn.br, E-mail: maritzargual@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    It is in progress at the Centro de Desenvolvimento da Tecnologia Nuclear - CDTN (Nuclear Technology Development Center), a research project that aims to investigate possible leaks in the fuel elements of the TRIGA reactor, located in this research center. This paper presents the final form of sipping test device for TRIGA reactor, and results of the first experiments setup. Mechanical support strength tests were made by knotting device on the crane, charged with water from the conventional water supply, and tests outside the reactor pool with the use of new non-irradiated fuel elements encapsulated in stainless steel, and available safe stored in this unit. It is expected that tests with graphite elements from reactor pool are done soon after and also the test experiment with the first fuel elements in service positioned in the B ring (central ring) of the reactor core in the coming months. (author)

  6. Fuel Burnup and Fuel Pool Shielding Analysis for Bushehr Nuclear Reactor VVER-1000

    Science.gov (United States)

    Hadad, Kamal; Ayobian, Navid

    Bushehr Nuclear power plant (BNPP) is currently under construction. The VVER-1000 reactor will be loaded with 126 tons of about 4% enriched fuel having 3-years life cycle. The spent fuel (SF) will be transferred into the spent fuel pool (SPF), where it stays for 8 years before being transferred to Russia. The SPF plays a crucial role during 8 years when the SP resides in there. This paper investigates the shielding of this structure as it is designed to shield the SF radiation. In this study, the SF isotope inventory, for different cycles and with different burnups, was calculated using WIMS/4D transport code. Using MCNP4C nuclear code, the intensity of γ rays was obtained in different layers of SFP shields. These layers include the water above fuel assemblies (FA) in pool, concrete wall of the pool and water laid above transferring fuels. Results show that γ rays leakage from the shield in the mentioned layers are in agreement with the plant's PSAR data. Finally we analyzed an accident were the water height above the FA in the pool drops to 47 cm. In this case it was observed that exposure dose above pool, 10 and 30 days from the accident, are still high and in the levels of 1000 and 758 R/hr.

  7. Results of detailed ground geophysical surveys for locating and differentiating waste structures in waste management area 'A' at Chalk River Laboratories, Ontario

    Energy Technology Data Exchange (ETDEWEB)

    Tomsons, D.K.; Street, P.J.; Lodha, G.S

    1999-07-01

    Waste Management Area 'A' (WMA 'A'), located in the outer area of the Chalk River Laboratories (CRL) was in use as a waste burial site from 1946 to 1955. Waste management structures include debris-filled trenches, concrete bunkers and miscellaneous contaminated solid materials, and ditches and pits used for liquid dispersal. In order to update historical records, it was proposed to conduct detailed ground geophysical surveys to define the locations of waste management structures in WMA 'A', assist in planning of the drilling and sampling program to provide ground truth for the geophysics investigation and to predict the nature and locations of unknown/undefined shallow structures. A detailed ground geophysical survey grid was established with a total of 127 grid lines, oriented NNE and spaced one metre apart. The geophysical surveys were carried out during August and September, 1996. The combination of geophysical tools used included the Geonics EM61 metal detector, the GSM-19 magnetometer/gradiometer and a RAMAC high frequency ground penetrating radar system. The geophysical surveys were successful in identifying waste management structures and in characterizing to some extent, the composition of the waste. The geophysical surveys are able to determine the presence of most of the known waste management structures, especially in the western and central portions of the grid which contain the majority of the metallic waste. The eastern portion of the grid has a completely different geophysical character. While historical records show that trenches were dug, they are far less evident in the geophysical record. There is clear evidence for a trench running between lines 30E and 63E at 70 m. There are indications from the radar survey of other trench-like structures in the eastern portion. EM61 data clearly show that there is far less metallic debris in the eastern portion. The geophysical surveys were also successful in identifying

  8. Application of neutron noise analysis to a swimming pool research reactor

    International Nuclear Information System (INIS)

    This work is part of a programme of establishing practical applications of neutron noise techniques to a swimming pool research reactor and deals with two different items: (1) The identification of local boiling caused e.g. by a partial blockage of the coolant flow in a fuel element. Local boiling can easily lead to a burn-out situation. The onset of boiling can be detected by neutron noise analysis and a boiling detection system is presently under development. (2) The measurement of the time evolution of the reactivity induced by xenon after reactor shut-down by an on-line reactivity meter based on neutron noise analysis. From the data, the prompt neutron decay constant at delayed critical, the equilibrium xenon reactivity worth, and an estimate of the average steady-state power flux in the core before reactor shut-down were obtained. (author)

  9. Radiation shielding considerations for the repair and maintenance of a swimming pool-type tokamak reactor

    International Nuclear Information System (INIS)

    The radiation shielding relevant to the repair and maintenance of a swimming pool-type tokamak reactor is considered. The dose rate during the reactor operation can be made low enough for personnel access into the reactor room if a 2m thick water layer is installed above the magnet cryostat. The dose rate 24 h after shutdown is such that the human access is allowed above the magnet cryostat. Sufficient water layer thickness is provided in the inboard space for the operation of automatic welder/cutter while retaining the magnet shielding capability. Some forced cooling is required for the decay heat removal in the first wall. The penetration shield thickness around the neutral beam injector port is estimated to be barely sufficient in terms of the magnet radiation damage. (orig.)

  10. 33 CFR 207.320 - Mississippi River, Twin City Locks and Dam, St. Paul and Minneapolis, Minn.; pool level.

    Science.gov (United States)

    2010-07-01

    ... Locks and Dam, St. Paul and Minneapolis, Minn.; pool level. 207.320 Section 207.320 Navigation and... § 207.320 Mississippi River, Twin City Locks and Dam, St. Paul and Minneapolis, Minn.; pool level. In.... 362-Minn., Ford Motor Co.), this section is prescribed for the control of the pool level created...

  11. Savannah River Site K-Reactor Probabilistic Safety Assessment

    International Nuclear Information System (INIS)

    This report gives the results of a Savannah River Site (SRS) K-Reactor Probabilistic Safety Assessment (PSA). Measures of adverse consequences to health and safety resulting from representations of severe accidents in SRS reactors are presented. In addition, the report gives a summary of the methods employed to represent these accidents and to assess the resultant consequences. The report is issued to provide useful information to the U. S. Department of Energy (DOE) on the risk of operation of SRS reactors, for insights into severe accident phenomena that contribute to this risk, and in support of improved bases for other DOE programs in Heavy Water Reactor safety

  12. Savannah River Site K-Reactor Probabilistic Safety Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Brandyberry, M.D.; Bailey, R.T.; Baker, W.H.; Kearnaghan, D.P.; O`Kula, K.R.; Wittman, R.S.; Woody, N.D. [Westinghouse Savannah River Co., Aiken, SC (United States); Amos, C.N.; Weingardt, J.J. [Science Applications International Corp. (United States)

    1992-12-01

    This report gives the results of a Savannah River Site (SRS) K-Reactor Probabilistic Safety Assessment (PSA). Measures of adverse consequences to health and safety resulting from representations of severe accidents in SRS reactors are presented. In addition, the report gives a summary of the methods employed to represent these accidents and to assess the resultant consequences. The report is issued to provide useful information to the U. S. Department of Energy (DOE) on the risk of operation of SRS reactors, for insights into severe accident phenomena that contribute to this risk, and in support of improved bases for other DOE programs in Heavy Water Reactor safety.

  13. Physics aspects of reload and approach-to-critical of the NRU reactor after vessel repair

    International Nuclear Information System (INIS)

    The National Research Universal (NRU) reactor at Chalk River shut down on 2009 May 14 and there was a subsequent outage of 15 months to repair leaks from the vessel. On 2010 August 17, NRU returned to full power operation and resumed isotope production. This paper describes the physics aspects of reload, and the approach-to-critical (ATC) tests conducted to restart the reactor safely. Five ATC's, each at a different number of reloaded assemblies, plus a final one before reactor startup, were completed to confirm the calculated physics predictions of the subcritical state and critical point. Activities for preparation of the ATC tests, the responsibilities of the physicists during execution of the ATC's, and plots of neutron signal data during the ATC's are presented. The final measured critical point of CR 14 @190 cm agreed well with the calculated physics prediction of CR 14 @185 cm, or within ∼0.5 mk. (author)

  14. Numerical simulation of sodium pool fires in liquid metal-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    In Liquid Metal-Cooled Fast Breeder Reactor (LMFBR), the leakage of sodium can result in sodium fires. Due to sodium's high chemical reactivity in contact with air and water, sodium fires will lead to an immediate increase of the air temperature and pressure in the containment. This will harm the integrity of the containment. In order to estimate and foresee the sequence of this accident, or to prevent the accident and alleviate the influence of the accident, it is necessary to develop programs to analyze such sodium fire accidents. Based on the work of predecessors, flame sheet model is produced and used to analyze sodium pool fire accidents. Combustion model and heat transfer model are included and expatiated. And the comparison between the analytical and experimental results shows the program is creditable and reasonable. This program is more realistic to simulate the sodium pool fire accidents and can be used for nuclear safety judgement. (authors)

  15. Online failed fuel identification using delayed neutron detector signals in pool type reactors

    International Nuclear Information System (INIS)

    In todays world, nuclear reactors are at the forefront of modern day innovation and reactor designs are increasingly incorporating cutting edge technology. It is of utmost importance to detect failure or defects in any part of a nuclear reactor for healthy operation of reactor as well as the safety aspects of the environment. Despite careful fabrication and manufacturing of fuel pins, there is a chance of clad failure. After fuel pin clad rupture takes place, it allows fission products to enter in to sodium pool. There are some potential consequences due to this such as Total Instantaneous Blockage (TIB) of coolant and primary component contamination. At present, the failed fuel detection techniques such as cover gas monitoring (alarming the operator), delayed neutron detection (DND-automatic trip) and standalone failed fuel localization module (FFLM) are exercised in various reactors. The first technique is a quantitative measurement of increase in the cover gas activity background whereas DND system causes automatic trip on detecting certain level of activity during clad wet rupture. FFLM is subsequently used to identify the failed fuel subassembly. The later although accurate, but mainly suffers from downtime and reduction in power during identification process. The proposed scheme, reported in this paper, reduces the operation of FFLM by predicting the faulty sector and therefore reducing reactor down time and thermal shocks. The neutron evolution pattern gets modulated because fission products are the delay neutron precursors. When they travel along with coolant to Intermediate heat Exchangers, experienced three effects i.e. delay; decay and dilution which make the neutron pulse frequency vary depending on the location of failed fuel sub assembly. This paper discusses the method that is followed to study the frequency domain properties, so that it is possible to detect exact fuel subassembly failure online, before the reactor automatically trips. (author)

  16. Local heat transfer from the corium melt pool to the boiling water reactor pressure vessel wall

    International Nuclear Information System (INIS)

    The present study considers in-vessel accident progression after core melt relocation to the lower head of a Boiling Water Reactor (BWR) and formation of a melt pool containing a forest of Control Rod Guide Tubes (CRGTs) cooled by purging flows. Descending streams of melt that flow along cooled surfaces of CRGT, and impinge on the bottom surface of the vessel wall can significantly increase local heat transfer. The area of enhanced heat transfer enlarges with decreasing of the melt Prandtl (Pr) number, while the peaking value of the heat transfer coefficient is a non-monotone function of Pr number. The melt Pr number depends on the melt composition (fractions of metallic and oxidic melt components) and thus is inherently uncertain parameter of the core melting and relocation scenarios. The effect of Pr number in the range of 1.02 - 0.03 on the local and integral thermal loads on the vessel wall is examined using Computational Fluid Dynamics (CFD). Heat transfer models obtained on the base of CFD simulations are implemented in the Phase-change Effective Convectivity Model (PECM) for simulation of reactor-scale accident progression heat transfer in real 3D geometry of the BWR lower plenum. We found that the influence of the low Pr number on the thermal loads in a big melt pool becomes more significant at later time, than rapid acceleration of the creep in the vessel wall. This result suggests that global vessel failure is insensitive to the melt composition in the considered 0.7 m deep melt pool configuration. However, it is not clear yet if the low Pr number effect has an influence on vessel failure mode in the other possible melt pool configurations. (author)

  17. Reactor Safety Research Programs Quarterly Report January - March 1980

    Energy Technology Data Exchange (ETDEWEB)

    Hagen, C. M

    1980-10-01

    This document summarizes the work performed by Pacific Northwest Laboratory from January 1 through March 31, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where serviceinduced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  18. Inspection of state of spent fuel elements stored in RA reactor spent fuel storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Aden, V.G.; Bulkin, S.Yu.; Sokolov, A.V. [Research and Development Institute of Power Engineering, Moscow (Russian Federation); Matausek, M.V.; Vukadin, Z. [VINCA Institute of Nuclear Science, Belgrade (Yugoslavia)

    1999-07-01

    About five thousand spent fuel elements from RA reactor have been stored for over 30 years in sealed aluminum barrels in the spent fuel storage pool. This way of storage does not provide complete information about the state of spent fuel elements or the medium inside the barrels, like pressure or radioactivity. The technology has recently been developed and the equipment has been manufactured to inspect the state of the spent fuel and to reduce eventual internal pressure inside the aluminum barrels. Based on the results of this inspection, a procedure will be proposed for transferring spent fuel to a more reliable storage facility. (author)

  19. Material test reactor fuel research at the BR2 reactor

    International Nuclear Information System (INIS)

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

  20. Natural convection test in Phenix reactor and associated CATHARE calculation

    International Nuclear Information System (INIS)

    The Phenix sodium cooled fast reactor started operation in 1973 and was stopped in 2009. Before the reactor was definitively stopped, final tests were performed, including a natural convection test in the primary circuit. One objective of this natural convection test in Phenix reactor is the qualification of plant dynamic codes as CATHARE code for future safety studies. The paper firstly describes the Phenix reactor primary circuit. The initial test conditions and the detailed transient scenario are presented. Then, the CATHARE modelling of the Phenix primary circuit is described. The whole transient scenario is calculated, including the nominal state, the steam generators dry out, the scram, the onset of natural convection in the primary circuit and the natural convection phases. The CATHARE calculations are compared to the Phenix measurements. A particular attention is paid to the significant decrease of the core power before the scram. Then, the evolution of main components inlet and outlet temperatures is compared. The need of coupling a system code with a CFD code to model the 3D behaviour of large pools is pointed out. This work is in progress. (author)

  1. Criticality calculations for a spent fuel storage pool for a BWR type reactor

    International Nuclear Information System (INIS)

    In this work, the methodology for the calculation of the constant of effective multiplication for the arrangement of spent fuel assemblies in the pool of a BWR type reactor is shown. Calculations were done for the pool of spent fuel specified in FSAR and for the assemblies that is thought a conservative composition of high enrichment and without Gadolinium, giving credit to the stainless steel boxes of the frames that keep the assemblies. To carry out this simulation, RECORD and MIXQUIC codes were used. With record code, macroscopic cross sections, two energy groups, for the characteristics of the thought assemblies were obtained. Cross sections, as well as the dimensions of the frames that keep the fuel assemblies were used as input data for MIXQUIC code. With this code, criticality calculations in two dimensions were done, supposing that there is not leak of neutrons along the axial of the main line. Additional calculations, supposing changes in the temperature, distance among fuel assemblies and the thickness of the stainless steel box of the frame were done. The obtained results, including the effect in tolerances due to temperature, weight and thickness, show that the arrangement in the pool, when frames are fully charged, is subcritical by less than 5% in δK. (Author)

  2. LOSS-OF-COOLANT ACIDENT SIMULATIONS IN THE NATIONAL RESEARCH UNIVERSAL REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, W D; Goodman, R L; Heaberlin, S W; Hesson, G M; Nealley, C; Kirg, L L; Marshall, R K; McNair, G W; Meitzler, W D; Neally, G W; Parchen, L J; Pilger, J P; Rausch, W N; Russcher, G E; Schreiber, R E; Wildung, N J

    1981-02-01

    Pressurized water reactor loss-of-coolant accident (LOCA) phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship among the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. Subsequent experiments establish the fuel rod failure characteristics at selected peak cladding temperatures. Fuel rod cladding pressurization simulates high burnup fission gas pressure levels of modern PWRs. This document contains both an experiment overview of the LOCA simulation program and a review of the safety analyses performed by Pacific Northwest Laboratory (PNL) to define the expected operating conditions as well as to evaluate the worst case operating conditions. The primary intent of this document is to supply safety information required by the Chalk River Nuclear Laboratories (CRNL), to establish readiness to proceed from one test phase to the next and to establish the overall safety of the experiment. A hazards review summarizes safety issues, normal operation and three worst case accidents that have been addressed during the development of the experiment plan.

  3. Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

    2011-08-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  4. Mark I 1/12-scale pressure suppression pool swell test program: Phase IV tests

    International Nuclear Information System (INIS)

    Additional 1/12-scale Mark I pressure suppression pool swell tests have been conducted to supplement test results previously reported. A total of 68 tests were run. Thirty-two tests were run to expand the data base for tests with nominal initial conditions and with an initial drywell/wetwell differential pressure. Thirty-six tests were run to scope the effects of test section gross vertical stiffness and torus side plate stiffness on key test results. Test section structural stiffness characteristics were found to have a pronounced effect on the magnitude of the maximum downforce applied to the torus. Values for the maximum upforce applied to the torus, the maximum pool momentum, and the pool surface velocity at ring header impact were not found to be significantly affected by test section structural stiffness characteristics for the range of stiffnesses tested. For tests at nominal conditions, large normally distributed data bases for the maximum upforce, maximum pool momentum and ring header impact velocity are provided by these Phase IV and previous 1/12-scale tests. For these data bases one standard deviation is at most 10% of the mean value

  5. Mark I 1/12-scale pressure suppression pool swell test program: Phase IV tests

    Energy Technology Data Exchange (ETDEWEB)

    Galyardt, D.L.

    1977-03-01

    Additional /sup 1///sub 12/-scale Mark I pressure suppression pool swell tests have been conducted to supplement test results previously reported. A total of 68 tests were run. Thirty-two tests were run to expand the data base for tests with nominal initial conditions and with an initial drywell/wetwell differential pressure. Thirty-six tests were run to scope the effects of test section gross vertical stiffness and torus side plate stiffness on key test results. Test section structural stiffness characteristics were found to have a pronounced effect on the magnitude of the maximum downforce applied to the torus. Values for the maximum upforce applied to the torus, the maximum pool momentum, and the pool surface velocity at ring header impact were not found to be significantly affected by test section structural stiffness characteristics for the range of stiffnesses tested. For tests at nominal conditions, large normally distributed data bases for the maximum upforce, maximum pool momentum and ring header impact velocity are provided by these Phase IV and previous /sup 1///sub 12/-scale tests. For these data bases one standard deviation is at most 10% of the mean value.

  6. Clinch River Breeder Reactor Plant Project: construction schedule

    International Nuclear Information System (INIS)

    The construction schedule for the Clinch River Breeder Reactor Plant and its evolution are described. The initial schedule basis, changes necessitated by the evaluation of the overall plant design, and constructability improvements that have been effected to assure adherence to the schedule are presented. The schedule structure and hierarchy are discussed, as are tools used to define, develop, and evaluate the schedule

  7. Seismic evaluation of safety systems at the Savannah River reactors

    International Nuclear Information System (INIS)

    A thorough review of all safety related systems in commercial nuclear power plants was prompted by the accident at the Three Mile Island Nuclear Power Plant. As a consequence of this review, the Nuclear Regulatory Commission (NRC) focused its attention on the environmental and seismic qualification of the industry's electrical and mechanical equipment. In 1980, the NRC issued Unresolved Safety Issue (USI) A-46 to verify the seismic adequacy of the equipment required to safely shut down a plant and maintain a stable condition for 72 hours. After extensive research by the NRC, it became apparent that traditional analysis and testing methods would not be a feasible mechanism to address this USI A-46 issue. The costs associated with utilizing the standard analytical and testing qualification approaches were exorbitant and could not be justified. In addition, the only equipment available to be shake table testing which is similar to the item being qualified is typically the nuclear plant component itself. After 8 years of studies and data collection, the NRC issued its ''Generic Safety Evaluation Report'' approving an alternate seismic qualification approach based on the use of seismic experience data. This experience-based seismic assessment approach will be the basis for evaluating each of the 70 pre-1972 commercial nuclear power units in the United States and for an undetermined number of nuclear plants located in foreign countries. This same cost-effective developed for the commercial nuclear power industry is currently being applied to the Savannah River Production Reactors to address similar seismic adequacy issues. This paper documents the results of the Savannah River Plant seismic evaluating program. This effort marks the first complete (non-trial) application of this state-of-the-art USI A-46 resolution methodology

  8. PAH occurrence in chalk river systems from the Jura region (France). Pertinence of suspended particulate matter and sediment as matrices for river quality monitoring.

    Science.gov (United States)

    Chiffre, Axelle; Degiorgi, François; Morin-Crini, Nadia; Bolard, Audrey; Chanez, Etienne; Badot, Pierre-Marie

    2015-11-01

    This study investigates the variations of polycyclic aromatic hydrocarbon (PAH) levels in surface water, suspended particulate matter (SPM) and sediment upstream and downstream of the discharges of two wastewater treatment plant (WWTP) effluents. Relationships between the levels of PAHs in these different matrices were also investigated. The sum of 16 US EPA PAHs ranged from 73.5 to 728.0 ng L(-1) in surface water and from 85.4 to 313.1 ng L(-1) in effluent. In SPM and sediment, ∑16PAHs ranged from 749.6 to 2,463 μg kg(-1) and from 690.7 μg kg(-1) to 3,625.6 μg kg(-1), respectively. Investigations performed upstream and downstream of both studied WWTPs showed that WWTP discharges may contribute to the overall PAH contaminations in the Loue and the Doubs rivers. Comparison between gammarid populations upstream and downstream of WWTP discharge showed that biota was impacted by the WWTP effluents. When based only on surface water samples, the assessment of freshwater quality did not provide evidence for a marked PAH contamination in either of the rivers studied. However, using SPM and sediment samples, we found PAH contents exceeding sediment quality guidelines. We conclude that sediment and SPM are relevant matrices to assess overall PAH contamination in aquatic ecosystems. Furthermore, we found a positive linear correlation between PAH contents of SPM and sediment, showing that SPM represents an integrating matrix which is able to provide meaningful data about the overall contamination over a given time span.

  9. Flow of kinetic parameters in a typical swimming pool type research reactor

    International Nuclear Information System (INIS)

    Calculations were performed to estimate the variation in kinetic parameters (delayed neutron fraction and prompt neutron generation time) in different core configurations of a typical swimming pool type research reactor. Pakistan research Reactor-1 (PARR-1) was employed for this study. The effect due to burnup of the core was also studied. Calculations were performed with the help of computer codes WIMSD/4 and CITATION. Precursors yield was modified according to the neutron flux averaging only. This is the simple way to calculate the precursor yield for a particular core. The kinetic parameters are different for different core configurations. The βeff decreases with 1.33 x 10-6/% burnup whereas prompt neutron generation time increases with 6.42 x 10-8 s/% burnup. The results were compared with safety analysis report and with published values and were found in good agreement. This study provides the confidence to understand the change in the kinetic parameters of research reactors with core change and also with burnup of the core

  10. Summary Report for the 2003 Breeding Season Avian Point Count Survey at the Long Island Complex, Mississippi River Pool 21

    Data.gov (United States)

    US Fish and Wildlife Service, Department of the Interior — During the 2003 breeding season, a point count survey project was conducted in Pool 21 of the Upper Mississippi River, Adams County, Illinois. The study area was...

  11. COMMIX-1C code estimation for the pool dynamics of Istanbul Technical University TRIGA MARK-II reactor

    International Nuclear Information System (INIS)

    In this study, the COMMIX-1C code is used to investigate the pool dynamics of Istanbul Technical University (ITU)TRIGA MARK-II reactor by simulating the velocity, pressure and temperature distributions in the reactor pool as a function of core design parameters and pool configuration. COMMIX-1C is multi-purpose, three-dimensional. transient, single-phase, thermal-hydraulics computer code. For the mass, momentum and energy equations, it uses a porous-medium formulation, a finite-volume algorithm, a flow modulated skew-upwind discretization scheme to reduce numerical diffusion and k-ε two-equation turbulence model. Its implementation for the particular system requires geometric and physical modelling decisions. ITU TRIGA MARK-II reactor pool is considered partly as continuum and partly as porous medium. All the major pool components are explicitly modelled in the simulation. Shape of the pool structure and computational cells are accounted for using the concept of directional surface permeability, volume porosity, distributed resistance, and distributed heat source or sink. The results are compared to the results of the computer codes TRISTAN, TRIGATH and TRIGATH-R

  12. Reactor Safety Research Programs Quarterly Report July - September 1981

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1982-01-01

    This document summarizes the work performed by Pacific Northwest laboratory (PNL) from July 1 through September 30, 1981, for the Division of Accident Evaluation, U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR} steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  13. Reactor Safety Research Programs Quarterly Report October - December 1981

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1982-03-01

    This document summarizes the work performed by Pacific Northwest laboratory (PNL) from October 1 through December 31, 1981, for the Division of Accident Evaluation, U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where serviceinduced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and post accident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  14. Full-length high-temperature severe fuel damage test No. 1

    Energy Technology Data Exchange (ETDEWEB)

    Rausch, W.N.; Hesson, G.M.; Pilger, J.P.; King, L.L.; Goodman, R.L.; Panisko, F.E.

    1993-08-01

    This report describes the first full-length high-temperature test (FLHT-1) performed by Pacific Northwest Laboratory (PNL) in the National Research Universal (NRU) reactor at Chalk River, Ontario, Canada. The test is part of a series of experiments being performed for the NRC as a part of their Severe Fuel Damage Program and is one of several planned for PNL`s Coolant Boilaway and Damage Progression Program. The report summarizes the test design and test plan. it also provides a summary and discussion of the data collected during the test and of the photos taken during the post-test examination. All objectives for the test were met. The key objective was to demonstrate that severe fuel damage tests on full-length fuel bundles can be safely conducted in the NRU reactor.

  15. L-Reactor operation, Savannah River Plant: environmental assessment

    International Nuclear Information System (INIS)

    The purpose of this document is to assess the significance of the effects on the human environment of the proposed resumption of L-reactor operation at the Savannah River Plant, scheduled for October 1983. The discussion is presented under the following section headings: need for resumption of L-Reactor operations and purpose of this environmental assessment; proposed action and alternative; affected environment (including, site location and description, land use, historic and archeological resources, socioeconomic and community characteristics, geology and seismology, hydrology, meteorology and climatology, ecology, and radiation environment); environmental consequences; summary of projected L-Reactor releases and impacts; and Federal and State permits and approval. The three appendices are entitled: radiation dose calculation methods and assumptions; floodplain/wetlands assessment - L-Reactor operations; and, conversion table. A list of references is included at the end of each chapter

  16. Compaction of microfossil and clay-rich chalk sediments

    DEFF Research Database (Denmark)

    Fabricius, Ida Lykke

    2001-01-01

    of microfossils and fine-grained silica and clay. Samples with relatively pure chalk mud supported texture compact along a common stress - matrix porosity trend. Microfossils thus have a passive role, apparently because they are supported by the chalk mud. Samples with fine-grained silica and clay can be modelled...... to Follow the same trend if we assume that a part of the fine-grained silica and clay are in the supporting frame and that the remaining silica and clay has a passive pore-filling role. The modelled part of the day and silica in the frame varies from 0% to 100%. Porosity and sonic velocity variations......The aim of this study was to evaluate the role of microfossils and clay in the compaction of chalk facies sediments. To meet this aim, chalk sediments with varying micro texture were studied. The sediments have been tested uniaxially confined in a stainless-steel compaction cell. The sediments are...

  17. Westinghouse independent safety review of Savannah River production reactors

    International Nuclear Information System (INIS)

    Westinghouse Electric Corporation has performed a safety assessment of the Savannah River production reactors (K, L, and P) as requested by the US Department of Energy. This assessment was performed between November 1, 1988, and April 1, 1989, under the transition contract for the Westinghouse Savannah River Company's preparations to succeed E.I. du Pont de Nemours ampersand Company as the US Department of Energy contractor for the Savannah River Project. The reviewers were drawn from several Westinghouse nuclear energy organizations, embody a combination of commercial and government reactor experience, and have backgrounds covering the range of technologies relevant to assessing nuclear safety. The report presents the rationale from which the overall judgment was drawn and the basis for the committee's opinion on the phased restart strategy proposed by E.I. du Pont de Nemours ampersand Company, Westinghouse, and the US Department of Energy-Savannah River. The committee concluded that it could recommend restart of one reactor at partial power upon completion of a list of recommended upgrades both to systems and their supporting analyses and after demonstration that the organization had assimilated the massive changes it will have undergone. 37 refs., 1 fig., 3 tabs

  18. Westinghouse independent safety review of Savannah River production reactors

    Energy Technology Data Exchange (ETDEWEB)

    Leggett, W.D.; McShane, W.J. (Westinghouse Hanford Co., Richland, WA (USA)); Liparulo, N.J.; McAdoo, J.D.; Strawbridge, L.E. (Westinghouse Electric Corp., Pittsburgh, PA (USA). Nuclear and Advanced Technology Div.); Toto, G. (Westinghouse Electric Corp., Pittsburgh, PA (USA). Nuclear Services Div.); Fauske, H.K. (Fauske and Associates, Inc., Burr Ridge, IL (USA)); Call, D.W. (Westinghouse Savannah R

    1989-04-01

    Westinghouse Electric Corporation has performed a safety assessment of the Savannah River production reactors (K,L, and P) as requested by the US Department of Energy. This assessment was performed between November 1, 1988, and April 1, 1989, under the transition contract for the Westinghouse Savannah River Company's preparations to succeed E.I. du Pont de Nemours Company as the US Department of Energy contractor for the Savannah River Project. The reviewers were drawn from several Westinghouse nuclear energy organizations, embody a combination of commercial and government reactor experience, and have backgrounds covering the range of technologies relevant to assessing nuclear safety. The report presents the rationale from which the overall judgment was drawn and the basis for the committee's opinion on the phased restart strategy proposed by E.I. du Pont de Nemours Company, Westinghouse, and the US Department of Energy-Savannah River. The committee concluded that it could recommend restart of one reactor at partial power upon completion of a list of recommended upgrades both to systems and their supporting analyses and after demonstration that the organization had assimilated the massive changes it will have undergone.

  19. Item Pool Design for an Operational Variable-Length Computerized Adaptive Test

    Science.gov (United States)

    He, Wei; Reckase, Mark D.

    2014-01-01

    For computerized adaptive tests (CATs) to work well, they must have an item pool with sufficient numbers of good quality items. Many researchers have pointed out that, in developing item pools for CATs, not only is the item pool size important but also the distribution of item parameters and practical considerations such as content distribution…

  20. Transient analysis of dissolution of a reactor bottom head into a melt pool

    International Nuclear Information System (INIS)

    The dissolution of the bottom head of a heavy-water reactor into a pool of molten fuel under severe accident conditions is investigated using a distributed-parameter model. The main objectives are to determine the rate of dissolution-front propagation and to estimate the extent to which the bottom head is thinned owing to dissolution. The model consists mainly of partitioning the bottom head into a number of rings and analyzing the transient dissolution of each ring with a localized lumped-parameter model. For each of the rings, the dissolution is modeled using a mass transfer coefficient, the temperature distribution is considered to be one dimensional and quasisteady and the heat flux across the melt-bottom head interface is modeled using a heat transfer coefficient. The distribution of the heat transfer coefficients is considered to be quasi-steady and is based on the heat transfer calculation results obtained using the ACCORD code. The model thus takes into account both the variation of heat fluxes over the melt pool-bottom head interface and the variations of interface mass transfer with time and with position along the interface. The basic equations and their solution method for the distributed-parameter model are described. Comparisons of calculation results with those obtained previously using the overall lumped-parameter model are presented

  1. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system (Bragg-Sitton, 2005). The current paper applies the same testing methodology to a direct drive gas cooled reactor system, demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. In each testing application, core power transients were controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. Although both system designs utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility.

  2. Assessment of simulation predictions of hydrocarbon pool fire tests.

    Energy Technology Data Exchange (ETDEWEB)

    Luketa-Hanlin, Anay Josephine

    2010-04-01

    An uncertainty quantification (UQ) analysis is performed on the fuel regression rate model within SIERRA/Fuego by comparing to a series of hydrocarbon tests performed in the Thermal Test Complex. The fuels used for comparison for the fuel regression rate model include methanol, ethanol, JP8, and heptane. The recently implemented flamelet combustion model is also assessed with a limited comparison to data involving measurements of temperature and relative mole fractions within a 2-m diameter methanol pool fire. The comparison of the current fuel regression rate model to data without UQ indicates that the model over predicts the fuel regression rate by 65% for methanol, 63% for ethanol, 95% for JP8, and 15% for heptane. If a UQ analysis is performed incorporating a range of values for transmittance, reflectance, and heat flux at the surface the current model predicts fuel regression rates within 50% of measured values. An alternative model which uses specific heats at inlet and boiling temperatures respectively and does not approximate the sensible heat is also compared to data. The alternative model with UQ significantly improves the comparison to within 25% for all fuels except heptane. Even though the proposed alternative model provides better agreement to data, particularly for JP8 and ethanol (within 15%), there are still outstanding issues regarding significant uncertainties which include heat flux gauge measurement and placement, boiling at the fuel surface, large scale convective motion within the liquid, and semi-transparent behavior.

  3. The sedimentology of redeposited chalk

    DEFF Research Database (Denmark)

    Anderskouv, Kresten; Surlyk, Finn; Gale, Andy

    , interpretation, and predictability of redeposited chalk facies remain uncertain. This project aims to improve existing facies models by investigating and comparing redeposited chalk units from a variety of settings. Long cores from the Danish and British sectors are interpreted in terms of depositional process......Redeposited facies in the Upper Cretaceous Chalk Group constitute major hydrocarbon reservoirs in the North Sea Central Graben. Existing facies models are largely based on publications from the early 1980's dealing with core material from the Norwegian sector. However, the recognition...

  4. Competitive sorption of organic contaminants in chalk

    Science.gov (United States)

    Graber, E. R.; Borisover, M.

    2003-12-01

    In the Negev desert, Israel, a chemical industrial complex is located over fractured Eocene chalk formations where transfer of water and solutes between fracture voids and matrix pores affects migration of contaminants in the fractures due to diffusion into the chalk matrix. This study tests sorption and sorption competition between contaminants in the chalk matrix to make it possible to evaluate the potential for contaminant attenuation during transport in fractures. Single solute sorption isotherms on chalk matrix material for five common contaminants ( m-xylene, ametryn, 1,2-dichloroethane, phenanthrene, and 2,4,6-tribromophenol) were found to be nonlinear, as confirmed in plots of Kd versus initial solution concentration. Over the studied concentration ranges, m-xylene Kd varied by more than a factor of 100, ametryn Kd by a factor of 4, 1,2-dichloroethane Kd by more than a factor of 3, phenanthrene Kd by about a factor of 2, and 2,4,6-tribromophenol Kd by a factor of 10. It was earlier found that sorption is to the organic matter component of the chalk matrix and not to the mineral phases (Chemosphere 44 (2001) 1121). Nonlinear sorption isotherms indicate that there is at least some finite sorption domain. Bi-solute competition experiments with 2,4,6-tribromophenol as the competitor were designed to explore the nature of the finite sorption domain. All of the isotherms in the bi-solute experiments are more linear than in the single solute experiments, as confirmed by smaller variations in Kd as a function of initial solution concentration. For both m-xylene and ametryn, there is a small nonlinear component or domain that was apparently not susceptible to competition by 2,4,6-tribromophenol. The nonlinear sorption domain(s) is best expressed at low solution concentrations. Inert-solvent-normalized single and bi-solute sorption isotherms demonstrate that ametryn undergoes specific force interactions with the chalk sorbent. The volume percent of phenanthrene

  5. Permeability prediction in chalks

    DEFF Research Database (Denmark)

    Alam, Mohammad Monzurul; Fabricius, Ida Lykke; Prasad, Manika

    2011-01-01

    The velocity of elastic waves is the primary datum available for acquiring information about subsurface characteristics such as lithology and porosity. Cheap and quick (spatial coverage, ease of measurement) information of permeability can be achieved, if sonic velocity is used for permeability......-permeability relationships were replaced by relationships between velocity of elastic waves and permeability using laboratory data, and the relationships were then applied to well-log data. We found that the permeability prediction in chalk and possibly other sediments with large surface areas could be improved...... significantly using the effective specific surface as the fluid-flow concept. The FZI unit is appropriate for highly permeable sedimentary rocks such as sandstones and limestones that have small surface areas....

  6. Operating experience of Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt / 13.2 MWe sodium cooled, loop type mixed carbide fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors and to serve as an irradiation facility for development of fuel and structural material for future fast reactors. The reactor achieved first criticality in October 1985 with small indigenously designed and fabricated Mark I core (70% PuC-30% UC). The reactor power was subsequently raised in steps to 17.4 MWt by addition of Mark II fuel subassemblies (55% PuC-45% UC) and with the Mark I fuel operating at the designed linear heat rating of 400 W/cm. The turbo-generator was synchronized with the grid in July 1997. The achieved peak burn-up is 137 000 MWd/t so far without any fuel-clad failure. Presently the reactor is being operated at a nominal power of 15.7 MWt for irradiation of a test fuel subassembly of the Prototype Fast Breeder Reactor, which is coming up at Kalpakkam. It is also planned to irradiate test subassemblies made of metallic fuel for future fast reactor program. Being a small reactor, all feed back coefficients of reactivity including void coefficient are negative and hence the reactor is inherently safe. This was confirmed by carrying out physics tests. The capability to remove decay heat under various incidental conditions including natural convection was demonstrated by carrying out engineering tests. Thermo couples are provided for on-line monitoring of fuel SA outlet temperature by dedicated real time computer and processed to generate trip signals for the reactor in case of power excursion, increase in clad hot spot temperature and subassembly flow blockage. All pipelines and capacities in primary main circuit are provided with segmented outer envelope to minimize and contain radioactive sodium leak while ensuring forced cooling through reactor to remove decay heat in case of failure of primary boundary. In secondary circuit, provision is

  7. Research of Distribution of Elements in Natural Waters of the Selenga River Pool

    CERN Document Server

    Ganbold, G; Gerbish, S; Dalhsuren, B; Bayarmaa, Z; Maslov, O D; Sevastiyanov, D V

    2001-01-01

    The distribution of heavy metals in natural waters of the Selenga river pool was investigated. The contents of elements were determined using X-ray analysis with complete external reflection (XRACER). The zones with excess of the average contents of elements in comparison with reference samples were found out, that specifies their pollution by metals. It is offered in these zones to organize the regular water quality monitoring for supervision over the condition of the water ecosystems and to carry out actions on decrease of anthropogenous load and pollution of natural waters.

  8. The radionuclides of primary coolant in HANARO and the recent activities performed to reduce the radioactivity or reactor pool water

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minjin [HANARO Research Reactor Centre, Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

    1998-10-01

    In HANARO reactor, there have been activities to identify the principal radionuclides and to quantify them under the normal operation. The purposes of such activities were to establish the measure by which we can reduce the radioactivity of the reactor pool water and detect, in early stage, the abnormal symptoms due to the leakage of radioactive materials from the irradiation sample or the damage of the nuclear fuel, etc. The typical radionuclides produced by the activation of reactor coolant are N{sup 16} and Ar{sup 41}. The radionuclides produced by the activation of the core structural material consist of Na{sup 24}, Mn{sup 56}, and W{sup 187}. Of the various radionuclides, governing the radiation level at the pool surface are Na{sup 24}, Ar{sup 41}, Mn{sup 58}, and W{sup 187}. By establishing the hot water layer system on the pool surface, we expected that the radionuclides such as Ar{sup 41} and Mn{sup 56} whose half-life are relatively short could be removed to a certain extent. Since the content of radioactivity of Na{sup 24} occupies about 60% of the total radioactivity, we assumed that the total radiation level would be greatly reduced if we could decrease the radiation level of Na{sup 24}. However the actual radiation level has not been reduced as much as we expected. Therefore, some experiments have been carried out to find the actual causes afterwards. What we learned through the experiments are that any disturbance in reactor pool water layer causes increase of the pool surface radiation level and even if we maintain the hot water layer well, reactor shutdown will be very much likely to happen once the hot water layer is disturbed. (author)

  9. ORIGEN2 model and results for the Clinch River Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A G; Bjerke, M A

    1982-06-01

    Reactor physics calculations and literature information acquisition have led to the development of a Clinch River Breeder Reactor (CRBR) model for the ORIGEN2 computer code. The model is based on cross sections taken directly from physics codes. Details are presented concerning the physical description of the fuel assemblies, the fuel management scheme, irradiation parameters, and initial material compositions. The ORIGEN2 model for the CRBR has been implemented, resulting in the production of graphical and tabular characteristics (radioactivity, thermal power, and toxicity) of CRBR spent fuel, high-level waste, and fuel-assembly structural material waste as a function of decay time. Characteristics for pressurized water reactors (PWRs), commercial liquid-metal fast breeder reactors (LMFBRs), and the Fast Flux Test Facility (FFTF) have also been included in this report for comparison with the CRBR data.

  10. Experiments on in-vessel melt pool formation and behavior in the LIVE-2D test facility

    International Nuclear Information System (INIS)

    Experiments were carried out to study natural convection heat transfer and crust formation at the boundaries of an internally heated molten pool using a non-eutectic melt (KNO3- NaNO3) as a simulant fluid. The experiments were performed in the LIVE-2D experimental facility in a semi-circular slice scaled 1/4 to a prototypic PWR type reactor. Besides the transient molten pool behavior, for which the LIVE-2D tests provide qualified data on temperature evolution in the molten pool and crust growth rates, the experiments address other important phenomena, such as the local distribution of heat flux, and the influence of solidification on the thermal-hydraulics of the pool. The effect of variation of heat input and boundary cooling conditions during and immediately after the pouring phase were studied for internal Rayleigh numbers varying from 5.1013 to 2.1014. These effects are important for the assessment of the reactor vessel integrity in case of a core melt accident and for the feasibility of an accident management strategy using in-vessel melt retention by cooling the lower head from outside. The experimental results provide the Ra-Nu correlations for the upward/downward heat flux distribution and show the influence of the heating regime on the crust thickness formed at the molten pool/vessel wall boundary. The experimental results are being used for development of mechanistic models for description of in-core molten pool behavior and their implementation in the severe accident codes like ASTEC. The paper summarizes the objectives of the LIVE-2D experiments and presents the main results obtained up to now. (author)

  11. Keeping research reactors relevant: A pro-active approach for SLOWPOKE-2

    International Nuclear Information System (INIS)

    The SLOWPOKE is a small, inherently safe, pool-type research reactor that was engineered and marketed by Atomic Energy of Canada Limited (AECL) in the 1970s and 80s. The original reactor, SLOWPOKE-1, was moved from Chalk River to the University of Toronto in 1970 and was operated until upgraded to the SLOWPOKE-2 reactor in 1973. In all, eight reactors in the two versions were produced and five are still in operation today, three having been decommissioned. All of the remaining reactors are designated as SLOWPOKE-2 reactors. These research reactors are prone to two major issues: aging components and lack of relevance to a younger audience. In order to combat these problems, one SLOWPOKE -2 facility has embraced a strategy that involves modernizing their reactor in order to keep the reactor up to date and relevant. In 2001, this facility replaced its aging analogue reactor control system with a digital control system. The system was successfully commissioned and has provided a renewed platform for student learning and research. The digital control system provides a better interface and allows flexibility in data storage and retrieval that was never possible with the analogue control system. This facility has started work on another upgrade to the digital control and instrumentation system that will be installed in 2010. The upgrade includes new computer hardware, updated software and a web-based simulation and training system that will allow licensed operators, students and researchers to use an online simulation tool for training, education and research. The tool consists of: 1) A dynamic simulation for reactor kinetics (e.g., core flux, power, core temperatures, etc). This tool is useful for operator training and student education; 2) Dynamic mapping of the reactor and pool container gamma and neutron fluxes as well as the vertical neutron beam tube flux. This research planning tool is used for various researchers who wish to do irradiations (e.g., neutron

  12. Present status of Japan materials testing reactor

    International Nuclear Information System (INIS)

    The Japan Materials Testing Reactor (JMTR) in Japan Atomic Energy Agency (JAEA) is a light water cooled tank type reactor with first criticality in March 1968. Owing to the connection between the JMTR and hot laboratory by a canal, easy re-irradiation tests can be conducted with safe and quick transportation of irradiated samples. The JMTR has been applied to fuel/material irradiation examinations for LWRs, HTGR, fusion reactor and RI production. However, the JMTR operation was once stopped in August 2006, and check and review on the reoperation had been conducted by internal as well as external committees. As a result of the discussion, the JMTR reoperation was determined, and refurbishment works started from the beginning of JFY 2007. The refurbishment works have finished in March 2011 taking four years from JFY 2007. Unfortunately, at the end of the JFY 2010 on March 11, the Great-Eastern-Japan-Earthquake occurred, and functional tests before the JMTR restart, such as cooling system, reactor control system and so on, were delayed by the earthquake. Moreover, a detail inspection found some damages such as slight deformation of the truss structure at the roof of the JMTR reactor building. Consequently, the restart of the JMTR will be delayed from June to next October, 2012. Now, the safety evaluation after the earthquake disaster is being carried out aiming at the restart of the JMTR. The renewed JMTR will be started from JFY 2012 and operated for a period of about 20 years until around JFY 2030. The usability improvement of the JMTR, e.g. higher reactor availability, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussed with users as the preparations for re-operation. (author)

  13. Experimental study for research and development of a super fast reactor. (2) Oscillatory condensation of high temperature vapor directly discharged into sub-cooled liquid pool

    International Nuclear Information System (INIS)

    The measurement of pressure oscillation and the observation of condensation behavior of a vapor discharge into sub-cooled liquid cool has been carried out to obtain a basic data for the evaluation of safety of the LOCA in the supercritical pressure light water cooled fast reactor (Super Fast Reactor). In the experiment, HCFC 123 is used as the test fluid. HCFC 123 is easy for handling due to its low critical pressure and temperature, and therefore, the experimental conditions can be set easily to make systematic data. The vapor at high temperature is discharged into the sub-cooled liquid pool through a submerged single pipe vertically fixed. The oscillatory condensation is observed. The condensation oscillation produces pressure oscillation in the liquid pool. The condensing interface area becomes small as the increase of the degree of sub-cooling. The pressure frequency has a period of millisecond order and the frequency and amplitude of the pressure oscillation increase with increasing the degree of sub-cooling and mass flux of the vapor, like the results of some conventional water vapor injection tests. In the present study, it is also consistently discussed the influence of the vapor temperature, mass flux, mass flow rate, back pressure of the liquid pool, pipe diameter and the degree of sub-cooling on the pressure amplitude and condensation behavior. (author)

  14. Thermal insulation system design and fabrication specification (nuclear) for the Clinch River Breeder Reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    1978-07-21

    This specification defines the design, analysis, fabrication, testing, shipping, and quality requirements of the Insulation System for the Clinch River Breeder Reactor Plant (CRBRP), near Oak Ridge, Tennessee. The Insulation System includes all supports, convection barriers, jacketing, insulation, penetrations, fasteners, or other insulation support material or devices required to insulate the piping and equipment cryogenic and other special applications excluded. Site storage, handling and installation of the Insulation System are under the cognizance of the Purchaser.

  15. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in the development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe (HP) cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system. Reactivity feedback calculations were then based on a bulk reactivity feedback coefficient and measured average core temperature. This paper presents preliminary results from similar dynamic testing of a direct drive gas cooled reactor system (DDG), demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. Although the HP and DDG designs both utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility. Planned system upgrades to allow implementation of higher fidelity dynamic testing are also discussed. Proposed DDG

  16. Axial power monitoring uncertainty in the Savannah River Reactors

    International Nuclear Information System (INIS)

    The results of this analysis quantified the uncertainty associated with monitoring the Axial Power Shape (APS) in the Savannah River Reactors. Thermocouples at each assembly flow exit map the radial power distribution and are the primary means of monitoring power in these reactors. The remaining uncertainty in power monitoring is associated with the relative axial power distribution. The APS is monitored by seven sensors that respond to power on each of nine vertical Axial Power Monitor (APM) rods. Computation of the APS uncertainty, for the reactor power limits analysis, started with a large database of APM rod measurements spanning several years of reactor operation. A computer algorithm was used to randomly select a sample of APSs which were input to a code. This code modeled the thermal-hydraulic performance of a single fuel assembly during a design basis Loss-of Coolant Accident. The assembly power limit at Onset of Significant Voiding was computed for each APS. The output was a distribution of expected assembly power limits that was adjusted to account for the biases caused by instrumentation error and by measuring 7 points rather than a continuous APS. Statistical analysis of the final assembly power limit distribution showed that reducing reactor power by approximately 3% was sufficient to account for APS variation. This data confirmed expectations that the assembly exit thermocouples provide all information needed for monitoring core power. The computational analysis results also quantified the contribution to power limits of the various uncertainties such as instrumentation error

  17. High flux testing reactor Petten. Replacement of the reactor vessel and connected components. Overall report

    International Nuclear Information System (INIS)

    The project of replacing the HFR originated in 1974 when results of several research programmes confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report contains the detailed chronology of events concerning preparation and execution of the replacement. After a 14 months' outage the reactor resumed routine operation on 14th February, 1985. At the end of several years of planning and preparation the reconstruction proceded in the following steps: unloading of the old core, decay of short-lived radioactivity in December 1983, removal of the old tank and of its peripheral equipment in January-February 1984, segmentation and waste disposal of the removed components in March-April, decontamination of the pools, bottom penetration overhauling in May-June, installation of the new tank and other new components in July-September, testing and commissioning, including minor modifications in October-December, and, trials runs and start-up preparation in January-February 1985. The new HFR Petten features increased and improved experimental facilities. Among others the obsolete thermal columns was replaced by two high flux beam tubes. Moreover the new plant has been designed for future increases of reactor power and neutron fluxes. For the next three to four years the reactor has to cope with a large irradiation programme, claiming its capacity to nearly 100%

  18. Less chalk more action

    Science.gov (United States)

    Mitriceski Andelkovic, Bojana; Jovic, Sladjana

    2016-04-01

    Less chalk more action Education should not be a mechanical system that operates according to the principles of the orders and implementation. Education should respect the basic laws of the develop and progress. Curiosity is the engine of achievement and children spontaneously and happily learn only if they get interested, if teacher wake up and stimulate their creativity and individuality. We would like to present classes that are realized as thematic teaching with several subjects involved: chemistry, geography, math, art and biology. Classes were organized for students at age from 10 to 13 years, every month during autumn and winter 2015. Better students identified themselves as teachers and presented peer education .Teachers were monitoring the process of teaching and help to develop links between younger and older students, where older students were educators to younger students. Also one student with special needs was involved in this activities and was supported by other students during the workshops The benefit from this project will be represented with evaluation marks. Evaluation table shows that group of ten students(age 10 to13 years) which are selected in October as children with lack of motivation for learning, got better marks, at the end of January , then they had it in the beginning of the semester.

  19. Ageing Management and Preventice Measures for Reactor Pool Liners, Beam Tubes and Spent Fuel Storage Tank at the Dalat Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dien, Nguyen Nhi; Dien, Nguyen Minh; Su, Trang Cao [Nuclear Research Institute, Henoi (Viet Nam)

    2013-07-01

    The 500-kw Dalat Nuclear Research Reactor (DNRR) was reconstructed from the original 250-kW TRIGA Mark II as named of VN-001. In the framework of the reconstruction project during the 1982-1984 period, some structures of the TRIGA reactor constructed in the early sixties, such as the aluminum tank, graphite reflector, thermal column, four horizontal beam tubes, etc. have been remained. It means, such components are more than 50 years old and are facing with ageing issues. The structural materials of the pool liner and other components of TRIGA were made of aluminum alloy 6061 and aluminum cladding fuel assemblies. Some other parts, such as reactor core, irradiation rotary rack around the core, vertical irradiation facilities, etc. were replaced by the former Soviet Union's design with structural materials of aluminum alloy CAV-1. The reactor core has been loaded with HEU VVR-M2 fuel assemblies of 36% enrichment alloy CAV-1. The reactor core has been loaded with HEU VVR-M2 fuel assemblies of U-Al alloy 36% and of UO{sub 2} 19.75% enrichment used aluminum as fuel cladding. For ageing management and preventive measures of corrosion, an underwater high-resolution video camera system had been designed for visual inspections. A home-made cleaning system was also designed for cleaning the pool and other components. Water chemistry of the reactor pool and spent fuel storage was monitored regularly. In September-November 2011, all four horizontal channels were cleaned inside and visual inspection was done using special camera system. It was the first time from 1963 such activity could be done. Based on results obtained we could convince that inside all horizontal channels are in good condition and leakage could not be occurred. All 106 HEU spent fuel assemblies stored in the spent fuel pool in good condition. The visual inspection was done using under water camera too. The results obtained show that the surface of all HEU SFA is good and leakage was not occurred. The

  20. Sidestream Elevated Pool Aeration, a Technology for Improving Water Quality in Urban Rivers

    Science.gov (United States)

    Motta, D.; Garcia, T.; Abad, J. D.; Bombardelli, F. A.; Waratuke, A.; Garcia, M. H.

    2010-12-01

    Dissolved Oxygen (DO) levels are frequently depleted in rivers located in urban areas, as in the case of the Matanza-Riachuelo River in Buenos Aires, Argentina. This stream receives both domestic and industrial loads which have received minor or no treatment before being discharged into the water body. Major sources of pollution include, but are not limited, to leather and meat packing industries. Additionally, deep slow moving water in the river is associated with limited reaeration and facilitates deposition of organic-rich sediment, therefore exacerbating the DO consumption through sediment oxygen demand. In this study we assessed the efficiency of Sidestream Elevated Pool Aeration (SEPA) stations as a technology for alleviating conditions characterized by severely low DO levels. A SEPA station takes water from the stream at low DO concentrations, through a screw pump; then, water is transported to an elevated pool from where it flows over a series of weirs for water reaeration; finally, the aerated water is discharged back into the river sufficiently downstream from the intake point. This system mimics a phenomenon that occurs in mountain streams, where water is purified by bubbling over rocks. The impact of the use of SEPA stations on the DO concentrations in the Matanza-Riachuelo River was evaluated at both local and reach scales: this was done by deploying and monitoring an in situ pilot SEPA station, and by performing numerical modeling for the evaluation of the hydrodynamics in the SEPA station and the water quality in the reach where SEPA stations are planned to be implemented. An efficiency of aeration of 99% was estimated from DO measurements in the pilot SEPA, showing the potential of this technology for DO recovery in urban streams. Three-dimensional hydrodynamic modeling, besides assisting in the design of the pilot SEPA, has allowed for designing a prototype SEPA to be built soon. Finally, one-dimensional water quality modeling has provided the

  1. Jules Horowitz Reactor: a high performance material testing reactor

    Science.gov (United States)

    Iracane, Daniel; Chaix, Pascal; Alamo, Ana

    2008-04-01

    The physical modelling of materials' behaviour under severe conditions is an indispensable element for developing future fission and fusion systems: screening, design, optimisation, processing, licensing, and lifetime assessment of a new generation of structure materials and fuels, which will withstand high fast neutron flux at high in-service temperatures with the production of elements like helium and hydrogen. JANNUS and other analytical experimental tools are developed for this objective. However, a purely analytical approach is not sufficient: there is a need for flexible experiments integrating higher scales and coupled phenomena and offering high quality measurements; these experiments are performed in material testing reactors (MTR). Moreover, complementary representative experiments are usually performed in prototypes or dedicated facilities such as IFMIF for fusion. Only such a consistent set of tools operating on a wide range of scales, can provide an actual prediction capability. A program such as the development of silicon carbide composites (600-1200 °C) illustrates this multiscale strategy. Facing the long term needs of experimental irradiations and the ageing of present MTRs, it was thought necessary to implement a new generation high performance MTR in Europe for supporting existing and future nuclear reactors. The Jules Horowitz Reactor (JHR) project copes with this context. It is funded by an international consortium and will start operation in 2014. JHR will provide improved performances such as high neutron flux ( 10 n/cm/s above 0.1 MeV) in representative environments (coolant, pressure, temperature) with online monitoring of experimental parameters (including stress and strain control). Experimental devices designing, such as high dpa and small thermal gradients experiments, is now a key objective requiring a broad collaboration to put together present scientific state of art, end-users requirements and advanced instrumentation. To cite this

  2. Three-dimensional fluid-structure interaction dynamics of a pool-reactor in-tank component. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Kulak, R.F.

    1979-01-01

    The safety evaluation of reactor-components often involves the analysis of various types of fluid/structural components interacting in three-dimensional space. For example, in the design of a pool-type reactor several vital in-tank components such as the primary pumps and the intermediate heat exchangers are contained within the primary tank. Typically, these components are suspended from the deck structure and largely submersed in the sodium pool. Because of this positioning these components are vulnerable to structural damage due to pressure wave propagation in the tank during a CDA. In order to assess the structural integrity of these components it is necessary to perform a dynamic analysis in three-dimensional space which accounts for the fluid-structure coupling. A model is developed which has many of the salient features of this fluid-structural component system.

  3. Advanced burner test reactor preconceptual design report.

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

    2008-12-16

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an

  4. Instrumentation to Enhance Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

    2009-09-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  5. Instrumentation to Enhance Advanced Test Reactor Irradiations

    International Nuclear Information System (INIS)

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  6. Internal fluid flow management analysis for Clinch River Breeder Reactor Plant sodium pumps

    International Nuclear Information System (INIS)

    The Clinch River Breeder Reactor Plant (CRBRP) sodium pumps are currently being designed and the prototype unit is being fabricated. In the design of these large-scale pumps for elevated temperature Liquid Metal Fast Breeder Reactor (LMFBR) service, one major design consideration is the response of the critical parts to severe thermal transients. A detailed internal fluid flow distribution analysis has been performed using a computer code HAFMAT, which solves a network of fluid flow paths. The results of the analytical approach are then compared to the test data obtained on a half-scale pump model which was tested in water. The details are presented of pump internal hydraulic analysis, and test and evaluation of the half-scale model test results

  7. Design of Seismic Test Rig for Control Rod Drive Mechanism of Jordan Research and Training Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Jongoh; Kim, Gyeongho; Yoo, Yeonsik; Cho, Yeonggarp; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The reactor assembly is submerged in a reactor pool filled with water and its reactivity is controlled by locations of four control absorber rods(CARs) inside the reactor assembly. Each CAR is driven by a stepping motor installed at the top of the reactor pool and they are connected to each other by a tie rod and an electromagnet. The CARs scram the reactor by de-energizing the electromagnet in the event of a safe shutdown earthquake(SSE). Therefore, the safety function of the control rod drive mechanism(CRDM) which consists of a drive assembly, tie rod and CARs is to drop the CAR into the core within an appropriate time in case of the SSE. As well known, the operability for complex equipment such as the CRDM during an earthquake is very hard to be demonstrated by analysis and should be verified through tests. One of them simulates the reactor assembly and the guide tube of the CAR, and the other one does the pool wall where the drive assembly is installed. In this paper, design of the latter test rig and how the test is performed are presented. Initial design of the seismic test rig and excitation table had its first natural frequency at 16.3Hz and could not represent the environment where the CRDM was installed. Therefore, experimental modal analyses were performed and an FE model for the test rig and table was obtained and tuned based on the experimental results. Using the FE model, the design of the test rig and table was modified in order to have higher natural frequency than the cutoff frequency. The goal was achieved by changing its center of gravity and the stiffness of its sliding bearings.

  8. Process, policy, and implementation of pool-wide drawdowns on the Upper Mississippi River: a promising approach for ecological restoration of large impounded rivers

    Science.gov (United States)

    Kenow, Kevin P.; Gretchen Benjamin,; Tim Schlagenhaft,; Ruth Nissen,; Mary Stefanski,; Gary Wege,; Scott A. Jutila,; Newton, Teresa J.

    2016-01-01

    The Upper Mississippi River (UMR) has been developed and subsequently managed for commercial navigation by the U.S. Army Corps of Engineers (USACE). The navigation pools created by a series of lock and dams initially provided a complex of aquatic habitats that supported a variety of fish and wildlife. However, biological productivity declined as the pools aged. The River Resources Forum, an advisory body to the St. Paul District of the USACE, established a multiagency Water Level Management Task Force (WLMTF) to evaluate the potential of water level management to improve ecological function and restore the distribution and abundance of fish and wildlife habitat. The WLMTF identified several water level management options and concluded that summer growing season drawdowns at the pool scale offered the greatest potential to provide habitat benefits over a large area. Here we summarize the process followed to plan and implement pool-wide drawdowns on the UMR, including involvement of stakeholders in decision making, addressing requirements to modify reservoir operating plans, development and evaluation of drawdown alternatives, pool selection, establishment of a monitoring plan, interagency coordination, and a public information campaign. Three pool-wide drawdowns were implemented within the St. Paul District and deemed successful in providing ecological benefits without adversely affecting commercial navigation and recreational use of the pools. Insights are provided based on more than 17 years of experience in planning and implementing drawdowns on the UMR. 

  9. A new impulse in the development of nuclear pool-type reactors for underground heating plant: Designing, running background and possible perspectives

    International Nuclear Information System (INIS)

    This paper considers the concept of energy supply with using ultimately safe pool-type integral nuclear reactors. Safety and reliability of these reactors has already been demonstrated to the public by the long-term operation of this type various research reactors. The reactor and power plant design features, new approach to the nuclear safety, the nuclear upgrading of existing energy system in a small Russian town are considered in the paper

  10. Assessment of the National Research Universal Reactor Proposed New Stack Sampling Probe Location for Compliance with ANSI/HPS N13.1-1999

    Energy Technology Data Exchange (ETDEWEB)

    Glissmeyer, John A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Antonio, Ernest J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Flaherty, Julia E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-02-29

    This document reports on a series of tests conducted to assess the proposed air sampling location for the National Research Universal reactor (NRU) complex exhaust stack, located in Chalk River, Ontario, Canada, with respect to the applicable criteria regarding the placement of an air sampling probe. Due to the age of the equipment in the existing monitoring system, and the increasing difficulty in acquiring replacement parts to maintain this equipment, a more up-to-date system is planned to replace the current effluent monitoring system, and a new monitoring location has been proposed. The new sampling probe should be located within the exhaust stack according to the criteria established by the American National Standards Institute/Health Physics Society (ANSI/HPS) N13.1-1999, Sampling and Monitoring Releases of Airborne Radioactive Substances from the Stack and Ducts of Nuclear Facilities. These criteria address the capability of the sampling probe to extract a sample that represents the effluent stream. The internal Pacific Northwest National Laboratory (PNNL) project for this task was 65167, Atomic Energy Canada Ltd. Chalk River Effluent Duct Flow Qualification. The testing described in this document was guided by the Test Plan: Testing of the NRU Stack Air Sampling Position (TP-STMON-032).

  11. TRIGA reactor dynamics: Frequency response tests

    International Nuclear Information System (INIS)

    In this work, the results of frequency response tests conducted on ITU TRIGA Reactor are presented. To conduct the experiments, a special 'micro control rod' and its submersible stepping-motor drive mechanism was designed and constructed. The experiments cover a frequency range of 0.002 - 2 Hz., and 0.02, 4, 200 kW nominal power levels. Zero-power and at-power reactivity to % power transfer functions are presented as gain, and phase shift vs. frequency diagrams. Low power response is in close agreement with the point reactor zero-power transfer function. Response at 200 kW is studied with the help of a Nyquist diagram, and found to be stable. An elaboration on the main features of the feedback mechanism is also given. Power to reactivity feedback was measured to be just about 1.5 cent / % power change. (authors)

  12. Decommissioning of the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    E. Perry; J. Chrzanowski; C. Gentile; R. Parsells; K. Rule; R. Strykowsky; M. Viola

    2003-10-28

    The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D&D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D&D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D&D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget.

  13. Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D and D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D and D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D and D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget

  14. C Reactor overbore test facility review

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, P.A.; Nilson, R.

    1964-04-24

    In 1961, large-size, smooth-bore, Zircaloy process tubes were installed in C-Reactor graphite channels that had been enlarged to 2.275 inches. These tubes were installed to provide a test and demonstration facility for the concept of overboring as a means of securing significant improvement in the production capability of the reactors, After two years of facility operation, it is now appropriate to consider the extent to which original objectives have been achieved, to re-examine the original objectives, and to consider the best future use of this unique facility. This report presents the general results of such a review and re-examination in more detail.

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

    International Nuclear Information System (INIS)

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

  16. The terrestrial carbon inventory on the Savannah River Site: Assessing the change in Carbon pools 1951-2001.

    Energy Technology Data Exchange (ETDEWEB)

    Dai, Zhaohua; Trettin, Carl, C.; Parresol, Bernard, R.

    2011-11-30

    The Savannah River Site (SRS) has changed from an agricultural-woodland landscape in 1951 to a forested landscape during that latter half of the twentieth century. The corresponding change in carbon (C) pools associated land use on the SRS was estimated using comprehensive inventories from 1951 and 2001 in conjunction with operational forest management and monitoring data from the site.

  17. Spatial and Temporal Patterns of Nitrification Rates in Forested Floodplain Wetland Soils of Upper Mississippi River Pool 8, Journal Article

    Science.gov (United States)

    Overbank flooding is thought to be a critical process controlling nitrogen retention and cycling. In this study we investigated the effects of season and flood frequency on soil nitrification rates at ten sites in forested floodplains of Upper Mississippi River, Pool 8...A rough ...

  18. Parametric evaluation of mixed (low and high enriched) fuel core for a swimming pool type research reactor

    International Nuclear Information System (INIS)

    A study has been carried out to evaluate the performance of a swimming pool type research reactor core comprised of mixed (low and high enriched) uranium fuel. The study includes the calculations of core reactivity, worth of control rods and core criticality at the Beginning Of Life (BOL) of the core and for two operating conditions Cold Zero Power (CZP) and Hot Full Power (HFP). Further, to ensure safe and stable operation of the core from nuclear design point of view, average power densities in the fuel region, power peaking factors, axial power distribution in the hot channel and reactivity feed back coefficients have also been calculated. Two group fluxes have also been determined at different irradiation locations. All these calculations have been carried out employing reactor lattice code WIMS-D14 and reactor analysis code CITATION The calculated results show reasonably good agreement with the quoted operational data of the previous LEU cores. (Author)

  19. 10 CFR Appendix P to Subpart B of... - Uniform Test Method for Measuring the Energy Consumption of Pool Heaters

    Science.gov (United States)

    2010-01-01

    ... of Pool Heaters P Appendix P to Subpart B of Part 430 Energy DEPARTMENT OF ENERGY ENERGY CONSERVATION... Subpart B of Part 430—Uniform Test Method for Measuring the Energy Consumption of Pool Heaters 1. Test method. The test method for testing pool heaters is as specified in American National Standards...

  20. An automated test facility for neutronic amplifiers

    International Nuclear Information System (INIS)

    Neutronic amplifiers are used at the Chalk River Laboratory in applications such as neutron flux monitoring and reactor control systems. Routine preventive maintenance of control and safety systems included annual calibration and characterization of the neutronic amplifiers. An investigation into the traditional methods of annual routine maintenance of amplifiers concluded that frequency and phase response measurements in particular were labour intensive and subject to non-repeatable errors. A decision was made to upgrade testing methods and facilities by using programmable test equipment under the control of a computer. In order to verify the results of the routine measurements, expressions for the transfer functions were derived from the circuit diagrams. Frequency and phase responses were then calculated and plotted thus providing a bench-mark to which the test results can be compared. (author)

  1. Thermal Hydraulic Tests for Reactor Core Safety

    Energy Technology Data Exchange (ETDEWEB)

    Moon, S. K.; Baek, W. P.; Chun, S. Y. (and others)

    2007-06-15

    The main objectives of the present project are to resolve the current issues of reactor core thermal hydraulics, to develop an advanced measurement and analytical techniques, and to perform reactor core safety verification tests. 6x6 reflood experiments, various heat transfer experiments using Freon, and experiments on the spacer grids effects on the post-dryout are carried out using spacer grids developed in Korea in order to resolve the current issues of the reactor core thermal hydraulics. In order to develop a reflood heat transfer model, the detailed reflood phenomena are visualized and measured using round tube and 2x2 rod bundle. A detailed turbulent mixing phenomenon for subchannels is measured using advanced measurement techniques such as LDV and PIV. MARS and MATRA codes developed in Korea are assessed, verified and improved using the obtained experimental data. Finally, a systematic quality assurance program and experimental data generation system has been constructed in order to increase the reliability of the experimental data.

  2. 33 CFR 207.170 - Federal Dam, Oklawaha River, Moss Bluff, Fla.; pool level.

    Science.gov (United States)

    2010-07-01

    ... Bluff, Fla.; pool level. 207.170 Section 207.170 Navigation and Navigable Waters CORPS OF ENGINEERS..., Moss Bluff, Fla.; pool level. (a) The level of the pool shall normally be maintained at elevation 56.5 feet above sea level: Provided, That the level of the pool may be raised to not exceeding 58.5...

  3. 33 CFR 207.60 - Federal Dam, Hudson River, Troy, N.Y.; pool level.

    Science.gov (United States)

    2010-07-01

    ..., N.Y.; pool level. 207.60 Section 207.60 Navigation and Navigable Waters CORPS OF ENGINEERS..., N.Y.; pool level. (a) Whenever the elevation of the pool created by the Federal dam at Troy, N.Y... automatically when the pool level rises to an elevation of +18.5 feet mean sea level, and conform in...

  4. Fuel irradiation test plan at the Japan materials testing reactor

    International Nuclear Information System (INIS)

    Development of high performance fuels, which enables burnup extension and high duty uses of light water reactors (LWRs) by means of power up rates and flexible operating cycles, is one of key technical issues for extending the uses for longer periods. Introduction of new design fuel rods with new cladding alloys and wider utilization of mixed oxide fuels is expected in Japan. Fuel irradiation tests for development and safety demonstration are quite important, in order to realize theses progress. Operational management on water chemistry, minimizing the long term degradation of reactor components, could have unfavorable influence on the integrity of the fuel rods. Japanese government and the Japan Atomic Energy Agency have decided to re new the Japan Materials Testing Reactor (JMTR) and to install new test rigs, in order to play an active role solving the issues on the development and the safety of the fuel and the plant aging. Fuel integrity under abnormal transient conditions will be investigated using a special capsule type test rig, which has its own power control system under simulated LWR cooling conditions. Water loops for simulation of high duty operation, e.g. high power, high burnup and high rod internal pressure conditions, are proposed for the development and safety examination of the high performance fuels. Combination of the JMTR tests with simulated reactivity initiated accident tests in the Nuclear Safety Research Reactor and loss of coolant accident tests in hot laboratories would provide a comprehensive data for safety evaluation and design progress of the high performance fuels at extended burnups, covering from the normal to the accident conditions, including abnormal transients

  5. Melt-dilute treatment of spent nuclear fuel assemblies from research and test reactors

    International Nuclear Information System (INIS)

    The Savannah River Site is the U.S. Department of Energy's preferred site for return and treatment of all aluminum-base, spent, research and test reactor fuel assemblies. There are over 20,000 spent fuel assemblies now stored in different countries around the world, and by 2035 many will be returned to SRS for treatment and interim storage, in preparation for disposal in a geologic repository. The early fuel assemblies for research and test reactors were made using aluminum clad plates that were fabricated from highly enriched (93%) uranium-aluminum alloy. Later, powder metallurgical fabrication methods were developed to produce plate fuels with higher uranium contents using either uranium aluminide, uranium oxide or uranium silicide powders mixed with aluminum. Silicide fuel elements generally are fabricated with low enriched uranium containing less than 20% 2'35U. Following irradiation, the spent fuel assemblies are discharged from the reactor, and most assemblies have been stored in underwater pools, some since the early 1950's. A number of disposition options including direct/co-disposal and melt-dilute treatment were evaluated recently. The melt-dilute technique was identified as the preferred method for treatment of aluminum-base spent fuel. The technique consists of melting the spent fuel assembly and adding depleted uranium to the melt for isotopic dilution to 2'35U. Aluminum is added, if necessary, to produce a predetermined alloy composition. Additionally, neutron poisons may be added to the melt where they form solid solution phases or compounds with uranium and/or aluminum. Lowering the enrichment reduces both criticality and proliferation concerns for storage. Consolidation by melting also reduces the number of storage canisters. Laboratory and small-scale process demonstration using irradiated fuel is underway. Tests of the off gas absorption system have been initiated using both surrogate and irradiated RERTR mini fuel plates. An experimental L

  6. Successful scaling-up of self-sustained pyrolysis of oil palm biomass under pool-type reactor.

    Science.gov (United States)

    Idris, Juferi; Shirai, Yoshihito; Andou, Yoshito; Mohd Ali, Ahmad Amiruddin; Othman, Mohd Ridzuan; Ibrahim, Izzudin; Yamamoto, Akio; Yasuda, Nobuhiko; Hassan, Mohd Ali

    2016-02-01

    An appropriate technology for waste utilisation, especially for a large amount of abundant pressed-shredded oil palm empty fruit bunch (OFEFB), is important for the oil palm industry. Self-sustained pyrolysis, whereby oil palm biomass was combusted by itself to provide the heat for pyrolysis without an electrical heater, is more preferable owing to its simplicity, ease of operation and low energy requirement. In this study, biochar production under self-sustained pyrolysis of oil palm biomass in the form of oil palm empty fruit bunch was tested in a 3-t large-scale pool-type reactor. During the pyrolysis process, the biomass was loaded layer by layer when the smoke appeared on the top, to minimise the entrance of oxygen. This method had significantly increased the yield of biochar. In our previous report, we have tested on a 30-kg pilot-scale capacity under self-sustained pyrolysis and found that the higher heating value (HHV) obtained was 22.6-24.7 MJ kg(-1) with a 23.5%-25.0% yield. In this scaled-up study, a 3-t large-scale procedure produced HHV of 22.0-24.3 MJ kg(-1) with a 30%-34% yield based on a wet-weight basis. The maximum self-sustained pyrolysis temperature for the large-scale procedure can reach between 600 °C and 700 °C. We concluded that large-scale biochar production under self-sustained pyrolysis was successfully conducted owing to the comparable biochar produced, compared with medium-scale and other studies with an electrical heating element, making it an appropriate technology for waste utilisation, particularly for the oil palm industry.

  7. Successful scaling-up of self-sustained pyrolysis of oil palm biomass under pool-type reactor.

    Science.gov (United States)

    Idris, Juferi; Shirai, Yoshihito; Andou, Yoshito; Mohd Ali, Ahmad Amiruddin; Othman, Mohd Ridzuan; Ibrahim, Izzudin; Yamamoto, Akio; Yasuda, Nobuhiko; Hassan, Mohd Ali

    2016-02-01

    An appropriate technology for waste utilisation, especially for a large amount of abundant pressed-shredded oil palm empty fruit bunch (OFEFB), is important for the oil palm industry. Self-sustained pyrolysis, whereby oil palm biomass was combusted by itself to provide the heat for pyrolysis without an electrical heater, is more preferable owing to its simplicity, ease of operation and low energy requirement. In this study, biochar production under self-sustained pyrolysis of oil palm biomass in the form of oil palm empty fruit bunch was tested in a 3-t large-scale pool-type reactor. During the pyrolysis process, the biomass was loaded layer by layer when the smoke appeared on the top, to minimise the entrance of oxygen. This method had significantly increased the yield of biochar. In our previous report, we have tested on a 30-kg pilot-scale capacity under self-sustained pyrolysis and found that the higher heating value (HHV) obtained was 22.6-24.7 MJ kg(-1) with a 23.5%-25.0% yield. In this scaled-up study, a 3-t large-scale procedure produced HHV of 22.0-24.3 MJ kg(-1) with a 30%-34% yield based on a wet-weight basis. The maximum self-sustained pyrolysis temperature for the large-scale procedure can reach between 600 °C and 700 °C. We concluded that large-scale biochar production under self-sustained pyrolysis was successfully conducted owing to the comparable biochar produced, compared with medium-scale and other studies with an electrical heating element, making it an appropriate technology for waste utilisation, particularly for the oil palm industry. PMID:26612557

  8. Analytical Study of High Concentration PCB Paint at the Heavy Water Components Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lowry, N.J.

    1998-10-21

    This report provides results of an analytical study of high concentration PCB paint in a shutdown nuclear test reactor located at the US Department of Energy's Savannah River Site (SRS). The study was designed to obtain data relevant for an evaluation of potential hazards associated with the use of and exposure to such paints.

  9. Analytical Study of High Concentration PCB Paint at the Heavy Water Components Test Reactor

    International Nuclear Information System (INIS)

    This report provides results of an analytical study of high concentration PCB paint in a shutdown nuclear test reactor located at the US Department of Energy's Savannah River Site (SRS). The study was designed to obtain data relevant for an evaluation of potential hazards associated with the use of and exposure to such paints

  10. Distribution and origin of suspended matter and organic carbon pools in the Tana River Basin, Kenya

    Directory of Open Access Journals (Sweden)

    F. Tamooh

    2012-08-01

    Full Text Available We studied patterns in organic carbon pools and their origin in the Tana River Basin (Kenya, in February 2008 (dry season, September–November 2009 (wet season, and June–July 2010 (end of wet season, covering the full continuum from headwater streams to lowland mainstream sites. A consistent downstream increase in total suspended matter (TSM, 0.6 to 7058 mg l−1 and particulate organic carbon (POC, 0.23 to 119.8 mg l−1 was observed during all three sampling campaigns, particularly pronounced below 1000 m above sea level, indicating that most particulate matter exported towards the coastal zone originated from the mid and low altitude zones rather than from headwater regions. This indicates that the cascade of hydroelectrical reservoirs act as an extremely efficient particle trap. Although 7Be / 210Pbxs ratios/age of suspended sediment do not show clear seasonal variation, the gradual downstream increase of suspended matter during end of wet season suggests its origin is caused by inputs of older sediments from bank erosion and/or river sediment resuspension. During wet season, higher TSM concentrations correspond with relatively young suspended matter, suggesting a contribution from recently eroded material. With the exception of reservoir waters, POC was predominantly of terrestrial origin as indicated by generally high POC : chlorophyll a (POC : Chl a ratios (up to ~41 000. Stable isotope signatures of POC (δ13CPOC ranged between −32 and −20‰ and increased downstream, reflecting an increasing contribution of C4-derived carbon in combination with an expected shift in δ13C for C3 vegetation towards the more semi-arid lowlands. δ13C values in sediments from the main reservoir (−19.5 to −15.7‰ were higher than those found in any of the riverine samples, indicating selective retention of particles associated with C4

  11. Fast Shutdown System tests in the Georgia Tech Research Reactor

    International Nuclear Information System (INIS)

    The Fast Shutdown System (FSS) is a new safety system design concept being considered for in installation in the Savannah River (SRS) production reactors. This system is expected to mitigate the consequences of a Design Basis Loss of Coolant Accident, and therefore allow higher operational power levels. A test of this system in the Georgia Tech Research Reactor is proposed to demonstrate the efficacy of this concept. Three tests will be conducted at full power (5MW) and one at low power (100kw). Two full power tests will be conducted with the FSS rod backfilled with one (1) atmosphere of He-4, and one with the rod evacuated. The low power conducted with the FSS rod evacuated. Neutron flux and pressure data will be collected with an independent data acquisition system (DAS). Safety issues associated with the performance of the Fast Shutdown System experiments are addressed in this report. The credible accident scenarios were analyzed using worst case scenarios to demonstrate that no significant nuclear or personnel safety hazards would result from the performance of the proposed experiments

  12. Mark I 1/12-scale pressure suppression pool swell tests

    International Nuclear Information System (INIS)

    A second series of 1/12-scale tests of the Mark I BWR containment have been run to define the torus and ring header loads resulting from pressure suppression pool swell following a postulated LOCA. These tests were conducted following modification of the test section used for earlier tests to tighten the sealing and improve control of the test conditions. Forty-two tests were performed, investigating the sensitivity of the pool response to drywell pressurization rate, initial downcomer submergence, initial system pressure and initial drywell/wetwell pressure differential, covering the range of scaled Mark I expected conditions

  13. Mark I 1/12-scale pressure suppression pool swell tests. [BWR

    Energy Technology Data Exchange (ETDEWEB)

    Torbeck, J.E.; Galyardt, D.L.; Walker, J.P.

    1976-05-01

    A second series of 1/12-scale tests of the Mark I BWR containment have been run to define the torus and ring header loads resulting from pressure suppression pool swell following a postulated LOCA. These tests were conducted following modification of the test section used for earlier tests to tighten the sealing and improve control of the test conditions. Forty-two tests were performed, investigating the sensitivity of the pool response to drywell pressurization rate, initial downcomer submergence, initial system pressure and initial drywell/wetwell pressure differential, covering the range of scaled Mark I expected conditions (on the basis of the FSAR).

  14. SRS reactor stack plume marking tests

    Energy Technology Data Exchange (ETDEWEB)

    Petry, S.F.

    1992-03-01

    Tests performed in 105-K in 1987 and 1988 demonstrated that the stack plume can successfully be made visible (i.e., marked) by introducing smoke into the stack breech. The ultimate objective of these tests is to provide a means during an emergency evacuation so that an evacuee can readily identify the stack plume and evacuate in the opposite direction, thus minimizing the potential of severe radiation exposure. The EPA has also requested DOE to arrange for more tests to settle a technical question involving the correct calculation of stack downwash. New test canisters were received in 1988 designed to produce more smoke per unit time; however, these canisters have not been evaluated, because normal ventilation conditions have not been reestablished in K Area. Meanwhile, both the authorization and procedure to conduct the tests have expired. The tests can be performed during normal reactor operation. It is recommended that appropriate authorization and procedure approval be obtained to resume testing after K Area restart.

  15. Dielectric Heaters for Testing Spacecraft Nuclear Reactors

    Science.gov (United States)

    Sims, William Herbert; Bitteker, Leo; Godfroy, Thomas

    2006-01-01

    A document proposes the development of radio-frequency-(RF)-driven dielectric heaters for non-nuclear thermal testing of the cores of nuclear-fission reactors for spacecraft. Like the electrical-resistance heaters used heretofore for such testing, the dielectric heaters would be inserted in the reactors in place of nuclear fuel rods. A typical heater according to the proposal would consist of a rod of lossy dielectric material sized and shaped like a fuel rod and containing an electrically conductive rod along its center line. Exploiting the dielectric loss mechanism that is usually considered a nuisance in other applications, an RF signal, typically at a frequency .50 MHz and an amplitude between 2 and 5 kV, would be applied to the central conductor to heat the dielectric material. The main advantage of the proposal is that the wiring needed for the RF dielectric heating would be simpler and easier to fabricate than is the wiring needed for resistance heating. In some applications, it might be possible to eliminate all heater wiring and, instead, beam the RF heating power into the dielectric rods from external antennas.

  16. Impact of supercritical CO2 injection on petrophysical and rock mechanics properties of chalk: an experimental study on chalk from South Arne field, North Sea

    DEFF Research Database (Denmark)

    Alam, Mohammad Monzurul; Hjuler, Morten Leth; Christensen, Helle Foged;

    2011-01-01

    Changes in chalk due to EOR by injecting supercritical CO2 (CO2-EOR) can ideally be predicted by applying geophysical methods designed from laboratory-determined petrophysical and rock mechanics properties. A series of petrophysical and rock mechanics tests were performed on Ekofisk Formation...... and Tor Formation chalk of the South Arne field to reveal the changes in petrophysical and rock mechanics properties of chalk due to the injection of CO2 at supercritical state. An increase in porosity and decrease in specific surface was observed due to injection of supercritical CO2. This indicates...... as indicated by NMR T2 relaxation time was observed. Rock mechanics testing indicates that in 30% porosity chalk from the South Arne field, injection of supercritical CO2 has no significant effect on shear strength and compaction properties, while there is probably a slight decrease in stiffness properties...

  17. Numerical modelling of low-Reynolds number direct contact condensation in a suppression pool test facility

    International Nuclear Information System (INIS)

    Highlights: • A low-Reynolds number direct contact condensation mode was simulated. • Eulerian two-fluid approach was used without interfacial tracking. • The numerical results were validated with the steam blowdown test. • The surface divergence model predicted the condensation phenomena satisfactory. - Abstract: In the safety pressure suppression pool systems of Boiling Water Reactors (BWRs), the condensation rate has to be maintained high enough in order to fulfill their safety function. A major part of this condensation occurs as direct contact condensation (DCC), which governs different modes varying from vigorous chugging of collapsing bubbles to mild condensation on almost flat steam–water interface. This paper discusses the Computational Fluid Dynamics (CFD) simulations of the latter, low-Reynolds number weak condensation regime. The numerical simulations were performed with two CFD codes, NEPTUNECFD and OpenFOAM, in which the DCC phenomenon was modelled by using the Eulerian two-fluid approach of interpenetrating continua without interfacial tracking. The interfacial heat transfer between steam and water was modelled by using the DCC models based on the surface renewal and the surface divergence theories. Flow turbulence was solved by employing the standard k–∊ turbulence model. The CFD results of this study were validated against the test results of the POOLEX facility of Lappeenranta University of Technology. In the reference test STB-31, the condensation phenomena were limited to only occur on a stable steam–water interface by very low steam mass flux applied and thermal insulation of the blowdown pipe. The simulation results demonstrated that the surface divergence model predicted the condensation phenomena quite accurately both qualitatively and quantitatively while the surface renewal model overestimated it strongly

  18. Alkali metal pool boiler life tests for a 25 kWe advanced Stirling conversion system

    Science.gov (United States)

    Anderson, W. G.; Rosenfeld, J. H.; Noble, J.

    The overall operating temperature and efficiency of solar-powered Stirling engines can be improved by adding an alkali metal pool boiler heat transport system to supply heat more uniformly to the heater head tubes. One issue with liquid metal pool boilers is unstable boiling. Stable boiling is obtained with an enhanced boiling surface containing nucleation sites that promote continuous boiling. Over longer time periods, it is possible that the boiling behavior of the system will change. An 800-h life test was conducted to verify that pool boiling with the chosen fluid/surface combination remains stable as the system ages. The apparatus uses NaK boiling on a - 100 + 140 stainless steel sintered porous layer, with the addition of a small amount of xenon. Pool boiling remained stable to the end of life test. The pool boiler life test included a total of 82 cold starts, to simulate startup each morning, and 60 warm restarts, to simulate cloud cover transients. The behavior of the cold and warm starts showed no significant changes during the life test. In the experiments, the fluid/surface combination provided stable, high-performance boiling at the operating temperature of 700 C. Based on these experiments, a pool boiler was designed for a full-scale 25-kWe Stirling system.

  19. Composition of the seed bank in drawdown areas of navigation pool 8 of the upper Mississippi river

    Science.gov (United States)

    Kenow, K.P.; Lyon, J.E.

    2009-01-01

    In an effort to enhance aquatic plant production and habitat diversity on the Upper Mississippi River (UMR), resource managers considered water level reduction as a management tool to increase the area of emergent and submersed aquatic vegetation by natural seed germination. To quantify the availability of seed, we assessed the potential seed bank of selected areas of Navigation Pool 8 of the UMR from substrate samples collected in spring 2000. We tested these samples for viable seed content under four hydrologic conditions: dry, moist, shallow flooded and submerged. Forty-seven species were identified in the seed bank, including 27 obligate wetland, 10 facultative wetland and 7 upland species. Dominant taxa within the seed bank included Sagittaria spp., Lindernia dubia, Zosterella dubia, Cyperus spp., Eragrostis spp. and Leersia oryzoides. Of the four hydrologic treatments, moist substrates had the greatest species diversity and were the most productive, yielding an average density of 1420 seedlings m-2. Emergent and submersed aquatic species were widely distributed, each type occurring in more than 90% of the samples. Timing of seedling germination varied among species and has implications for scheduling drawdowns to promote establishment of desired species. Seed bank results were correlated with the vegetation response on substrates exposed during a reduction of water levels of Pool 8 during summer 2001. Experimentally determining the composition and viability of seed banks from drawdown areas provides information useful in predicting the types of vegetation that may develop on exposed substrates. Further, these findings provide resource managers a better understanding of the potential for achieving desired vegetation response through water level reductions.

  20. Evaluation of LOCA in a swimming-pool type reactor using the 3D-AIRLOCA code

    International Nuclear Information System (INIS)

    The 3D-AIRLOCA code was used to calculate core temperature evolution curves in the wake of a full LOCA in a swimming pool type reactor, resulting in complete core exposure and dryout within about 1000 sec of the initiating event. The results show that fuel integrity loss thresholds (450 C for softening and 650 C for melting) are reached and exceeded over large fractions of the core at powr levels as low as 2 MW. At 4.5 MW, the softening threshold is reached even when the accident occurs up to 12 hours after reactor shutdown for continuous operation, and up to 2 hrs after shutdown for intermittent (6 hrs/day, 4 days a week) operation. The situation is even more severe in blockage cases, when the air flow through the core is blocked by residual water at the grid plate level. It is concluded that substantial fission product releases are quite likely in this class of accidents. (orig.)

  1. TREAT [Transient Reactor Test Facility] reactor control rod scram system simulations and testing

    International Nuclear Information System (INIS)

    Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent ampersand Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs

  2. Corrosion of spent Advanced Test Reactor fuel

    International Nuclear Information System (INIS)

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

  3. Proposal of world network on material testing reactors

    International Nuclear Information System (INIS)

    Establishment of an international cooperation system of worldwide testing reactor network (world network) is proposed in order to achieve efficient facility utilization and provide high quality irradiation data by role sharing of irradiation tests with materials testing reactors in the world. As for the first step, mutual understanding among materials testing reactors is thought to be necessary. From this point, an international symposium on materials testing reactors (ISMTR) was held to construct the world network from 2008, and a common understanding of world network has begun to be shared. (author)

  4. High Temperature Gas-Cooled Test Reactor Options Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    Preliminary scoping calculations are being performed for a 100 MWt gas-cooled test reactor. The initial design uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to identify some reactor design features to investigate further. Current status of the effort is described.

  5. Continuous-flow stirred-tank reactor 20-L demonstration test: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D.D.; Collins, J.L.

    2000-02-01

    One of the proposed methods of removing the cesium, strontium, and transuranics from the radioactive waste storage tanks at Savannah River is the small-tank tetraphenylborate (TPB) precipitation process. A two-reactor-in-series (15-L working volume each) continuous-flow stirred-tank reactor (CSTR) system was designed, constructed, and installed in a hot cell to test the Savannah River process. The system also includes two cross-flow filtration systems to concentrate and wash the slurry produced in the process, which contains the bulk of radioactivity from the supernatant processed through the system. Installation, operational readiness reviews, and system preparation and testing were completed. The first test using the filtration systems, two CSTRs, and the slurry concentration system was conducted over a 61-h period with design removal of Cs, Sr, and U achieved. With the successful completion of Test 1a, the following tests, 1b and 1c, were not required.

  6. A sipping test simulator for identifying defective fuels in MTR type nuclear research reactor

    International Nuclear Information System (INIS)

    Highlights: • This simulator based on windows application of C# programming language. • This simulator could be useful for training of technicians in spent nuclear fuels storage facility. • This simulator is user friendly and easy to learn. - Abstract: Integrity of fuel assemblies is critical to continuous operation of any nuclear reactor. NDT methods and sipping test are practical techniques which are used for this purpose. Assessing the fuel integrity by NDT is a troublesome process which could incur personal overdose due to high radiation, requiring large space, and heavy equipment. Therefore to overcome problems associated with the NDT process, sipping test is widely used. The main purpose of this article is introducing sipping test simulator (STS) which is so important for training. Also, this article describes the procedure and methodology used to perform sipping test on the fuel assemblies either in reactor pool or spent fuel storage pool. A unique ability of this simulator is analyzing direct spectroscopy files from experimental data of a real operating reactor. The sipping test simulator is a full-feature training curriculum in spent nuclear fuels storage technology with a PC-based simulator. This simulator is written in C# programming language for a Windows based computer. The simulator will teach everything needed to know for identifying the fuel defects using sipping test process. As learning the basics of sipping test step wise, a freshman operator will soon be able to accomplish all steps in practice

  7. The Live program - Results of test L1 and joint analyses on transient molten pool thermal hydraulics

    International Nuclear Information System (INIS)

    The development of a corium pool in the lower head and its behaviour is still a critical issue. This concerns, in general, the understanding of a severe accident with core melting, its course, major critical phases and timing, and the influence of these processes on the accident progression as well as, in particular, the evaluation of in-vessel melt retention by external vessel flooding as an accident mitigation strategy. Previous studies were especially related to the in-vessel retention question and often just concentrated on the quasi-steady state behaviour of a large molten pool in the lower head, considered as a bounding configuration. However, non-feasibility of the in-vessel retention concept for high power density reactors and uncertainties e. g. due to layering effects even for low or medium power reactors, turns this to be insufficient. Rather, it is essential to consider the whole evolution of the accident, including e. g. formation and growth of the in-core melt pool, characteristics of corium arrival in the lower head, and molten pool behaviour after the debris re-melting. These phenomena have a strong impact on a potential termination of a severe accident. The general objective of the LIVE program at FZK is to study these phenomena resulting from core melting experimentally in large-scale 3D geometry and in supporting separate-effects tests, with emphasis on the transient behaviour. Up to now, several tests on molten pool behaviour have been performed within the LIVE experimental program with water and with non-eutectic melts (KNO3-NaNO3) as simulant fluids. The results of these experiments, performed in nearly adiabatic and in isothermal conditions, allow a direct comparison with findings obtained earlier in other experimental programs (SIMECO, ACOPO, BALI, etc. ) and will be used for the assessment of the correlations derived for the molten pool behaviour. Complementary to other international programs with real corium melts, the results of the LIVE

  8. Radiation exposure: Cytogenetic tests. Chernobyl reactor accident

    International Nuclear Information System (INIS)

    Forty test subjects who, either during or after the reactor accident of Chernobyl (26th April 1986), stayed at a building site at Shlobin 150 km away, were examined for spontaneously occurring as well as mitomycin C-induced Sister Chromatid Exchanges (SCE). The building site staff, who underwent a whole-body radionuclide count upon their return to Austria (June through September 1986), were used for the cytogenetic tests. The demonstration of the SCE was made from whole-blood cultures by the fluorescence/Giemse technique. At last 20 Metaphases of the 2nd mitotic cycle were evaluated per person. The radiation doses of the test subjects were calculated by adding the external exposure determined on the building site, the estimated thyroid dose through I-131, and the measured incorporation of Cs-134 and Cs-137. The subjects were divided into two groups for statistical analysis: One was a more exposed group (proven stay at Shlobin between 26th April and 31st May 1986, mostly working in the open air) and the other a less exposed group for comparison (staying at Shlobin from 1st Juni 1986 and working mainly indoors). (orig.)

  9. Similarity Analysis for Reactor Flow Distribution Test and Its Validation

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Soon Joon; Ha, Jung Hui [Heungdeok IT Valley, Yongin (Korea, Republic of); Lee, Taehoo; Han, Ji Woong [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    The newly derived dimensionless groups are slightly different from Hetsroni's. Reynolds number, relative wall roughness, and Euler don't appear, instead, friction factor appears newly. In order to conserve friction factor Reynolds number and relative wall roughness should be conserved. Since the effect of Reynolds number in high range is small, and since the scaled model is far smaller than prototype the conservation of friction factor is easily obtained by making the model wall just smooth. It is much easier to implement the test design than Hetsroni's because the Reynolds number and relative wall roughness do not appear explicitly. In case that there is no free surface within the interested domain of the reactor, the gravity is of second importance, and in this case the pressure drops should be compensated for in order to compare them between prototype and model. The gravity head compensated pressure drop is directly same to the measured value by a differential pressure transmitter. In order to conserve the gravity effect Froude number should be conserved. In pool type SFR (Sodium Cooled Fast Reactor) there exists liquid level difference, and if the level difference is desired to be conserved, the Froude number should be conserved. Euler number, which represents pressure terms in momentum equation, should be well conserved according to Hetsroni's approach. It is not a wrong statement, but it should be noted that Euler number is NOT an independent variable BUT a dependent variable according to Hong et al. It means that if all the geometrical similarity and the dimensionless numbers are conserved, Euler number is automatically conserved. So Euler number need not be considered in case that the perfect geometrical similarity is kept. However, even in case that the geometrical similarity is not conserved, it possible to conserved the velocity field similarity by just conserve Euler number. It gives tolerance to the engineer who designs the test

  10. Similarity Analysis for Reactor Flow Distribution Test and Its Validation

    International Nuclear Information System (INIS)

    The newly derived dimensionless groups are slightly different from Hetsroni's. Reynolds number, relative wall roughness, and Euler don't appear, instead, friction factor appears newly. In order to conserve friction factor Reynolds number and relative wall roughness should be conserved. Since the effect of Reynolds number in high range is small, and since the scaled model is far smaller than prototype the conservation of friction factor is easily obtained by making the model wall just smooth. It is much easier to implement the test design than Hetsroni's because the Reynolds number and relative wall roughness do not appear explicitly. In case that there is no free surface within the interested domain of the reactor, the gravity is of second importance, and in this case the pressure drops should be compensated for in order to compare them between prototype and model. The gravity head compensated pressure drop is directly same to the measured value by a differential pressure transmitter. In order to conserve the gravity effect Froude number should be conserved. In pool type SFR (Sodium Cooled Fast Reactor) there exists liquid level difference, and if the level difference is desired to be conserved, the Froude number should be conserved. Euler number, which represents pressure terms in momentum equation, should be well conserved according to Hetsroni's approach. It is not a wrong statement, but it should be noted that Euler number is NOT an independent variable BUT a dependent variable according to Hong et al. It means that if all the geometrical similarity and the dimensionless numbers are conserved, Euler number is automatically conserved. So Euler number need not be considered in case that the perfect geometrical similarity is kept. However, even in case that the geometrical similarity is not conserved, it possible to conserved the velocity field similarity by just conserve Euler number. It gives tolerance to the engineer who designs the test

  11. Welding of stainless steel pool of pressurized water reactor nuclear power station

    International Nuclear Information System (INIS)

    The construction of stainless steel lining of million kilowatt grade pressurized water reactor nuclear power station is a new technology. The author introduces its welding method, parameter verification measure and key factors of construction quality control and so on

  12. Distribution and origin of suspended sediments and organic carbon pools in the Tana River Basin, Kenya

    Directory of Open Access Journals (Sweden)

    F. Tamooh

    2012-03-01

    Full Text Available We studied patterns in organic carbon pools and their origin in the Tana River Basin (Kenya, in February 2008 (dry season, September–November 2009 (wet season, and June–July 2010 (end of wet season, and covering the full continuum from headwater streams to lowland mainstream sites. A consistent downstream increase in total suspended matter (TSM, 0.6 to 7058 mg l−1 and particulate organic carbon (POC, 0.23 to 119.8 mg l−1 was observed during all three sampling campaigns, particularly pronounced below 1000 m above sea level, indicating that most particulate matter exported towards the coastal zone originated from the mid and low altitude zones rather than from headwater regions. This indicates that the cascade of hydroelectrical reservoirs act as an extremely efficient particle trap. The decrease in 7Be/210Pbxs ratios of TSM downstream (range: 0.43 to 1.93 during the wet season indicated that the increasing sediment load in the lower Tana was largely due to recent surface erosion. During lower flow conditions, however, the gradual longitudinal increase in TSM coincided was more variable 7Be/210Pbxs ratios (0 to 4.5, suggesting that bank erosion and/or remobilisation of older sediments are the sources of the increasing TSM concentrations downstream. With the exception of reservoir waters, POC was predominantly of terrestrial origin as indicated by generally high POC/Chl-a ratios (up to ∼ 41 000. Stable isotope signatures of POC (δ13CPOC ranged between –32 and –20 ‰ and increased downstream, reflecting an increasing contribution of C4-derived carbon in combination with an expected shift in δ13C for C3 vegetation towards the more semi-arid lowlands. Sediments from the main reservoir (Masinga showed δ13C values higher (–19.5 to –15.7 ‰ than found in any of the riverine samples, indicating

  13. An evaluation of the relative quality of dike pools for benthic macroinvertebrates in the Lower Missouri River, USA

    Science.gov (United States)

    Poulton, B.C.; Allert, A.L.

    2012-01-01

    A habitat-based aquatic macroinvertebrate study was initiated in the Lower Missouri River to evaluate relative quality and biological condition of dike pool habitats. Water-quality and sediment-quality parameters and macroinvertebrate assemblage structure were measured from depositional substrates at 18 sites. Sediment porewater was analysed for ammonia, sulphide, pH and oxidation-reduction potential. Whole sediments were analysed for particle-size distribution, organic carbon and contaminants. Field water-quality parameters were measured at subsurface and at the sediment-water interface. Pool area adjacent and downstream from each dike was estimated from aerial photography. Macroinvertebrate biotic condition scores were determined by integrating the following indicator response metrics: % of Ephemeroptera (mayflies), % of Oligochaeta worms, Shannon Diversity Index and total taxa richness. Regression models were developed for predicting macroinvertebrate scores based on individual water-quality and sediment-quality variables and a water/sediment-quality score that integrated all variables. Macroinvertebrate scores generated significant determination coefficients with dike pool area (R2=0.56), oxidation–reduction potential (R2=0.81) and water/sediment-quality score (R2=0.71). Dissolved oxygen saturation, oxidation-reduction potential and total ammonia in sediment porewater were most important in explaining variation in macroinvertebrate scores. The best two-variable regression models included dike pool size + the water/sediment-quality score (R2=0.84) and dike pool size + oxidation-reduction potential (R2=0.93). Results indicate that dike pool size and chemistry of sediments and overlying water can be used to evaluate dike pool quality and identify environmental conditions necessary for optimizing diversity and productivity of important aquatic macroinvertebrates. A combination of these variables could be utilized for measuring the success of habitat enhancement

  14. An Evaluation of liquid metal leak detection methods for the Clinch River Breeder Reactor Plant

    Energy Technology Data Exchange (ETDEWEB)

    Morris, C.J.; Doctor, S.R.

    1977-12-01

    This report documents an independent review and evaluation of sodium leak detection methods described in the Clinch River Breeder Reactor Preliminary Safety Analysis Report. Only information in publicly available documents was used in making the assessments.

  15. Repair of the NRU Reactor Vessel: Technical Challenges and Lessons Learned

    International Nuclear Information System (INIS)

    Full text: In May 2009, following a Class 4 power outage that affected most of Eastern Ontario, including the Chalk River Laboratories site, Atomic Energy of Canada Limited (AECL) announced to its various stakeholders that a small heavy-water leak in the NRU reactor had been detected during routine monitoring while the reactor was being readied for return to service. Over the next 15 months AECL located, inspected, repaired and returned the NRU reactor to service. This presentation will focus on the extensive efforts required to support the unique activities associated with reactor vessel inspection and repair including initial assessment, repair site challenges, repair preparation and finally repair execution. The presentation will summarize: - Initial leak search and assessment of the vessel condition through the use of specialized tooling and non-destructive evaluation which resulted in one of the largest single NDE inspection campaigns ever carried out in the nuclear industry; - Challenges of executing a repair through 12 cm access ports at a distance of nine meters including the development of the specialized tooling; - The importance of development of repair techniques through mock up testing to perform welding repairs on a thin wall aluminium vessel and the measures taken and engineering challenges overcome to achieve a successful repair; - The final repair process, including site preparation, weld execution and final NDE inspection techniques; - Challenges encountered and lesson learned during the execution of weld repair, NDE inspections, and return-to-service of the reactor. (author)

  16. Pooled Nucleic Acid Testing to Detect Antiretroviral Treatment Failure in Mexico

    Science.gov (United States)

    Tilghman, Myres W.; Guerena, Don Diego; Licea, Alexei; Pérez-Santiago, Josué; Richman, Douglas D.; May, Susanne; Smith, Davey M.

    2010-01-01

    Background Similar to other resource-limited settings, cost restricts availability of viral load monitoring for most patients receiving antiretroviral therapy in Tijuana, Mexico. We evaluated if a pooling method could improve efficiency and reduce costs while maintaining accuracy. Methods We evaluated 700 patient blood plasma specimens at a reference laboratory in Tijuana for detectable viremia, individually and in 10 × 10 matrix pools. Thresholds for virologic failure were set at ≥500, ≥1000 and ≥1500 HIV RNA copies per milliliter. Detectable pools were deconvoluted using pre-set algorithms. Accuracy and efficiency of the pooling method were compared with individual testing. Quality assurance (QA) measures were evaluated after 1 matrix demonstrated low efficiency relative to individual testing. Results Twenty-two percent of the cohort had detectable HIV RNA (≥50 copies/mL). Pooling methods saved approximately one third of viral load assays over individual testing, while maintaining negative predictive values of >90% to detect samples with virologic failure (≥50 copies/mL). One matrix with low relative efficiency would have been detected earlier using the developed QA measures, but its exclusion would have only increased relative efficiency from 39% to 42%. These methods would have saved between $13,223 and $14,308 for monitoring this cohort. Conclusions Despite limited clinical data, high prevalence of detectable viral loads and a contaminated matrix, pooling greatly improved efficiency of virologic monitoring while maintaining accuracy. By improving cost-effectiveness, these methods could provide sustainability of virologic monitoring in resource-limited settings, and incorporation of developed QA measures will most likely maximize pooling efficiency in future uses. PMID:21124228

  17. Feynman-alpha technique for measurement of detector dead time using a 30 kW tank-in-pool research reactor

    CERN Document Server

    Akaho, E H K; Intsiful, J D K; Maakuu, B T; Nyarko, B J B

    2002-01-01

    Reactor noise analysis was carried out for Ghana Research Reactor-1 GHARR-1, a tank-in-pool type reactor using the Feynman-alpha technique (variance-to-mean method). Measurements made at different detector positions and under subcritical conditions showed that the technique could not be used to determine the prompt decay constant for the reactor which is Be reflected with photo-neutron background. However, for very low dwell times the technique was used to measure the dead time of the detector which compares favourably with the value obtained using the alpha-conventional method.

  18. SAVANNAH RIVER SITE R-REACTOR DISASSEMBLY BASIN IN-SITU DECOMMISSIONING -10499

    Energy Technology Data Exchange (ETDEWEB)

    Langton, C.; Serrato, M.; Blankenship, J.; Griffin, W.

    2010-01-04

    The US DOE concept for facility in-situ decommissioning (ISD) is to physically stabilize and isolate intact, structurally sound facilities that are no longer needed for their original purpose, i.e., generating (reactor facilities), processing(isotope separation facilities) or storing radioactive materials. The 105-R Disassembly Basin is the first SRS reactor facility to undergo the in-situ decommissioning (ISD) process. This ISD process complies with the 105-R Disassembly Basin project strategy as outlined in the Engineering Evaluation/Cost Analysis for the Grouting of the R-Reactor Disassembly Basin at the Savannah River Site and includes: (1) Managing residual water by solidification in-place or evaporation at another facility; (2) Filling the below grade portion of the basin with cementitious materials to physically stabilize the basin and prevent collapse of the final cap - Sludge and debris in the bottom few feet of the basin will be encapsulated between the basin floor and overlying fill material to isolate it from the environment; (3) Demolishing the above grade portion of the structure and relocating the resulting debris to another location or disposing of the debris in-place; and (4) Capping the basin area with a concrete slab which is part of an engineered cap to prevent inadvertent intrusion. The estimated total grout volume to fill the 105-R Reactor Disassembly Basin is 24,384 cubic meters or 31,894 cubic yards. Portland cement-based structural fill materials were designed and tested for the reactor ISD project, and a placement strategy for stabilizing the basin was developed. Based on structural engineering analyses and material flow considerations, maximum lift heights and differential height requirements were determined. Pertinent data and information related to the SRS 105-R Reactor Disassembly Basin in-situ decommissioning include: regulatory documentation, residual water management, area preparation activities, technology needs, fill material

  19. Poroelasticity of high porosity chalk under depletion

    DEFF Research Database (Denmark)

    Andreassen, Katrine Alling; Fabricius, Ida Lykke

    2013-01-01

    levels of pore pressure. The chalk is oil-saturated Lixhe chalk from a quarry near Liège, Belgium, with a general porosity of 45%. Additionally, we compare the theoretical lateral stress to the experimentally determined lateral stress at the onset of pore collapse. The static Biot coefficient based...

  20. Self Compacting Concrete with Chalk Filler

    DEFF Research Database (Denmark)

    Sørensen, Eigil V.

    2007-01-01

    at 28 days from about 35 MPa down to about 13 MPa. The cementing efficiency factor of the chalk filler was found to be in the range 0.21 - 0.42. The chalk filler performed equally well with a grey and a white cement; the latter opens the possibility to produce white SCC more cost effectively....

  1. A CFD based approach for thermal hydraulic design of main vessel cooling system of pool type fast reactors

    International Nuclear Information System (INIS)

    Highlights: ► We study thermal hydraulic design of main vessel cooling system of fast reactors. ► A CFD based approach is proposed for determination of coolant flow rate. ► Effect of cooling system ovality on temperature asymmetry is quantified. ► Suitable flow distribution device is identified to achieve acceptable flow field. ► To compare efficacy of various devices, a flow mal-distribution index is defined. - Abstract: A computational fluid dynamics (CFDs) based approach is proposed for the thermal hydraulic design of the main vessel cooling system for pool type sodium cooled reactors. Usage of the proposed method is demonstrated by applying it to a future Indian commercial fast breeder reactor. Towards quantifying the amount of sodium flow rate for the main vessel cooling system, two-dimensional CFD investigations have been performed. The conjugate conduction–convection models adopted for this purpose are validated against sodium experiments available in literature. The required flow fraction has been determined to be 2.6% of core flow, which is 175.6 kg/s at full power conditions. The heat loss from the hot pool to the cold pool through the main vessel cooling system is estimated to be 10.6 MW at full power and 3.7 MW at 20% power conditions. By detailed three-dimensional CFD studies, the effect of ovality in the main vessel cooling annuli due to manufacturing tolerances has been assessed and the associated circumferential temperature difference in the main vessel is determined to be 14 °C, which is less than the permissible upper limit of 30 °C. The uniformity of sodium flow in the cooling annulus has been investigated by a three-dimensional hydraulic analysis with a view to identify a suitable passive device that can render a uniform velocity distribution. To compare the effectiveness of various devices, a flow mal-distribution index is defined. Detailed parametric studies have been carried out to identify an appropriate porous jet breaker

  2. An Innovative Passive Residual Heat Removal System of an Open-Pool Type Research Reactor with Pump Flywheel and Gravity Core Cooling Tank

    Directory of Open Access Journals (Sweden)

    Kwon-Yeong Lee

    2015-01-01

    Full Text Available In an open-pool type research reactor, the primary cooling system can be designed to have a downward flow inside the core during normal operation because of the plate type fuel geometry. There is a flow inversion inside the core from the downward flow by the inertia force of the primary coolant to the upward flow by the natural circulation when the pump is turned off. To delay the flow inversion time, an innovative passive system with pump flywheel and GCCT is developed to remove the residual heat. Before the primary cooling pump starts up, the water level of the GCCT is the same as that of the reactor pool. During the primary cooling pump operation, the water in the GCCT is moved into the reactor pool because of the pump suction head. After the pump stops, the potential head generates a downward flow inside the core by moving the water from the reactor pool to the GCCT and removes the residual heat. When the water levels of the two pools are the same again, the core flow has an inversion of the flow direction, and natural circulation is developed through the flap valves.

  3. Pool spacing, channel morphology, and the restoration of tidal forested wetlands of the Columbia River, U.S.A.

    Energy Technology Data Exchange (ETDEWEB)

    Diefenderfer, Heida L.; Montgomery, David R.

    2008-10-09

    Tidal forested wetlands have sustained substantial areal losses, and restoration practitioners lack a description of many ecosystem structures associated with these late-successional systems in which surface water is a significant controlling factor on the flora and fauna. The roles of large woody debris in terrestrial and riverine ecosystems have been well described compared to functions in tidal areas. This study documents the role of large wood in forcing channel morphology in Picea-sitchensis (Sitka spruce) dominated freshwater tidal wetlands in the floodplain of the Columbia River, U.S.A. near the Pacific coast. The average pool spacing documented in channel surveys of three freshwater tidal forested wetlands near Grays Bay were 2.2 ± 1.3, 2.3 ± 1.2, and 2.5 ± 1.5. There were significantly greater numbers of pools on tidal forested wetland channels than on a nearby restoration site. On the basis of pool spacing and the observed sequences of log jams and pools, the tidal forested wetland channels were classified consistent with a forced step-pool class. Tidal systems, with bidirectional flow, have not previously been classified in this way. The classification provides a useful basis for restoration project design and planning in historically forested tidal freshwater areas, particularly in regard to the use of large wood in restoration actions and the development of pool habitats for aquatic species. Significant modifications by beaver on these sites warrant further investigation to explore the interactions between these animals and restoration actions affecting hydraulics and channel structure in tidal areas.

  4. New Sensors for Irradiation Testing at Materials and Test Reactors

    International Nuclear Information System (INIS)

    Enhanced instrumentation, capable of providing real-time measurements of parameters during fuels and material irradiations, is required to support irradiation testing requested by US nuclear research programs. For example, several research programs funded by the US Department of Energy (US DOE) are emphasizing the use of first principle models to characterize the performance of fuels and materials. To facilitate this approach, high fidelity, real-time data are essential to demonstrate the performance of these new fuels and materials during irradiation testing. Furthermore, sensors that obtain such data in US MTRs, such as the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL), must be miniature, reliable, and able to withstand high fluxes and high temperatures. Depending on program requirements, sensors may need to obtain data in inert gas, pressurized water, or liquid metal environments. To address these needs, INL has developed and deployed several new sensors to support irradiation testing in US DOE programs. The paper identifies the sensors currently available to support higher flux US MTR irradiations. Recent results and products from sensor research and development are highlighted. In particular, progress in deploying enhanced in-pile sensors for detecting temperature, elongation, and thermal conductivity is emphasized. Finally, initial results from research to evaluate the viability of ultrasonic and fiber optic technologies for irradiation testing are summarized. (author)

  5. TR-EDB: Test Reactor Embrittlement Data Base, Version 1

    Energy Technology Data Exchange (ETDEWEB)

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K. [Oak Ridge National Lab., TN (United States)

    1994-01-01

    The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.

  6. TR-EDB: Test Reactor Embrittlement Data Base, Version 1

    International Nuclear Information System (INIS)

    The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results

  7. Sediment transport and siltation of brown trout (Salmo trutta L.) spawning gravels in chalk streams

    Science.gov (United States)

    Acornley, R. M.; Sear, D. A.

    1999-02-01

    Deposition rates of fine sediment into brown trout spawning gravels were measured at monthly intervals for a period of one year in a small channel of the River Test, Hampshire. Data were also collected on stream discharge, water depth, flow velocity and suspended sediment concentrations. Deposition rates followed a seasonal pattern and were maximal during periods of high discharge in the late winter/early spring when suspended sediment concentrations were high. The material deposited in the spawning gravels included silts and fine sands (<250 m) that were transported in suspension and coarser fragments of low density tufa-like material that were transported as bed load. The ecological implications of fine sediment deposition for salmonid egg survival in chalk streams are considered.

  8. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 19

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  9. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 17

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  10. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 1

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  11. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 13

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  12. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 5

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  13. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 5a

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  14. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 3

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  15. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 22

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  16. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 24

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  17. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 16

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  18. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 26

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  19. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 2

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  20. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 8

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  1. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 4

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  2. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 7

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  3. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 20

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  4. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 15

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  5. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 14

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  6. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 25

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  7. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 11

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  8. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 10

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  9. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 21

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  10. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 18

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  11. 1890's Land Cover/Use - Mississippi River Commission Surveys, Pool 12

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — In the late 1880's and early 1900's the Mississippi River Commission (MRC) conducted an extensive high-resolution survey of the Mississippi River from Cairo,...

  12. Optimal Item Pool Design for a Highly Constrained Computerized Adaptive Test

    Science.gov (United States)

    He, Wei

    2010-01-01

    Item pool quality has been regarded as one important factor to help realize enhanced measurement quality for the computerized adaptive test (CAT) (e.g., Flaugher, 2000; Jensema, 1977; McBride & Wise, 1976; Reckase, 1976; 2003; van der Linden, Ariel, & Veldkamp, 2006; Veldkamp & van der Linden, 2000; Xing & Hambleton, 2004). However, studies are…

  13. Optimisation of the flow path in a conceptual pool type reactor under natural circulation with lead coolant

    International Nuclear Information System (INIS)

    This contribution investigates the effects of a bypass flow blocking bottom plate and the influence of the heat transfer between the hot and cold leg in a small pool type reactor cooled through natural convection with lead coolant. The computations are carried out using 3D computational fluid dynamics, where small-detail parts, such as the core and heat exchangers are modeled using a porous media approach. The introduction of full conjugate heat transfer shows that the heat transfer between the hot and cold leg can deteriorate flow in the cold leg and lead to recirculation zones. These zones become even more pronounced with the introduction of a bottom plate, which on the other hand also increases the flow through the core and lowers the maximum temperature in the core by approximately 150 K. Based on the results, redesign suggestions for the bottom plate and the internal wall are made. (author)

  14. Test results from a full-scale sodium reflux pool-boiler solar receiver

    Science.gov (United States)

    Moreno, J. B.; Andraka, C. E.; Diver, R. B.; Ginn, W. C.; Dudley, V.; Rawlinson, K. S.

    1990-01-01

    A sodium reflux pool-boiler solar receiver has been tested on a nominal 75 kW sub t parabolic-dish concentrator. The purpose was to demonstrate the feasibility of reflux-receiver technology for application to Stirling-engine dish-electric systems. In this application, pool boilers (and more generally liquid-metal reflux receivers) have a number of advantages over directly-illuminated tube receivers. The advantages, to be discussed, include more uniform temperature, which results in longer lifetime and higher temperature available to the engine.

  15. Criticality safety assessment of a TRIGA reactor spent fuel pool under accident conditions

    International Nuclear Information System (INIS)

    An overview paper on the criticality safety analysis of a pool type storage for a TRIGA spent fuel at the ''Jozef Stefan'' Institute in Ljubljana, Slovenia, is presented. It was shown in that subcriticality is not guaranteed for some postulated accidents (an earthquake with subsequent fuel rack disintegration resulting in contact fuel pitch). To mitigate this deficiency, a study was made about replacing a certain number of fuel elements in the rack with absorber rods in order to lower the probability for supercriticality to acceptable level. (author)

  16. Current Status of Activation Analysis Using Tsing Hua Open-Pool Reactor

    International Nuclear Information System (INIS)

    Activation analysis may be considered to be one of the most practical applications in the utilization of research reactor. One of the obvious advantages of this method is its high sensitivity for many elements. Using the average research reactor, it is feasible to determine the content of element in various matrices, to an order of as low as 0,001 microgram. Although the detection, sensitivities of other analytical methods also have been improved very much in recent years by using newly developed instruments, these methods can rival the activation analysis in sensitivity for only a certain limited number of elements. Thus, at present, the technique of activation analysis seems to offer the highest sensitivity for the greatest number of elements

  17. Concept of a BNCT line with in-pool fission converter at MARIA reactor in Swierk

    Science.gov (United States)

    Pytel, Krzysztof; Andrzejewski, Krzysztof; Golnik, Natalia; Osko, Jakub

    2009-01-01

    BNCT facility in the Institute of Atomic Energy in Otwock-Swierk is under construction at the horizontal channel H2 of the research reactor MARIA. Measurements of the neutron energy spectrum performed at the front of the H2 experimental channel, have shown that flux of epithermal neutrons (above 10 keV) at the BNCT irradiation port was below 109 n cm-2 s-1 i.e. it was too low to be directly used for the BNCT treatment. Therefore, a fission converter will be placed between the reactor core and the periphery of the graphite reflector of MARIA reactor. The uranium converter will be powered by the densely packed EK-10 fuel elements with 10% enrichment. Preliminary calculations have shown that the total neutron flux in the converter will be about 1013 n cm-2 s-1 and flux of epithermal neutrons at the entrance to the filter/moderator of the beam will be about 2·1013 n cm-2 s-1.

  18. Some aspects of criticality safety of TRIGA reactor spent fuel pool

    International Nuclear Information System (INIS)

    Additional criticality safety analysis of a pool type storage for TRIGA spent fuel at ''Jozef Stefan'' Institute in Ljubljana, Slovenia, is presented. Previous results have shown that subcriticality is not guaranteed for some postulated accidents. To mitigate this deficiency, a study was made about replacing a certain number of fuel elements in the rack with absorber rods. For this purpose Monte Carlo computer code MCNP4A with ENDF-B/V library and detailed three dimensional fuel rack model was used. A short study of different absorber rods design is presented. At first the analysis was done about the number of uniformly mixed absorber rods in the lattice needed to sustain the subcriticality of the storage when pitch is decreased from rack design pitch of 8cm to contact, assuming that the absorber rods remain in their proper positions. Because of supercriticality possibility due to random mixing of the absorber rods during lattice compaction, a probabilistic study was made, sampling the probability density functions for random lattice loadings of the absorber rods. The results show reasonably low probabilities for supercriticality even for fresh 12 wt% enriched standard TRIGA fuel stored in the spent fuel pool. (orig.)

  19. Status of the irradiation test vehicle for testing fusion materials in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.C.; Smith, D.L. [Argonne National Lab., IL (United States); Palmer, A.J.; Ingram, F.W. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States); Wiffen, F.W. [Dept. of Energy, Germantown, MD (United States). Office of Fusion Energy

    1998-09-01

    The design of the irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) has been completed. The main application for the ITV is irradiation testing of candidate fusion structural materials, including vanadium-base alloys, silicon carbide composites, and low-activation steels. Construction of the vehicle is underway at the Lockheed Martin Idaho Technology Company (LMITCO). Dummy test trains are being built for system checkout and fine-tuning. Reactor insertion of the ITV with the dummy test trains is scheduled for fall 1998. Barring unexpected difficulties, the ITV will be available for experiments in early 1999.

  20. A river basin as a common-pool resource: a case study for the Jaguaribe basin in the semi-arid Northeast of Brazil

    NARCIS (Netherlands)

    Oel, van Pieter R.; Krol, Maarten S.; Hoekstra, Arjen Y.

    2009-01-01

    This paper applies 'common-pool resource' concepts to analyse to which extent the physical characteristics of a river basin facilitate or impede good management of water in different parts of a river basin. In addition, we compare the apparent manageability of water in the different parts of the bas

  1. Heterogeneity of soil carbon pools and fluxes in a channelized and a restored floodplain section (Thur River, Switzerland

    Directory of Open Access Journals (Sweden)

    E. Samaritani

    2011-01-01

    Full Text Available Due to their spatial complexity and dynamic nature, floodplains provide a wide range of ecosystem functions. However, because of flow regulation, many riverine floodplains have lost their characteristic heterogeneity. Restoration of floodplain habitats and the rehabilitation of key ecosystem functions has therefore become a major goal of environmental policy. Many important ecosystem functions are linked to organic carbon (C dynamics in riparian soils. The fundamental understanding of the factors that drive the processes involved in C cycling in heterogeneous and dynamic systems such as floodplains is however only fragmentary.

    We quantified soil organic C pools (microbial C and water extractable organic C and fluxes (soil respiration and net methane production in functional process zones of adjacent channelized and widened sections of the Thur River, NE Switzerland, on a seasonal basis. The objective was to assess how spatial heterogeneity and temporal variability of these pools and fluxes relate to physicochemical soil properties on one hand, and to soil environmental conditions and flood disturbance on the other hand.

    Overall, factors related to seasonality and flooding (temperature, water content, organic matter input affected soil C dynamics more than soil properties did. Coarse-textured soils on gravel bars in the restored section were characterized by low base-levels of organic C pools due to low TOC contents. However, frequent disturbance by flood pulses led to high heterogeneity with temporarily and locally increased pools and soil respiration. By contrast, in stable riparian forests, the finer texture of the soils and corresponding higher TOC contents and water retention capacity led to high base-levels of C pools. Spatial heterogeneity was low, but major floods and seasonal differences in temperature had additional impacts on both pools and fluxes. Soil properties and base levels of C pools in the dam foreland of the

  2. Heterogeneity of soil carbon pools and fluxes in a channelized and a restored floodplain section (Thur River, Switzerland

    Directory of Open Access Journals (Sweden)

    E. Samaritani

    2011-06-01

    Full Text Available Due to their spatial complexity and dynamic nature, floodplains provide a wide range of ecosystem functions. However, because of flow regulation, many riverine floodplains have lost their characteristic heterogeneity. Restoration of floodplain habitats and the rehabilitation of key ecosystem functions, many of them linked to organic carbon (C dynamics in riparian soils, has therefore become a major goal of environmental policy. The fundamental understanding of the factors that drive the processes involved in C cycling in heterogeneous and dynamic systems such as floodplains is however only fragmentary.

    We quantified soil organic C pools (microbial C and water extractable organic C and fluxes (soil respiration and net methane production in functional process zones of adjacent channelized and widened sections of the Thur River, NE Switzerland, on a seasonal basis. The objective was to assess how spatial heterogeneity and temporal variability of these pools and fluxes relate to physicochemical soil properties on one hand, and to soil environmental conditions and flood disturbance on the other hand.

    Overall, factors related to seasonality and flooding (temperature, water content, organic matter input affected soil C dynamics more than soil properties did. Coarse-textured soils on gravel bars in the restored section were characterized by low base-levels of organic C pools due to low TOC contents. However, frequent disturbance by flood pulses led to high heterogeneity with temporarily and locally increased C pools and soil respiration. By contrast, in stable riparian forests, the finer texture of the soils and corresponding higher TOC contents and water retention capacity led to high base-levels of C pools. Spatial heterogeneity was low, but major floods and seasonal differences in temperature had additional impacts on both pools and fluxes. Soil properties and base levels of C pools in the dam foreland of the channelized section

  3. Proceedings of the international symposium on materials testing reactors

    International Nuclear Information System (INIS)

    This report is the Proceedings of the International Symposium on Materials Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The symposium was held on July 16 to 17, 2008, at the Oarai Research and Development Center of JAEA. This symposium was also held for the 40th anniversary ceremony of Japan Materials Testing Reactor (JMTR) from achieving its first criticality. The objective of the symposium is to exchange the information on current status, future plan and so on among each testing reactors for the purpose of mutual understanding. There were 138 participants from Argentina, Belgium, France, Indonesia, Kazakhstan, Korea, the Russian Federation, Sweden, the United State, Vietnam and Japan. The symposium was divided into four technical sessions and three topical sessions. Technical sessions addressed the general topics of 'status and future plan of materials testing reactors', 'material development for research and testing reactors', irradiation technology (including PIE technology)' and 'utilization with materials testing reactors', and 21 presentations were made. Also the topical sessions addressed 'establishment of strategic partnership', 'management on re-operation work at reactor trouble' and 'basic technology for neutron irradiation tests in MTRs', and panel discussion was made. The 21 of the presented papers are indexed individually. (J.P.N.)

  4. Development, utilization, and future prospects of materials test reactors

    International Nuclear Information System (INIS)

    Reactor radiation affects the chemical and physical properties of materials. These changes can be very drastic in certain cases. Special test reactors have therefore been built since the 1950's and specific skills were developed to expose materials specimens to the precise irradiation conditions required. Materials testing reactors are those research reactor facilities which are designed and operated predominantly for studies into radiation damage. About a dozen plants in European communities (EC) Member States and in the US can be identified in this category, with 5 to 100 MW fission power and neutron fluxes between 5 x 1013 and 1015 cm-2s-1. The paper elaborates common aspects of development, utilization, and future prospects of US and EC materials testing reactors, and indicates the most significant differences

  5. N-16 power monitoring system of the RP-10 pool-type reactor

    International Nuclear Information System (INIS)

    The preliminary results of monitoring of power of the RP-10 nuclear reactor by measuring the activity of gamma radiation 16N content in the coolant are presented. A detector NaI(Tl) placed in a window that communicates the decay tank and the pump room of the primary cooling circuit was used. Measurements were performed for different levels of power, from 0,5 to 10 MW. Results show a linear behavior between the power of operation and the activity of 16N. (orig.)

  6. Analyses of pool swell tests by two-dimensional hydrodynamic computer code

    Energy Technology Data Exchange (ETDEWEB)

    Shimegi, Nobuo; Suzuki, Kenichi

    1988-10-01

    A two-dimensional hydrodynamic computer code SOLA-VOF was examined on the analytical capability for dynamic loads by pool swell in the MARK-I type BWR suppression chamber under LBLOCA (Large Break Loss of Coolant Accident) conditions. Two pool swell tests, (LLL 1/5-scale and EPRI 1/12-scale tests) were selected for this purpose and analyzed by the SOLA-VOF code modified with incorporation of a simple downcomer flow model. In these analyses, it was necessary to take account of three-dimensional effect of pool swell behavior along the chamber axis by use of a method such as spatially weighting function experimentally determined, because a simple two-dimensional calculation by the SOLA-VOF code gave too much conservative evaluation for the impact load on the ring header. Applications of this method gave a good agreement between the calculation and measurement. The vertical loads on the suppression chamber wall were well analyzed by this code. It might be because the local pressure difference caused by the nonuniform pool swelling disappeared owing to pressure integration on the surface of suppression chamber wall.

  7. Analyses of pool swell tests by two-dimensional hydrodynamic computer code

    International Nuclear Information System (INIS)

    A two-dimensional hydrodynamic computer code SOLA-VOF was examined on the analytical capability for dynamic loads by pool swell in the MARK-I type BWR suppression chamber under LBLOCA (Large Break Loss of Coolant Accident) conditions. Two pool swell tests, (LLL 1/5-scale and EPRI 1/12-scale tests) were selected for this purpose and analyzed by the SOLA-VOF code modified with incorporation of a simple downcomer flow model. In these analyses, it was necessary to take account of three-dimensional effect of pool swell behavior along the chamber axis by use of a method such as spatially weighting function experimentally determined, because a simple two-dimensional calculation by the SOLA-VOF code gave too much conservative evaluation for the impact load on the ring header. Applications of this method gave a good agreement between the calculation and measurement. The vertical loads on the suppression chamber wall were well analyzed by this code. It might be because the local pressure difference caused by the nonuniform pool swelling disappeared owing to pressure integration on the surface of suppression chamber wall. (author)

  8. Fukushima - calculation of the reactor core inventory and storage pools Dai-ichi 1 to Dai-ichi 4, an estimation of a source term

    International Nuclear Information System (INIS)

    Inventory of the reactor core and spent fuel storage pool of the reactors at Dai-ichi 1 to Dai-ichi 4 was determined to need a realistic estimate of the source (released into the atmosphere environment) and modelling of radiological impact of the events in Fukushima NPP. Calculations of inventories were carried out by the methodology that is used in systems to support emergency response and crisis management anymore. Calculations were made based on a model that respects knowledge of real fuels and fuel cycles for individual reactors Dai-ichi. Necessary input data for training the model and calculate inventories are obtained from the IAEA PRIS database.

  9. Safety evaluation report related to the renewal of the operating license for the Worcester Polytechnic Institute open-pool training reactor, Docket No. 50-134

    International Nuclear Information System (INIS)

    This Safety Evaluation Report for the application filed by the Worcester Polytechnic Institute (WPI) for a renewal of Operating License R-61 to continue to operate the WPI 10-kW open-pool training reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by the Worcester Polytechnic Institute and is located on the WPI campus in Worcester, Worcester County, Massachusetts. The staff concludes that the reactor facility can continue to be operated by WPI without endangering the health and safety of the public

  10. Safety of research reactors (Design and Operation)

    International Nuclear Information System (INIS)

    The primary objective of this thesis is to conduct a comprehensive up-to-date literature review on the current status of safety of research reactor both in design and operation providing the future trends in safety of research reactors. Data and technical information of variety selected historical research reactors were thoroughly reviewed and evaluated, furthermore illustrations of the material of fuel, control rods, shielding, moderators and coolants used were discussed. Insight study of some historical research reactors was carried with considering sample cases such as Chicago Pile-1, F-1 reactor, Chalk River Laboratories,. The National Research Experimental Reactor and others. The current status of research reactors and their geographical distribution, reactor category and utilization is also covered. Examples of some recent advanced reactors were studied like safety barriers of HANARO of Korea including safety doors of the hall and building entrance and finger print identification which prevent the reactor from sabotage. On the basis of the results of this research, it is apparent that a high quality of safety of nuclear reactors can be attained by achieving enough robust construction, designing components of high levels of efficiency, replacing the compounds of the reactor in order to avoid corrosion and degradation with age, coupled with experienced scientists and technical staffs to operate nuclear research facilities.(Author)

  11. Nickel adsorption on chalk and calcite

    DEFF Research Database (Denmark)

    Belova, Dina Alexandrovna; Lakshtanov, Leonid; Carneiro, J.F.;

    2014-01-01

    Nickel uptake from solution by two types of chalk and calcite was investigated in batch sorption studies. The goal was to understand the difference in sorption behavior between synthetic and biogenic calcite. Experiments at atmospheric partial pressure of CO2, in solutions equilibrated with calcite...... = - 1.12 on calcite and log KNi = - 0.43 and - 0.50 on the two chalk samples. The study confirms that synthetic calcite and chalk both take up nickel, but Ni binds more strongly on the biogenic calcite than on inorganically precipitated, synthetic powder, because of the presence of trace amounts...

  12. Significance of coast down time on safety and availability of a pool type fast breeder reactor

    International Nuclear Information System (INIS)

    Highlights: • Plant dynamics studies for quantifying the benefits of flow coast down time. • Establishment of minimum flow coast down time required for safety. • Assessment of influence of flow coast down on enhancing plant availability. • Synthesis of thermo mechanical benefits of flow coast down time on component design. - Abstract: Plant dynamic investigation towards establishing the influence of flow coast down time of primary and secondary sodium systems on safety and availability of plant has been carried out based on one dimensional analysis. From safety considerations, a minimum flow coast down time for primary sodium circuit is essential to be provided to limit the consequences of loss of flow event within allowable limits. Apart from safety benefits, large primary coast down time also improves plant availability by the elimination of reactor SCRAM during short term power failure events. Threshold values of SCRAM parameters also need optimization. By suitably selecting the threshold values for SCRAM parameters, significant reduction in the inertia of pumping systems can be derived to obtain desirable results on plant availability. With the optimization of threshold values and primary flow coast down behaviour equivalent to a halving time of 8 s, there is a possibility to eliminate reactor SCRAM during short term power failure events extending up to 0.75 s duration. Benefits of secondary flow halving on reducing transient thermal loading on components have also been investigated and mixed effects have been observed

  13. Tests of Selection in Pooled Case-Control Data: An Empirical Study

    Directory of Open Access Journals (Sweden)

    Nitin eUdpa

    2011-11-01

    Full Text Available For smaller organisms with faster breeding cycles, artificial selection can be used to create sub-populations with different phenotypic traits. Genetic tests can be employed to identify the causal markers for the phenotypes, as a precursor to engineering strains with a combination of traits. Traditional approaches involve analyzing crosses of inbred strains to test for co-segregation with genetic markers. Here we take advantage of cheaper next generation sequencing techniques to identifygenetic signatures of adaptation to the selection constraints. Obtaining individual sequencing data is often unrealistic due to cost and sample issues, so we focus on pooled genomic data.In this paper, we explore a series of statistical tests for selection using pooled case (under selection and control populations. Extensive simulations are used to show that these approaches work well for a wide range of population divergence times and strong selective pressures. We show that pooling does not have a significant impact on statistical power. The tests are also robust to reasonable variations in several different parameters, including window size, base-calling error rate, and sequencing coverage. We then demonstrate the viability (and the challenges of one of these methods in two independent Drosophila populations (Drosophila melanogaster bred under selectionfor hypoxia and accelerated development, respectively. Testing for extreme hypoxia tolerance showed clear signals of selection, pointing to loci that are important for hypoxia adaptation.Overall, we outline a strategy for finding regions under selection using pooled sequences, then devise optimal tests for that strategy. The approaches show promise for detecting selection, even several generations after fixation of the beneficial allele has occurred.

  14. Status and future plan of Japan materials testing reactor

    International Nuclear Information System (INIS)

    The Japan Materials Testing Reactor (JMTR) of Japan Atomic Energy Agency (JAEA) is a light water cooling tank typed reactor. JMTR has been used for fuel and material irradiation studies for LWRs, HTGR, fusion reactor and RI production. Since the JMTR is connected with hot laboratory through the canal, re-irradiation tests can conduct easily by safety and quick transportation of irradiation samples. First criticality was achieved in March 1968, and operation was stopped from August, 2006 for the refurbishment. The reactor facilities are refurbished during four years from the beginning of FY 2007, and necessary examination and work are carrying out on schedule. The renewed and upgraded JMTR will start from FY 2011 and operate for a period of about 20 years (until around FY 2030). The usability improvement of the JMTR, such as higher reactor available factor, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussing as the preparations for re-operation. (author)

  15. Irradiations in swimming-pool type reactors from room temperature up to 2000 deg C

    International Nuclear Information System (INIS)

    The irradiations which have been, and are being carried out in the Melusine and Siloe reactors in connection with pure or applied research projects, are effected in widely varying conditions; amongst these, for example, the temperature may vary from -250 deg C to +2000 deg C The eight devices presented are designed for irradiations effected at temperatures of from room temperature up to 2000 deg C. 1. Irradiation device for irradiation at normal temperatures 2. The 'PEF' device 3. The 'CHOUCA' device, 150 to 900 deg C 4. The 'CYRANO' device for EL 4 conditions 5. 'HT' capsules, 800-1000 deg C 6. The 'HEBE' furnace 1400 deg C 7. The 'PEC' device, 1400 deg C 8. The 'HF' furnace 2000 deg C. (authors)

  16. Design of a decay tank for a pool type research reactor with a CFD model

    International Nuclear Information System (INIS)

    A conceptual primary cooling system (PCS) was designed for adequate cooling of the core of a research reactor. The primary coolant after passing through the reactor core contains many kinds of radio-nuclides. A decay tank provides a delayed transit time to ensure that the N-16 activity decreases enough before the coolant leaves the decay tank's shielding room. The size of the decay tank should be enlarged to provide sufficient transit time. However, there was a limitation: to minimize the tank size, it should be designed with an internal baffle, which affects the pressure loss in the system and net positive suction head (NPSH) of the PCS pump. Therefore, the decay tank should be optimized for size and the internal baffle. A vertical type decay tank was chosen to optimize the geometrical arrangement of PCS and the vertical internal baffle was installed to minimize the number of internal structures. The preliminary geometry of the tank and the internal baffle were determined to satisfy the required delayed transit time by calculating the maximum velocity and the flow path length of the circular and the annular sections of the tank. The commercially available CFD model, FLUENT, which solves the Navier-Stokes and turbulent models, was used to specifically design the decay tank with the preliminarily calculated geometry and the related flow rate. Several turbulence models, standard k-ε model, renormalization group (RNG) model, and realizable k-ε model, were conducted to isolate the root cause of these differences. By comparing the results of the velocity profile and the characteristics of each model, a detailed design study was simulated using the realizable k-ε model. A user-defined scalar equation was solved to estimate the delayed transit time. The size and the internal baffle that satisfy the required transit time were determined based on the CFD results. (author)

  17. Preliminary steps in partial decommissioning of a swimming pool type reactor Apsara: a health physics experience

    International Nuclear Information System (INIS)

    Full text: Apsara reactor after 50 years of extensive use in production of radio-isotopes, neutron radiography, neutron beam research, shielding experiments etc., is undergoing a partial decommissioning to facilitate refurbishment and up gradation to 2 MWth power using lower enriched Uranium fuel. Partial decommissioning involves defueling and removal of core components as a first step. Radiological safety in defueling is discussed in this paper. Defueling is carried out from top of pile where the radiation level was < 0.10 mR/h, under the strict stipulation that no fuel is to be lifted above water surface during transfer and hence no dose is consumed for this job. Collective dose consumed in the job was only in the SFSB area in Dhruva reactor and was 8.20 person mSv (59% of budgeted dose). This was possible by thorough mock up at Dhruva, SFSB for irradiated fuel handling, satisfying ALARA and refining the procedure. Also radiation mapping of core components, grid plate before and after removal of dry and wet guide tubes were carried out. It was observed that the top portion of grid plate showed a maximum radiation level which was 2-3 times that at bottom portion. The positions around the dry guide tubes G 1 and G 7 showed high radiation levels of 30 R/h. On removing them, the radiation levels reduced to 0.3 -0.5 R/h in all positions. This acted as an input for planning to cut and remove various core components, as also for segregation of high active/low active/inactive components from waste disposal point of view. For example the dry guide tubes G 1 and G 7 was cut and disposed as active waste with upper portion as Cat I and lower portion (was in core region) as Cat II radioactive waste

  18. Research reactors

    International Nuclear Information System (INIS)

    This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

  19. Improving the proliferation resistance of research and test reactors

    International Nuclear Information System (INIS)

    Elimination, or substantial reduction, of the trade in highly enriched fuel elements for research and test reactors would significantly reduce the proliferation risk associated with the current potential for diversion of these materials. To this end, it is the long-term goal of U.S. policy to fuel all new and existing research and test reactors with uranium of less than 20% enrichment (but substantially greater than natural) excepting, perhaps, only a small number of high-power, high-performance, reactors. The U.S. development program for enrichment reduction in research and test reactor designs currently using 90-93% enriched uranium is based on the practical criterion that enrichment reduction should not cause significant flux performance (flux per unit power) or burnup performance degradation relative to the unmodified reactor design. A program is now beginning in the U.S. to develop the necessary fuel technology, but several years of work will be needed. Accordingly, as an immediate interim step, the U.S. is proposing to convert existing research and test reactors (and new designs) from the use of 90-93% enriched fuel to the use of 30-45% enriched fuel wherever this can be done without unacceptable reactor performance degradation

  20. A Note on the Pooling of Individual PANIC Unit Root Tests

    OpenAIRE

    Westerlund, Joakim

    2007-01-01

    One of the most cited studies in recent years within the field of nonstationary panel data analysis is that of Bai and Ng (2004, A PANIC Attack on Unit Roots and Cointegration. Econometrica 72, 1127-1177), in which the authors propose PANIC, a new framework for analyzing the nonstationarity of panels with idiosyncratic and common components. This paper shows that, although valid at the level of the individual unit, PANIC is not an asymptotically valid framework for pooling tests at the aggreg...

  1. Effect of coupon orientation on corrosion behavior of aluminium alloy coupons in the spent fuel storage section of the IEA-R1 research reactor pool

    International Nuclear Information System (INIS)

    Full text: The objectives of the IAEA sponsored Regional Technical Co-operation Project for Latin America (Argentina, Brazil, Chile, Mexico, and Peru) are to provide the basic conditions to define a regional strategy for managing spent fuel and to provide solutions, taking into consideration the economic and technological realities of the countries involved. In particular, to determine the basic conditions for managing research reactor (RR) spent fuel during operation and interim storage as well as final disposal, and to establish forms of regional cooperation for spent fuel characterization, safety, regulation and public communication. This project is divided into 4 subprojects: (1) spent fuel characterization; (2) safety and regulation; (3) options for spent fuel storage and disposal; (4) public information and communication. Corrosion monitoring and surveillance is one of the activities of the subproject 'spent fuel characterization'. The dominant fuel type used in the Latin American (LA) RRs is plate-type (MTR), LEU, oxide fuel (U3O8-Al) clad in Al, followed by TRIGA-type (U-Zr-H) rods. Almost all the spent fuels from LA RRs are stored in storage racks within the reactor pool, in decay pools or in away-from-reactor wet basins. Two of the countries participating in the Regional Project (Argentina and Brazil) were (and continue to be) participants of the IAEA sponsored coordinated research project (CRP) on 'Corrosion of research reactor Al-clad spent fuel in water'. The corrosion surveillance activities of the Regional Project are based on this CRP. The main objective of this activity is to evaluate the effect of LA spent fuel basin parameters on the corrosion behaviour of RR fuel cladding. This evaluation consists of exposing racks containing Al alloy and stainless steel coupons at the different spent fuel basins. Al alloy coupons fabricated under conditions similar to those used to prepare fuel plates have been used. The surface conditions of the Al coupons

  2. The North Sea reservoir chalk characterization

    OpenAIRE

    Kallesten, Emanuela

    2015-01-01

    A significant amount of the hydrocarbon production in the North Sea is related to chalk reservoirs. Since 1969, the chalk play remains one of the most important oil sources in Norway. With the initial expected recovery factor 17%, development in technologies and methods contributed to a substantial increase in oil recovery to an approximately 40%. Much of the reserves in place are yet to be extracted, and secondary and tertiary recovery methods need to advance in order to mobilize the remaini...

  3. Chalk: composition, diagenesis and physical properties

    OpenAIRE

    Fabricius, Ida Lykke

    2007-01-01

    Chalk is a sedimentary rock of unusually high homogeneity on the scale where physical properties are measured, but the properties fall in wide ranges. Chalk may thus be seen as the ideal starting point for a physical understanding of rocks in general. Properties as porosity, permeability, capillary entry pressure, and elastic moduli are consequences of primary sediment composition and ofsubsequent diagenetic history as caused by microbial action, burial stress, temperature, and pore pressure....

  4. Subsidence and capillary effects in chalks

    OpenAIRE

    Delage, Pierre; Schroeder, Christian; Cui, Yu-Jun

    1996-01-01

    Based on the concepts of the mechanics of unsaturated soils where capillary phenomena arise between the wetting fluid (water) and the non-wetting one (air), the subsidence of chalks containing oil (non-wetting fluid) during water injection (wetting fluid) is analysed. It is shown that the collapse phenomenon of unsaturated soils under wetting provides a physical explanation and a satisfactory prediction of the order of magnitude of the subsidence of the chalk. The use of a well established co...

  5. Preliminary Test Estimators and Phi-divergence Measures in Pooling Binomial Data

    Directory of Open Access Journals (Sweden)

    N. Martin

    2012-07-01

    Full Text Available Two independent random samples are drawn from two Binomial populations with parameters theta1 and  theta2 respectively. Ahmed (1991 considered a preliminary test estimator based on maximum likelihood estimator for estimating theta1 when it is suspected that theta1=theta2.  In this paper we combine minimum phi-divergence estimators as well as phi-divergence test statistics in order to define a preliminary phi-divergence test estimators. These new estimators are compared with the classical estimator as well as the pooled estimator.

  6. Differential mobilization of terrestrial carbon pools in Eurasian Arctic river basins

    NARCIS (Netherlands)

    Feng, X.; Vonk, J.E.; van Dongen, B.E.; Gustafsson, Ö.; Semiletov, I.P.; Dudarev, O.V.; Wang, Z.; Montluçon, D.B.; Wacker, L.; Eglinton, T.I.

    2013-01-01

    Mobilization of Arctic permafrost carbon is expected to increase with warming-induced thawing. However, this effect is challenging to assess due to the diverse processes controlling the release of various organic carbon (OC) pools from heterogeneous Arctic landscapes. Here, by radiocarbon dating var

  7. First on-sun test of NaK pool-boiler solar receiver

    Science.gov (United States)

    Moreno, J. B.; Andraka, C. E.; Moss, T. A.; Cordeiro, P. G.; Dudley, V. E.; Rawlinson, K. S.

    During 1989-1990, a refluxing liquid-metal pool-boiler solar receiver designed for dish/Stirling application at 75 kW(sub t) throughput was successfully demonstrated at Sandia National Laboratories. Significant features of this receiver included (1) boiling sodium as the heat transfer medium, and (2) electric-discharge-machined (EDM) cavities as artificial nucleation sites to stabilize boiling. Following this first demonstration, a second-generation pool-boiler receiver that brings the concept closer to commercialization has been designed, constructed, and successfully tested. For long life, the new receiver is built from Haynes Alloy 230. For increased safety factors against film boiling and flooding, the absorber area and vapor-flow passages have been enlarged. To eliminate the need for trace heating, sodium has been replaced by the sodium-potassium alloy NaK-78. To reduce manufacturing costs, the receiver has a powdered-metal coating instead of EDM cavities for stabilization of boiling. To control incipient-boiling superheats, especially during hot restarts, it contains a small amount of xenon. In this paper, we present the receiver design and report the results of on-sun tests using a nominal 75 kW(sub t) test-bed concentrator to characterize boiling stability, hot-restart behavior, and thermal efficiency at temperatures up to 750 C. We also report briefly on late results from an advanced-concepts pool-boiler receiver.

  8. Irradiation testing of miniature fuel plates for the RERTR program. [Reduced Enrichment Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Senn, R.L.; Martin, M.M.

    1981-07-01

    An irradiation test facility, which provides a test bed for irradiating a variety of miniature fuel plates (miniplates) for the Reduced Enrichment Research and Test Reactors (RERTR) program, has been placed into operation. These tests screen various candidate fuel materials on their suitability for replacing the highly enriched uranium fuel materials currently used by the world's test and research reactors with a lower enrichment fuel material, without significantly degrading reactor operating characteristics and power levels. The use of low uranium enrichment of about 20% /sup 235/U in place of highly enriched fuel for these reactors would reduce the potential for /sup 235/U diversion. The irradiation test facility, designated as HFED, is operating in core position E-7 in the Oak Ridge Research Reactor (ORR), a 30-MW water-moderated reactor. The miniplates will achieve burnups of up to approx. 2.2 x 10/sup 27/ fissions/m/sup 3/ of fuel.

  9. In-Research Reactor Tests for SCWR Fuel Verifications

    International Nuclear Information System (INIS)

    The Supercritical water cooled reactors (SCWRs) are essentially light water reactors (LWRs) operating at higher pressure and temperature. The SCWRs achieve high thermal efficiency (i.e., about 45% vs. about 35% efficiency for advanced LWRs) and are simpler plants as the need for many of the traditional LWR components is eliminated. The SCWRs build upon two proven technologies, the LWR and the supercritical coal-fired boiler. The main mission of the SCWR is production of low-cost electricity. Thus the SCWR is also suited for hydrogen generation with electrolysis, and can support the development of the hydrogen economy in the near term. In this paper, the SCWR fuel performance verification tests are reviewed. Based on this review results, in-research reactor verification tests to be performed in a fuel test loop through the international joint program are proposed. In addition, capsule tests and fuel test loop tests to be performed in HANARO are also proposed

  10. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 6

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  11. 2011 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 21

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  12. Bathymetric data for the Upper Mississippi and Illinois Rivers -- Pool 21

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — Water depth is an important feature of aquatic systems. On the Upper Mississippi River System (UMRS), water depth data are important for describing the physical...

  13. 2010 UMRS Color Infrared Aerial Photo Mosaic - Illinois River, Peoria Pool

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  14. 2002 Land Cover/Use Data for Pool 8 - Upper Mississippi River

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) has created a high-resolution land cover/use data set for Mississippi River...

  15. Bathymetric data for the Upper Mississippi and Illinois Rivers -- La Grange Pool

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — Water depth is an important feature of aquatic systems. On the Upper Mississippi River System (UMRS), water depth data are important for describing the physical...

  16. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 8 Color Infrared

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  17. Bathymetric data for the Upper Mississippi and Illinois Rivers -- Pool 4

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — Water depth is an important feature of aquatic systems. On the Upper Mississippi River System (UMRS), water depth data are important for describing the physical...

  18. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 9

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  19. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 8

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  20. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 13 North

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  1. 2011 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 14

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  2. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 10

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  3. 2011 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 16

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  4. 2011 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 26

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  5. 2011 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 22

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  6. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 1

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  7. 2011 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 18

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  8. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 4 South

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  9. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 7

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  10. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 5

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  11. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 12

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  12. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 13 South

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  13. 2010 UMRS Color Infrared Aerial Photo Mosaic - Illinois River, Marseilles Pool

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  14. 2011 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 24

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  15. 2010 UMRS Color Infrared Aerial Photo Mosaic - Illinois River, Alton Pool

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  16. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 5a

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  17. 2010 UMRS Color Infrared Aerial Photo Mosaic - Illinois River, LaGrange Pool North

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  18. 2011 UMRS Color Infrared Aerial Photo Mosaic - Illinois River, Brandon Pool

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  19. 2010 UMRS Color Infrared Aerial Photo Mosaic - Illinois River, LaGrange Pool South

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  20. 2010 UMRS Color Infrared Aerial Photo Mosaic - Illinois River, Starved Rock Pool

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  1. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 11

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  2. 2011 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 20

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  3. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 3

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  4. 2011 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 25

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  5. 2011 UMRS Color Infrared Aerial Photo Mosaic - Illinois River, Dresden Pool

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  6. 2011 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 15

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  7. 2011 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 19

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  8. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 2

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  9. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 4 North

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  10. 2011 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 17

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  11. 2011 UMRS Color Infrared Aerial Photo Mosaic - Illinois River, Lockport Pool

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Upper Mississippi River Restoration-Environmental Management Program (UMRR-EMP) partnered with the U.S. Fish and Wildlife Service (USFWS) Region 3 to collect...

  12. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 5a

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) collects aerial photography of the Upper Mississippi River System (UMRS) floodplain...

  13. SMORN-III benchmark test on reactor noise analysis methods

    International Nuclear Information System (INIS)

    A computational benchmark test was performed in conjunction with the Third Specialists Meeting on Reactor Noise (SMORN-III) which was held in Tokyo, Japan in October 1981. This report summarizes the results of the test as well as the works made for preparation of the test. (author)

  14. HFR irradiation testing of light water reactor (LWR) fuel

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-09-01

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

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

    International Nuclear Information System (INIS)

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

  17. Reactor numerical simulation and hydraulic test research

    Energy Technology Data Exchange (ETDEWEB)

    Yang, L. S. [Nuclear Power Institute of China, Beijing (China)

    2009-07-01

    In recent years, the computer hardware was improved on the numerical simulation on flow field in the reactor. In our laboratory, we usually use the Pro/e or UG commercial software. After completed topology geometry, ICEM-CFD is used to get mesh for computation. Exact geometrical similarity is maintained between the main flow paths of the model and the prototype, with the exception of the core simulation design of the fuel assemblies. The drive line system is composed of drive mechanism, guide bush assembly, fuel assembly and control rod assembly, and fitted with the rod level indicator and drive mechanism power device.

  18. The reactor core analysis code CITATION-1000VP for High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Reactor core analysis with full core model has been necessary for the High Temperature Engineering Test Reactor (HTTR) design. The CITATION-1000VP code has been developed to enable reactor core analysis of HTTR with full core model through extending the number of zones and meshes, and enhancing the calculation speed of CITATION code. This report describes the program changes for extending the number of zones and meshes, and for vectorization. The maximum number of zones and meshes becomes 999 and 500, respectively. The calculation speed is enhanced up to 21 times. (author)

  19. Startup of the FFTF sodium cooled reactor. [Acceptance Test Program

    Energy Technology Data Exchange (ETDEWEB)

    Redekopp, R.D.; Umek, A.M.

    1981-03-01

    The Fast Flux Test Facility (FFTF), located on the Department of Energy (DOE) Hanford Reservation near Richland, Washington, is a 3 Loop 400 MW(t) sodium cooled fast reactor with a primary mission to test fuels and materials for development of the Liquid Metal Fast Breeder Reactor (LMFBR). Bringing FFTF to a condition to accomplish this mission is the goal of the Acceptance Test Program (ATP). This program was the mechanism for achieving startup of the FFTF. Highlights of the ATP involving the system inerting, liquid metal and inerted cell testing and initial ascent to full power are discussed.

  20. Testing Homeopathy in Mouse Emotional Response Models: Pooled Data Analysis of Two Series of Studies

    Directory of Open Access Journals (Sweden)

    Paolo Bellavite

    2012-01-01

    Full Text Available Two previous investigations were performed to assess the activity of Gelsemium sempervirens (Gelsemium s. in mice, using emotional response models. These two series are pooled and analysed here. Gelsemium s. in various homeopathic centesimal dilutions/dynamizations (4C, 5C, 7C, 9C, and 30C, a placebo (solvent vehicle, and the reference drugs diazepam (1 mg/kg body weight or buspirone (5 mg/kg body weight were delivered intraperitoneally to groups of albino CD1 mice, and their effects on animal behaviour were assessed by the light-dark (LD choice test and the open-field (OF exploration test. Up to 14 separate replications were carried out in fully blind and randomised conditions. Pooled analysis demonstrated highly significant effects of Gelsemium s. 5C, 7C, and 30C on the OF parameter “time spent in central area” and of Gelsemium s. 5C, 9C, and 30C on the LD parameters “time spent in lit area” and “number of light-dark transitions,” without any sedative action or adverse effects on locomotion. This pooled data analysis confirms and reinforces the evidence that Gelsemium s. regulates emotional responses and behaviour of laboratory mice in a nonlinear fashion with dilution/dynamization.

  1. The ''CAMERA'' test facility in the OSIRIS reactor

    International Nuclear Information System (INIS)

    CAMERA is an irradiation installation conceived to measure under neutronic flux and continuously the dimension variations of a fuel pencil of PWR reactors. The device, set in the periphery of the OSIRIS reactor, can receive new, preirradiated or reconstituted pencils. The principles of measurements is explained. Then, a brief description of the installation is given: in-pile part; out-of-pile part; connections. The technical characteristics of the installation are presented. A first qualification test of the installation under flux has been carried out at the end of the first semester 1984 in the OSIRIS reactor

  2. Fish communities in sandy pool of Magela Creek, Alligator Rivers Region

    International Nuclear Information System (INIS)

    Physico-chemical conditions, changes in fish communities and characteristics of species populations of eight permanent sandy pools along Magela Creek during the 1981 Dry season are described. Causes of mortality in each species, especially Craterocephalus marianae, were investigated. It is emphasised that in using baseline data to assess the impact of mining and animal communities, it may sometimes be difficult to differentiate natural mortality from mortality resulting from pollution. The aim of this study was to distinguish the most important environmental factors responsible for fish mortality. The study indicates that the mortality was low (<50% of the original population) in most pools. In populations that did suffer high mortality, anoxic conditions may have been an important cause. 67 refs., 36 tabs., 21 figs., ills

  3. Reactor fault simulation at the closure of the Windscale advanced gas-cooled reactor: analysis of reactor transient tests

    International Nuclear Information System (INIS)

    The testing of fault transient analysis methods by direct simulation of fault sequences on a commercial reactor is clearly excluded on safety and economic grounds. The closure of the Windscale prototype advanced gas-cooled reactor (WAGR) therefore offered a unique opportunity to test fault study methods under extreme conditions relatively unfettered by economic constraints, although subject to appropriate safety regulations. One aspect of these important experiments was a series of reactor transient tests. The objective of these reactor transients was to increase confidence in the fault study computer models used for commercial AGR safety assessment by extending their range of validation to cover large amplitude and fast transients in temperature, power and flow, relevant to CAGR faults, and well beyond the conditions achievable experimentally on commercial reactors. A large number of tests have now been simulated with the fault study code KINAGRAX. Agreement with measurement is very good and sensitivity studies show that such discrepancies as exist may be due largely to input data errors. It is concluded that KINAGRAX is able to predict steady state conditions and transient amplitudes in both power and temperature to within a few percent. (author)

  4. High-Temperature Gas-Cooled Test Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

    2016-04-01

    A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

  5. Entrained Flow Reactor Test of Potassium Capture by Kaolin

    DEFF Research Database (Denmark)

    Wang, Guoliang; Jensen, Peter Arendt; Wu, Hao;

    2015-01-01

    In the present study a method to simulate the reaction between gaseous KCl and kaolin at suspension fired condition was developed using a pilot-scale entrained flow reactor (EFR). Kaolin was injected into the EFR for primary test of this method. By adding kaolin, KCl can effectively be captured......-bed reactor. The method using the EFR developed in this study will be applied for further systematic investigation of different additives....

  6. River restoration success depends on the species pool of the immediate surroundings.

    Science.gov (United States)

    Sundermann, Andrea; Stoll, Stefan; Haase, Peter

    2011-09-01

    Previous studies evaluating the success of river restorations have rarely found any consistent effects on benthic invertebrate assemblages. In this study, we analyzed data from 24 river restoration projects in Germany dating back 1 to 12 years and 1231 data sets from adjacent river reaches that lie within 0-5, 5-10, and 10-15 km rings centered on the restored sites. We calculated restoration success and recolonization potential of adjacent river reaches based on stream-type-specific subsets of taxa indicative for good or bad habitat quality. On average, the restorations did not improve the benthic invertebrate community quality. However, we show that restoration success depends on the presence of source populations of desired taxa in the surrounding of restored sites. Only where source populations of additional desired taxa existed within a 0-5 km ring around the restored sites were benthic invertebrate assemblages improved by the restoration. Beyond the 5-km rings, this recolonization effect was no longer detected. We present here the first field results to support the debated argument that a lack of source populations in the areas surrounding restored sites may play an important role in the failure to establish desired invertebrate communities by the means of river restorations. In contrast, long-range dispersal of invertebrates seems to play a subordinate role in the recolonization of restored sites. However, because the surroundings of the restored sites were far from good ecological quality, the potential for improvement of restored sites was limited. PMID:21939037

  7. An isotope separation magnet for the injector test experiment (MITE)

    International Nuclear Information System (INIS)

    A magnet has been designed for space-charge neutralization studies on the Injector Test Experiment at the Chalk River Nuclear Laboratories. Augmented by suitable collectors, the magnet could also be used for pilot-scale isotope separations. The present report documents the design of this particular magnet and illustrates the process of designing beam transport magnets in general

  8. Surveillance tests for light-water cooled nuclear power reactor vessels in IMEF

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Yong-Sun; Ahn, Sang-Bok; Park, Dae-Gyu; Jung, Yang-Hong; Yoo, Byung-Ok; Oh, Wan-Ho; Baik, Seung-Je; Koo, Dae-Seo; Lee, Key-Soon [Korea Atomic Energy Research Institute, Irradiated Materials Examination Facility, Taejon (Korea)

    1999-09-01

    The surveillance tests for light-water cooled nuclear power reactor vessels were established to monitor the radiation-induced changes in the mechanical properties of ferritic materials in the beltline according to US NRC 10 CFR 50 App. G, US NRC RG1.99-rev.2, ASTM E185-82 and E185-94 in Irradiated Materials Examination Facility(IMEF). The surveillance capsule was transported from NPPs pool sites to KAERI IMEF by using a shipping cask. The capsule was cut and dismantled by capsule cutting machine and milling machine in M2 hot cell. Charpy tests and tension tests were performed in M5a and M5b hot cells respectively. Especially the EPMA located at hot lab was used to analyze the Ni and Cu wt% composition of base metal and weld for predicting the adjusted reference temperature(ART). The established process and test results were summarized in this paper. (author)

  9. An Experimental Test of Buffer Utility as a Technique for Managing Pool-Breeding Amphibians.

    Science.gov (United States)

    Powell, Jessica S Veysey; Babbitt, Kimberly J

    2015-01-01

    Vegetated buffers are used extensively to manage wetland-dependent wildlife. Despite widespread application, buffer utility has not been experimentally validated for most species. To address this gap, we conducted a six-year, landscape-scale experiment, testing how buffers of different widths affect the demographic structure of two amphibian species at 11 ephemeral pools in a working forest of the northeastern U.S. We randomly assigned each pool to one of three treatments (i.e., reference, 100m buffer, 30m buffer) and clearcut to create buffers. We captured all spotted salamanders and wood frogs breeding in each pool and examined the impacts of treatment and hydroperiod on breeding-population abundance, sex ratio, and recapture rate. The negative effects of clearcutting tended to increase as forest-buffer width decreased and be strongest for salamanders and when other stressors were present (e.g., at short-hydroperiod pools). Recapture rates were reduced in the 30m, but not 100m, treatment. Throughout the experiment for frogs, and during the first year post-cut for salamanders, the predicted mean proportion of recaptured adults in the 30m treatment was only 62% and 40%, respectively, of that in the reference treatment. Frog sex ratio and abundance did not differ across treatments, but salamander sex ratios were increasingly male-biased in both cut treatments. By the final year, there were on average, only about 40% and 65% as many females predicted in the 100m and 30m treatments, respectively, compared to the first year. Breeding salamanders at short-hydroperiod pools were about 10% as abundant in the 100m versus reference treatment. Our study demonstrates that buffers partially mitigate the impacts of habitat disturbance on wetland-dependent amphibians, but buffer width and hydroperiod critically mediate that process. We provide the first experimental evidence showing that 30-m-wide buffers may be insufficient for maintaining resilient breeding populations of pool

  10. An Experimental Test of Buffer Utility as a Technique for Managing Pool-Breeding Amphibians.

    Science.gov (United States)

    Powell, Jessica S Veysey; Babbitt, Kimberly J

    2015-01-01

    Vegetated buffers are used extensively to manage wetland-dependent wildlife. Despite widespread application, buffer utility has not been experimentally validated for most species. To address this gap, we conducted a six-year, landscape-scale experiment, testing how buffers of different widths affect the demographic structure of two amphibian species at 11 ephemeral pools in a working forest of the northeastern U.S. We randomly assigned each pool to one of three treatments (i.e., reference, 100m buffer, 30m buffer) and clearcut to create buffers. We captured all spotted salamanders and wood frogs breeding in each pool and examined the impacts of treatment and hydroperiod on breeding-population abundance, sex ratio, and recapture rate. The negative effects of clearcutting tended to increase as forest-buffer width decreased and be strongest for salamanders and when other stressors were present (e.g., at short-hydroperiod pools). Recapture rates were reduced in the 30m, but not 100m, treatment. Throughout the experiment for frogs, and during the first year post-cut for salamanders, the predicted mean proportion of recaptured adults in the 30m treatment was only 62% and 40%, respectively, of that in the reference treatment. Frog sex ratio and abundance did not differ across treatments, but salamander sex ratios were increasingly male-biased in both cut treatments. By the final year, there were on average, only about 40% and 65% as many females predicted in the 100m and 30m treatments, respectively, compared to the first year. Breeding salamanders at short-hydroperiod pools were about 10% as abundant in the 100m versus reference treatment. Our study demonstrates that buffers partially mitigate the impacts of habitat disturbance on wetland-dependent amphibians, but buffer width and hydroperiod critically mediate that process. We provide the first experimental evidence showing that 30-m-wide buffers may be insufficient for maintaining resilient breeding populations of pool

  11. Reactor coolant pump testing using motor current signatures analysis

    Energy Technology Data Exchange (ETDEWEB)

    Burstein, N.; Bellamy, J.

    1996-12-01

    This paper describes reactor coolant pump motor testing carried out at Florida Power Corporation`s Crystal River plant using Framatome Technologies` new EMPATH (Electric Motor Performance Analysis and Trending Hardware) system. EMPATH{trademark} uses an improved form of Motor Current Signature Analysis (MCSA), technology, originally developed at Oak Ridge National Laboratories, for detecting deterioration in the rotors of AC induction motors. Motor Current Signature Analysis (MCSA) is a monitoring tool for motor driven equipment that provides a non-intrusive means for detecting the presence of mechanical and electrical abnormalities in the motor and the driven equipment. The base technology was developed at the Oak Ridge National Laboratory as a means for determining the affects of aging and service wear specifically on motor-operated valves used in nuclear power plant safety systems, but it is applicable to a broad range of electric machinery. MCSA is based on the recognition that an electric motor (ac or dc) driving a mechanical load acts as an efficient and permanently available transducer by sensing mechanical load variations, large and small, long-term and rapid, and converting them into variations in the induced current generated in the motor windings. The motor current variations, resulting from changes in load caused by gears, pulleys, friction, bearings, and other conditions that may change over the life of the motor, are carried by the electrical cables powering the motor and are extracted at any convenient location along the motor lead. These variations modulate the 60 Hz carrier frequency and appear as sidebands in the spectral plot.

  12. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris{o} National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately.

  13. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris oe National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately

  14. Radiation Resistance Test of Wireless Sensor Node and the Radiation Shielding Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Li, Liqan; Sur, Bhaskar [Atomic Energy of Canada Limited, Ontario (Canada); Wang, Quan [University of Western Ontario, Ontario (Canada); Deng, Changjian [The University of Electronic Science and Technology, Chengdu (China); Chen, Dongyi; Jiang, Jin [Applied Physics Branch, Ontario (Korea, Republic of)

    2014-08-15

    A wireless sensor network (WSN) is being developed for nuclear power plants. Amongst others, ionizing radiation resistance is one essential requirement for WSN to be successful. This paper documents the work done in Chalk River Laboratories of Atomic Energy of Canada Limited (AECL) to test the resistance to neutron and gamma radiation of some WSN nodes. The recorded dose limit that the nodes can withstand before being damaged by the radiation is compared with the radiation environment inside a typical CANDU (CANada Deuterium Uranium) power plant reactor building. Shielding effects of polyethylene, cadmium and lead to neutron and gamma radiations are also analyzed using MCNP simulation. The shielding calculation can be a reference for the node case design when high dose rate or accidental condition (like Fukushima) is to be considered.

  15. Inductive testing of reactor pressure vessels

    International Nuclear Information System (INIS)

    In Service Inspection of Reactor Pressure Vessels is mostly done with ultrasonics. Using special 2 crystal-probes good detectability is achieved for near surface defects. The problem is to detect closely spaced cracks, to decide if the defects are surface braking and, if not, to decide the remaining ligament. The purpose of this study is to investigate to what extent Eddy Current can solve these problems. Detecting surfacebreaking cracks and fields of cracks can be done using conventional Eddy Current techniques. Mapping of closely spaced cracks requires a small probe and a high frequency. Measurement of depths a larger probe, a lower frequency and knowledge of the crackfield since 2 closely spaced shallow cracks might be mistaken for one deep crack. Depths of singel cracks can be measured down to 7-8 mm. In closely spaced crackfields the depths can not be measured. The measurement is mostly based on amplitude. For not surface breaking defects the problem is to decide the ligament, i.e. the distance between surface and cracktip. To achieve good penetration a large probe, low frequency and high energy or pulsed energy is used. Ligament up to 4 mm can be measured with good accuracy. The measurements is mostly based on phase. Noise, which originates from rough surface, varied material structure and lift off, can be reduced using multi frequency mix, probe design and scanning pattern. (author)

  16. Restart of K-Reactor, Savannah River Site: Safety evaluation report

    International Nuclear Information System (INIS)

    This Safety Evaluation Report (SER) focuses on those issues required to support the restart of the K-Reactor at the Savannah River Plant. This SER provides the safety criteria for restart and documents the results of the staff reviews of the DOE and operating contractor activities to meet these criteria. To develop the restart criteria for the issues discussed in this SER, the Savannah River Restart Office and Savannah River Special Projects Office staffs relied, when possible, on commercial industry codes and standards and on NRC requirements and guidelines for the commercial nuclear industry. However, because of the age and uniqueness of the Savannah River reactors, criteria for the commercial plants were not always applicable. In these cases, alternate criteria were developed. The restart criteria applicable to each of the issues are identified in the safety evaluations for each issue. The restart criteria identified in this report are intended to apply only to restart of the Savannah River reactors. Following the development of the acceptance criteria, the DOE staff and their support contractors evaluated the results of the DOE and operating contractor (WSRC) activities to meet these criteria. The results of those evaluations are documented in this report. Deviations or failures to meet the requirements are either justified in the report or carried as open or confirmatory items to be completed and evaluated in supplements to this report before restart. 62 refs., 1 fig

  17. Restart of K-Reactor, Savannah River Site: Safety evaluation report

    Energy Technology Data Exchange (ETDEWEB)

    1991-04-01

    This Safety Evaluation Report (SER) focuses on those issues required to support the restart of the K-Reactor at the Savannah River Plant. This SER provides the safety criteria for restart and documents the results of the staff reviews of the DOE and operating contractor activities to meet these criteria. To develop the restart criteria for the issues discussed in this SER, the Savannah River Restart Office and Savannah River Special Projects Office staffs relied, when possible, on commercial industry codes and standards and on NRC requirements and guidelines for the commercial nuclear industry. However, because of the age and uniqueness of the Savannah River reactors, criteria for the commercial plants were not always applicable. In these cases, alternate criteria were developed. The restart criteria applicable to each of the issues are identified in the safety evaluations for each issue. The restart criteria identified in this report are intended to apply only to restart of the Savannah River reactors. Following the development of the acceptance criteria, the DOE staff and their support contractors evaluated the results of the DOE and operating contractor (WSRC) activities to meet these criteria. The results of those evaluations are documented in this report. Deviations or failures to meet the requirements are either justified in the report or carried as open or confirmatory items to be completed and evaluated in supplements to this report before restart. 62 refs., 1 fig.

  18. Clinch River Breeder Reactor environmental effects: general water-side corrosion

    International Nuclear Information System (INIS)

    Studies are described of the general corrosion of 21/4 Cr--1 Mo steel in pure superheated steam, in impure superheated and saturated steam, and under nucleate boiling conditions. The test parameters were selected to provide information relevant to the use of this steel for the Clinch River Breeder Reactor superheaters and evaporators. The oxidation rate of 21/4 Cr--1 Mo steel in superheated steam was measured under heat transfer conditions at 510 to 5400C (950 to 10050F), and was approximately 11/2 times that measured under isothermal conditions. Extensive general attack of stressed 21/4 Cr--1 Mo steel specimens occurred in cyclic tests in superheated and saturated steam with chloride and oxygen additions, although no cracking or localized attack was observed. Considerably less attack occurred under superheat conditions or in the absence of oxygen. Tests under nucleate boiling conditions were operated to evaluate crevice effects associated with porous films on heat transfer surfaces. Significant crevice corrosion was produced in water containing 10 ppm chloride; a heavier but more general attack occurred in treated cooling tower water

  19. Design of the Advanced Gas Reactor Fuel Experiments for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2005-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight particle fuel tests in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL) to support development of the next generation Very High Temperature Reactor (VHTR) in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments will be irradiated in an inert sweep gas atmosphere with on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The final design phase has just been completed on the first experiment (AGR-1) in this series and the support systems and fission product monitoring system that will monitor and control the experiment during irradiation. This paper discusses the development of the experimental hardware and support system designs and the status of the experiment.

  20. Decommissioning and dismantling of 305-M test pile at the Savannah River Plant

    International Nuclear Information System (INIS)

    The 305-M Test Pile was started up at the Savannah River Plant in 1952 and operated until 1981. The pile was used to measure the uranium content of reactor fuel. In 1984 work began to decommission and dismantle the pile. Extensive procedures were used that included a detailed description of the radiological controls and safety measures. These controls allowed the job to be completed with radiation doses as low as reasonably achievable

  1. Vibration tests on some models of PEC reactor core elements

    International Nuclear Information System (INIS)

    This paper describes the aims of the experimental tests carried out at ISMES, within an agreement with the Department of Fast Reactors of ENEA, on some models of the elements of PEC Fast Nuclear Reactor Core in the frame of the activities for the seismic verification of the PEC core. The seismic verification is briefly described with particular attention to the problems arising from the shocks among the various elements during an earthquake, as well as the computer code used, the purpose and the techniques used to perform tests, some results and the first comparison between the theory and the experimental data

  2. Sodium reflux pool-boiler solar receiver on-sun test results

    Energy Technology Data Exchange (ETDEWEB)

    Andraka, C E; Moreno, J B; Diver, R B; Moss, T A [Oak Ridge National Lab., TN (United States)

    1992-06-01

    The efficient operation of a Stirling engine requires the application of a high heat flux to the relatively small area occupied by the heater head tubes. Previous attempts to couple solar energy to Stirling engines generally involved directly illuminating the heater head tubes with concentrated sunlight. In this study, operation of a 75-kW{sub t} sodium reflux pool-boiler solar receiver has been demonstrated and its performance characterized on Sandia's nominal 75-kW{sub t} parabolic-dish concentrator, using a cold-water gas-gap calorimeter to simulate Stirling engine operation. The pool boiler (and more generally liquid-metal reflux receivers) supplies heat to the engine in the form of latent heat released from condensation of the metal vapor on the heater head tubes. The advantages of the pool boiler include uniform tube temperature, leading to longer life and higher temperature available to the engine, and decoupling of the design of the solar absorber from the engine heater head. The two-phase system allows high input thermal flux, reducing the receiver size and losses, therefore improving system efficiency. The receiver thermal efficiency was about 90% when operated at full power and 800{degree}C. Stable sodium boiling was promoted by the addition of 35 equally spaced artificial cavities in the wetted absorber surface. High incipient boiling superheats following cloud transients were suppressed passively by the addition of small amounts of xenon gas to the receiver volume. Stable boiling without excessive incipient boiling superheats was observed under all operating conditions. The receiver developed a leak during performance evaluation, terminating the testing after accumulating about 50 hours on sun. The receiver design is reported here along with test results including transient operations, steady-state performance evaluation, operation at various temperatures, infrared thermography, x-ray studies of the boiling behavior, and a postmortem analysis.

  3. Sodium reflux pool-boiler solar receiver on-sun test results

    Science.gov (United States)

    Andraka, C. E.; Moreno, J. B.; Diver, R. B.; Moss, T. A.

    1992-06-01

    The efficient operation of a Stirling engine requires the application of a high heat flux to the relatively small area occupied by the heater head tubes. Previous attempts to couple solar energy to Stirling engines generally involved directly illuminating the heater head tubes with concentrated sunlight. In this study, operation of a 75-kW(sub t) sodium reflux pool-boiler solar receiver has been demonstrated and its performance characterized on Sandia's nominal 75-kW(sub t) parabolic-dish concentrator, using a cold-water gas-gap calorimeter to simulate Stirling engine operation. The pool boiler (and more generally liquid-metal reflux receivers) supplies heat to the engine in the form of latent heat released from condensation of the metal vapor on the heater head tubes. The advantages of the pool boiler include uniform tube temperature, leading to longer life and higher temperature available to the engine, and decoupling of the design of the solar absorber from the engine heater head. The two-phase system allows high input thermal flux, reducing the receiver size and losses, therefore improving system efficiency. The receiver thermal efficiency was about 90 percent when operated at full power and 800 C. Stable sodium boiling was promoted by the addition of 35 equally spaced artificial cavities in the wetted absorber surface. High incipient boiling superheats following cloud transients were suppressed passively by the addition of small amounts of xenon gas to the receiver volume. Stable boiling without excessive incipient boiling superheats was observed under all operating conditions. The receiver developed a leak during performance evaluation, terminating the testing after accumulating about 50 hours on sun. The receiver design is reported here along with test results including transient operations, steady-state performance evaluation, operation at various temperatures, infrared thermography, x-ray studies of the boiling behavior, and a postmortem analysis.

  4. Chalk Formations as Natural Barriers towards Radionuclide Migration

    DEFF Research Database (Denmark)

    Pedersen, Walther Batsberg; Carlsen, Lars; Jensen, Bror Skytte

    1985-01-01

    A series of chalk samples from the cretaceous formation overlying the Erslev salt dome have been studied in order to establish permeabilities, porosities, dispersion-, diffusion-, and sorption characteristics of the chalk. The chalk was found to be porous (∊≈0.4), however, of rather low permeabil...

  5. Microdeformation and subcritical cracking in chalk

    Science.gov (United States)

    Bergsaker, Anne; Dysthe, Dag Kristian

    2016-04-01

    Deformation processes in chalks, both in relation to changing pore fluids and stress conditions has been of great interest as chalk is an important reservoir rock for both hydrocarbons and ground water. Lately it has also gained interest as a potential reservoir rock for captured CO2. Chalks are composed of large amounts of biogenic calcite grains, the skeletal debris of marine microorganisms. Its deformation is highly time and stress dependent, and governed by a transition from distributed to localized deformation at the onset of yield, affected by mechanisms such as subcritical crack growth and pore collapse. We present a microdeformation rig which makes use of thermal expansion as a means of subjecting small samples to strictly controlled tensile stresses. High resolution imaging provides resolutions down to 0.5 micrometers, enabling study of pore scale processes during slow deformation. Examples of localized and distributed deformation are presented.

  6. Elastic behaviour of North Sea chalk

    DEFF Research Database (Denmark)

    Gommesen, Lars; Fabricius, Ida Lykke; Mukerji, T.;

    2007-01-01

    -consistent approximation, which here represents the unrelaxed scenario where the pore spaces of the rock are assumed to be isolated, and the Gassmann theory, which assumes that pore spaces are connected, as tools for predicting the effect of hydrocarbons from the elastic properties of brine-saturated North Sea reservoir......We present two different elastic models for, respectively, cemented and uncemented North Sea chalk well-log data. We find that low Biot coefficients correlate with anomalously low cementation factors from resistivity measurements at low porosity and we interpret this as an indication of cementation...... chalk. In the acoustic impedance–Poisson's ratio plane, we forecast variations in porosity and hydrocarbon saturation from their influence on the elastic behaviour of the chalk. The Gassmann model and the self-consistent approximation give roughly similar predictions of the effect of fluid on acoustic...

  7. Managing Injected Water Composition To Improve Oil Recovery: A Case Study of North Sea Chalk Reservoirs

    DEFF Research Database (Denmark)

    Zahid, Adeel; Shapiro, Alexander; Stenby, Erling Halfdan;

    2012-01-01

    In recent years, many core displacement experiments of oil by seawater performed on chalk rock samples have reported SO42–, Ca2+, and Mg2+ as potential determining ions for improving oil recovery. Most of these studies were carried out with outcrop chalk core plugs. The objective of this study...... imbibition, which has been applied in most of the previous studies. Two different flooding schemes (with and without aging) were used for flooding North Sea reservoir chalk samples. For comparison, two tests were also carried out with Stevns Klint core plugs. The flooding tests were carried out...... with the following injecting fluids: distilled water, brine with and without sulfate, and brine containing only magnesium ions. The total oil recovery, recovery rate, and interaction mechanisms of ions with rock were studied for different injecting fluids at different temperatures and wettability conditions. Studies...

  8. Improving the proliferation resistance of research and test reactors

    International Nuclear Information System (INIS)

    Elimination, or substantial reduction, of the trade in unirradiated highly-enriched fuel elements for research and test reactors would significantly reduce the proliferation risk associated with the current potential for diversion of these materials. To this end, it is the long-term goal of U.S. policy to fuel all new and existing research and test reactors with uranium of less-than-20% enrichment (but substantially greater than natural) excepting, perhaps, only a small number of high-power, high-performance, reactors. The U.S. development program for enrichment reduction in research and test reactor designs currently using 90-93% enriched uranium is based on the practical criterion that enrichment reduction should not cause significant flux performance (flux per unit power) or burnup performance degradation relative to the unmodified reactor design. To first order, this implies the requirement that the 235U loading in the reduced-enrichment fuel elements be the same as the 235U loading in the 90-93% enriched fuel elements. This can be accomplished by substitution of higher uranium density fuel technology for currently-used fuel technology in the fuel meat volume of the current fuel element design and/or by increasing the usable fuel meat volume. For research and test reactors of power greater than 5-10 megawatts, fuel technology does not currently exist that would permit enrichment reductions to below 20% utilizing this criterion. A program is now beginning in the U.S. to develop the necessary fuel technology. Currently-proven fuel technology is capable, however, of accommodating enrichment reductions to the 30-45% range (from 90-93%) for many reactors in the 5-50MW range. Accordingly the U.S. is proposing to convert existing reactors (and new designs) in the 5-50MW range from the use of highly-enriched fuel to the use of 30-45% enriched fuel, and reactors of less that about 5MW to less-than-20% enrichment, wherever this can be done without significant performance

  9. Physics calculations for the Clinch River Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kalimullah; Kier, P.H.; Hummel, H.H.

    1977-06-01

    Calculations of distributions of power and sodium void reactivity, unvoided and voided Doppler coefficients and steel and fuel worths have been performed using diffusion theory and first-order perturbation theory for the LWR discharge Pu-fueled CRBR at BOL, the FFTF-grade Pu-fueled CRBR at BOL and for the beginning and end of equilibrium cycle of the LWR-Pu-fueled CRBR. The results of the burnup and breeding ratio calculations performed for obtaining the reactor compositions during the equilibrium cycle are also reported. Effects of sodium and steel contents on the distributions of sodium void reactivity and steel worth have also been studied. Errors and uncertainties in the reactivity coefficients due to cross-sections and the two-dimensional geometric representations of the reactor used in the calculations have also been estimated. Comparisons of the results with those in the CRBR PSAR are also discussed.

  10. Lessons learned from the licensing process for the Clinch River Breeder Reactor Plant

    International Nuclear Information System (INIS)

    This paper presents the experience of licensing a specific liquid-metal fast breeder reactor (LMFBR), the Clinch River Breader Reactor Plant (CRBRP). It was a success story in that the licensing process was accomplished in a very short time span. The actions of the applicant and the actions of the US Nuclear Regulatory Commission (NRC) in response are presented and discussed to provide guidance to future efforts to license unconventional reactors. The history is told from the perspective of the authors. As such, some of the reasons given for success or lack of success are subjective interpretations. Nevertheless, the authors' positions provided them an excellent viewpoint to make these judgements. During the second phase of the licensing process, they were the CRBRP Technical Director and the Licensing Manager, respectively, for the Westinghouse Electric Corporation, the prime contractor for the reactor plant

  11. The technology development for surveillance test of reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Sun Phil; Park, Day Young; Choi, Kwen Jai

    1997-12-01

    Benchmark test was performed in accordance with the requirement of US NRC Reg. Guide DG-1053 for Kori unit-1 in order to determine best-estimated fast neutron fluence irradiated into reactor vessel. Since the uncertainty of radiation analysis comes from the calculation error due to neutron cross-section data, reactor core geometrical dimension, core source, mesh density, angular expansion and convergence criteria, evaluation of calculational uncertainty due to analytical method was performed in accordance with the regulatory guide and the proof was performed for entire analysis by comparing the measurement value obtained by neutron dosimetry located in surveillance capsule. Best-estimated neutron fluence in reactor vessel was calculated by bias factor, neutron flux measurement value/calculational value, from reanalysis result from previous 1st through 4th surveillance testing and finally fluence prediction was performed for the end of reactor life and the entire period of plant life extension. Pressurized thermal shock analysis was performed in accordance with 10 CFR 50.61 using the result of neutron fluence analysis in order to predict the life of reactor vessel material and the criteria of safe operation for Kori unit 1 was reestablished. (author). 55 refs., 55 figs.

  12. Aerosol behavior during sodium pool fires in a large vessel: CSTF tests AB1 and AB2

    International Nuclear Information System (INIS)

    Two large-scale aerosol behavior tests, using sodium pool fires as the aerosol source, were performed in the Containment Systems Test Facility (CSTF). The tests were conducted to characterize the properties and behavior of sodium aerosol particles formed and aged in a large containment vessel. The 20-m high, 850-m3 CSTF containment building in regard to parameters that affect agglomeration and gravitational settling. In both tests, sodium burned for one hour in a 4.38-m2 pool, and the only difference between them was that steam was injected during the second test, simulating the release of water vapor from heated concrete

  13. In situ tests on the PEC fast reactor building

    International Nuclear Information System (INIS)

    This paper describes forced excitation tests carried out at the PEC reactor building, to determine seismic motion amplifications produced in the building itself. Experimental results are used to gauge numerical methodologies capable of assessing the margins existing in the design analysis. (orig./HP)

  14. Fuel and core testing plan for a target fueled isotope production reactor

    International Nuclear Information System (INIS)

    In recent years there has been an unstable supply of the critical diagnostic medical isotope 99Tc. Several concepts and designs have been proposed to produce 99Mo the parent nuclide of 99Tc, at a commercial scale sufficient to stabilize the world supply. This work lays out a testing and experiment plan for a proposed 2 MW open pool reactor fueled by Low Enriched Uranium (LEU) 99Mo targets. The experiments and tests necessary to support licensing of the reactor design are described and how these experiments and tests will help establish the safe operating envelop for a medical isotope production reactor is discussed. The experiments and tests will facilitate a focused and efficient licensing process in order to bring on line a needed production reactor dedicated to supplying medical isotopes. The Target Fuel Isotope Reactor (TFIR) design calls for an active core region that is approximately 40 cm in diameter and 40 cm in fuel height. It contains up to 150 cylindrical, 1-cm diameter, LEU oxide fuel pins clad with Zircaloy (zirconium alloy), in an annular hexagonal array on a ∼2.0 cm pitch surrounded, radially, by a graphite or a Be reflector. The reactor is similar to U.S. university reactors in power, hardware, and safety/control systems. Fuel/target pin fabrication is based on existing light water reactor fuel fabrication processes. However, as part of licensing process, experiments must be conducted to confirm analytical predictions of steady-state power and accident conditions. The experiment and test plan will be conducted in phases and will utilize existing facilities at the U.S. Department of Energy's Sandia National Laboratories. The first phase is to validate the predicted reactor core neutronics at delayed critical, zero power and very low power. This will be accomplished by using the Sandia Critical Experiment (CX) platform. A full scale TFIR core will be built in the CX and delayed critical measurements will be taken. For low power experiments, fuel

  15. Research on reactor physics using the Japan Materials Testing Reactor Critical Facility (JMTRC)

    International Nuclear Information System (INIS)

    The JMTRC of 100 W was installed for the purpose of carrying out the basic experiment on the nuclear characteristics of reactors and the preceding test related to the operation plan of the Japan material testing reactor (JMTR, 50 MW). After the attainment of the initial criticality in October, 1965, for obtaining the reactor physics characteristics, criticality experiment was begun. The items of the criticality experiment were critical mass, control rod worth, reactor dynamic characteristic parameters, shutdown margin and so on, and these experimental data were effectively utilized for the safety evaluation in the operation of the JMTR. The preceding test using the JMTRC has been carried out for obtaining the nuclear characteristics of samples and the thermal characteristics estimated from those results by simulating the JMTR core. In August, 1983, the degree of fuel enrichment for the JMTRC was reduced to 45 % U-235, and various experiments usig the MEU core were carried out. In this paper, the criticality experiment using the MEU core and the experiment on the characteristics of lithium-containing pellets are reported. (K.I.)

  16. Static and dynamic effective stress coefficient of chalk

    DEFF Research Database (Denmark)

    Alam, M. Monzurul; Fabricius, Ida Lykke; Christensen, Helle Foged

    2012-01-01

    stress coefficient is thus relevant for studying reservoir deformation and for evaluating 4D seismic for the correct pore pressure prediction. The static effective stress coefficient n is estimated from mechanical tests and is highly relevant for effective stress prediction because it is directly related...... to mechanical strain in the elastic stress regime. The corresponding dynamic effective stress coefficient α is easy to estimate from density and velocity of acoustic (elastic) waves. We studied n and α of chalk from the reservoir zone of the Valhall field, North Sea, and found that n and α vary...

  17. Quality assurance in technology development for The Clinch River Breeder Reactor Plant Project

    International Nuclear Information System (INIS)

    The Clinch River Breeder Reactor Plant Project is the nation's first large-scale demonstration of the Liquid Metal Fast Breeder Reactor (LMFBR) concept. The Project has established an overall program of plans and actions to assure that the plant will perform as required. The program has been established and is being implemented in accordance with Department of Energy Standard RDT F 2-2. It is being applied to all parts of the plant, including the development of technology supporting its design and licensing activity. A discussion of the program as it is applied to development is presented

  18. An empirical test of freshwater vicariance via river capture.

    Science.gov (United States)

    Burridge, Christopher P; Craw, Dave; Waters, Jonathan M

    2007-05-01

    River capture is a geomorphological process through which stream sections are displaced from one catchment to another, and it may represent a dominant facilitator of interdrainage transfer and cladogenesis in freshwater-limited taxa. However, few studies have been conducted in a manner to explicitly test the biological significance of river capture. Here we present a multispecies phylogeographical analysis to test whether the nonmigratory fish fauna of the Von River (South Island, New Zealand) is the product of a well-documented, Late Quaternary capture of a section of the Oreti River (Southland drainage). Specifically, we predict that nonmigratory fishes of the Von River will exhibit closer genetic affinities with those of Southland, rather than those of the Clutha system, into which the Von River presently drains. Mitochondrial DNA phylogeography (control region and cytochrome b sequence data) and analysis of nuclear orthologues of mtDNA sequences indicate that 'flathead'Galaxias of the Von River (n = 31, three sites) have greatest genetic affinities with those of Southland (Galaxias 'southern', n = 216, 38 sites), rather than with those of the Clutha River (Galaxias sp. 'D', n = 73, 32 sites). Likewise, Von River 'roundhead'Galaxias (n = 52, four sites) have greatest genetic affinities with those of Southland drainages (Galaxias gollumoides, n = 223, 58 sites), rather than with those of the Clutha River (Galaxias pullus, Galaxias anomalus, Galaxias gollumoides of the Nevis tributary; n = 68, 32 sites). These findings are consistent with our predictions that genetic affinities of the nonmigratory fish fauna in the Von River would reflect past, rather than present, drainage connections. Consequently, river capture is responsible for the nonmigratory fish fauna of the Von River. In a broader context, river capture has frequently influenced the distribution of genetic lineages among catchments in New Zealand freshwater-limited fish, and its biogeographical

  19. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1988-05-01

    The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm/sup 3/ was by then in routine use, illustrated how far work has progressed.

  20. Standard Guide for Benchmark Testing of Light Water Reactor Calculations

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with ...

  1. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm3 was by then in routine use, illustrated how far work has progressed

  2. Present status and future perspective of research and test reactors in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Baba, Osamu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Kaieda, Keisuke

    1999-08-01

    Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfil a major role in the study of nuclear energy and fundamental research. At present, four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR), are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has reached first criticality and is waiting for the power-up test. This paper introduce these reactors and describe their present operational status. The recent tendency of utilization and future perspectives are also reported. (author)

  3. In-reactor experiments in fast breeder test reactor for assessment of core structural materials

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, India is a sodium cooled reactor with neutron flux level of the order of 1015 n/cm2/s and temperature of coolant in the range of 650-790K (380-520oC). This reactor is being used as a test bed for the development of fuel and structural materials required for Indian Fast Reactor Programme. FBTR is also used as a test facility to carry out accelerated irradiation tests on thermal reactor structural materials. In-reactor experiments on core structural materials are being carried out by subjecting prefabricated specimens to desired conditions of temperature and neutron fluence levels in FBTR. Non-instrumented irradiation capsules that can be loaded at any location of FBTR core are used for the experiments. Pressurised capsules of zirconium alloys have been developed and subjected to irradiation in FBTR to determine the irradiation creep rate of indigenously developed zirconium alloys (Zircaloy-2 and Zr-2.5%Nb alloy) for life assessment of pressure tubes of Indian Pressurised Heavy Water Reactors (PHWRs). Technology development of pressurised capsules was carried out at IGCAR. These pressurised capsules were filled with argon and a small fraction of helium at a high pressure (5.0-6.5 MPa at room temperature) in such a way that the target stresses were attained in the walls of the pressurised capsules at the desired temperature of irradiation in the reactor. FBTR was operated at a low power of 8 MWt during this irradiation campaign to have an inlet temperature of about 579 K (306oC) which was close to the temperature of pressure tubes at full power in PHWR. Irradiation of thirty pressurised capsules was carried out in FBTR using six irradiation capsules for different durations (upto 79 days). The fluence levels attained by the pressurised capsules were up to 1.1 x 1021 n/cm2 (E> 1 MeV) at temperatures of 579 to 592 K. Post-irradiation increase in diameter of the pressurised

  4. ASME N510 test results for Savannah River Site AACS filter compartments

    Energy Technology Data Exchange (ETDEWEB)

    Paul, J.D.; Punch, T.M. [Westinghouse Savannah River Company, Aiken, SC (United States)

    1995-02-01

    The K-Reactor at the Savannah River Site recently implemented design improvements for the Airborne Activity Confinement System (AACS) by procuring, installing, and testing new Air Cleaning Units, or filter compartments, to ASME AG-11, N509, and N510 requirements. Specifically, these new units provide documentable seismic resistance to a Design Basis Accident earthquake, provide 2 inch adsorber beds with 0.25 second residence time, and meet all AG-1, N509, and N510 requirements for testability and maintainability. This paper presents the results of the Site acceptance testing and discusses an issue associated with sample manifold qualification testing.

  5. The decommissioning of the KEMA suspension test reactor

    International Nuclear Information System (INIS)

    In this report the decommissioning of the KEMA Suspension Test Reactor (KSTR) is described. This reactor was a 1 MWth aqueous homo-geneous nuclear reactor in which a suspension of a mixed oxide UO2/ ThO2 in light water was circulated in a closed loop through a sphere-shaped core vessel. The reactor, located on KEMA premises, made 150 MW of heat during its critical periods. Dismantling of this reactor, with its many connected subsystems, meant the mastering of activated components which were also contaminated on inner surfaces caused by small fuel deposits (alpha contaminants) and fission products (beta, gamma contaminants). A description is given of the save removal of the fuel, the remote dismantling of systems and components and the disposal of steel scrap and other materials. Important features are the measures to be taken and provisions needed for safe handling, for the reduction of the radiation dose for the working team and the prevention of spreading of activity over the working area and the environment. It has been demonstrated that safe dismantling and disposal of such systems can be achieved. Experience gained at KEMA for the proper dismantling and for safety measures to be taken for workers and the environment can be made available for similar dismantling projects. A cost break-down is included in the report. (author). 22 refs.; 52 figs.; 12 tabs

  6. Materials qualification testing for next generation nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hurst, R.; Haehner, P. (European Commission, JRC Institute for Energy, Petten (Netherlands))

    2010-05-15

    The development of next generation, innovative nuclear fission reactors, needed to replace or supplement the current designs of nuclear reactors within the next say 30 years, critically depends on the availability of advanced structural and functional materials systems which must withstand extreme conditions: intense neuron irradiation, high temperatures, and potentially strongly corrosive coolant environments, in combination with complex loading states and cyclic loading histories. The mechanical performance and reliability of those materials depends on the service and off-normal conditions in whichever of the six candidate systems for Generation IV reactors, under the global Generation IV International Forum (GIF) agreement, they will be applied. This paper gives an overview of the suite of six selected reactor systems indicating where research on materials and structural integrity is still needed. Some of these reactor systems have been under study for many years whereas others are relatively new concepts but all still require a major expenditure of effort before they can be considered as realistic contenders. In particular the materials selection and component integrity for service will play a major role in a final successful design. Specific issues include: the endurance and stability with respect to creep, fatigue and fracture mechanics loading, the need for in situ environmental testing versus pre-exposure of materials and advanced structural-functional materials systems for specific applications. Using examples taken from research projects in which the authors' laboratory has participated, the materials qualification high temperature testing for three crucial components, reactor pressure vessel and piping, gas turbines and heat exchangers is described in some detail. Finally pointers are derived as to not only the scale of the remaining research needs but also the mechanisms which are planned to be followed in Europe, not to mention globally, to obtain

  7. Late Maastrichtian chalk mounds, Stevns Klint, Denmark

    DEFF Research Database (Denmark)

    Anderskouv, Kresten; Damholt, Tove; Surlyk, Finn

    2007-01-01

    Upper Maastrichtian chalk exposed at the Sigerslev quarry, Stevns Klint, Denmark is characterized by wavy and mound-like bedding geometries outlined by bands of black flint nodules. Four morphological elements are recognized, although bedding geometries are highly variable: southward migrating...

  8. Late Maastrichtian chalk mounds, Stevns Klint, Denmark

    DEFF Research Database (Denmark)

    Anderskouv, Kresten; Damholt, Tove; Surlyk, Finn

    Upper Maastrichtian chalk exposed in the Sigerslev quarry, Stevns Klint, Denmark shows wavy and mound-like bedding geometries outlined by bands of black flint nodules. Bedding geometries are highly variable, but four morphological elements are recognized: Southward migrating mounds, eastward...

  9. Slope failure of chalk channel margins

    DEFF Research Database (Denmark)

    Gale, A.; Anderskouv, Kresten; Surlyk, Finn;

    2015-01-01

    The importance of mass transport and bottom currents is now widely recognized in the Upper Cretaceous Chalk Group of Northern Europe. The detailed dynamics and interaction of the two phenomena are difficult to study as most evidence is based on seismic data and drill core. Here, field observation...

  10. Electroosmotic dewatering of chalk sludge, iron hydroxide sludge, wet fly ash and biomass sludge

    DEFF Research Database (Denmark)

    Hansen, H.K.; Christensen, Iben Vernegren; Ottosen, Lisbeth M.;

    2003-01-01

    Electroosmotic dewatering has been tested in laboratory cells on four different porous materials: chalk sludge, iron hydroxide sludge, wet fly ash and biomass sludge from enzyme production. In all cases it was possible to remove water when passing electric DC current through the material....... Casagrande's coefficients were determined for the four materials at different water contents. The experiments in this work showed that chalk could be dewatered from 40% to 79% DM (dry matter), fly ash from 75 to 82% DM, iron hydroxide sludge from 2.7 to 19% DM and biomass from 3 to 33% DM by electroosmosis...

  11. Strength and Biot's coefficient for high-porosity oil- or water-saturated chalk

    DEFF Research Database (Denmark)

    Andreassen, Katrine Alling

    . The Biot coefficient states the degree of cementation or how the pore pressure contributes to the strain resulting from an external load for a porous material. It is here calculated from dynamic measurements and correlated with the strength of outcrop chalk characterized by the onset of pore collapse...... during hydrostatic loading. The hypothesis is that the Biot coefficient and the theory of poroelasticity may cover the fluid effect by including the increased fluid bulk modulus from oil to water. A high number of test results for both oil- and water-saturated high-porosity outcrop chalk show correlation...

  12. Integrated leak rate test results of JOYO reactor containment vessel

    International Nuclear Information System (INIS)

    Integrated leak rate tests of JOYO after the reactor coolant system had been filled with sodium have been performed two times since 1978 (February 1978 and December 1979). The tests were conducted with the in-containment sodium systems, primary argon cover gas system and air conditioning systems operating. Both the absolute pressure method and the reference chamber method were employed during the test. The results of both tests confirmed the functioning of the containment vessel, and leak rate limits were satisfied. In Addition, the adequancy of the test instrumentation system and the test method was demonstrated. Finally the plant conditions required to maintain reasonable accuracy for the leak rate testing of LMFBR were established. In this paper, the test conditions and the test results are described. (author)

  13. Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

    Science.gov (United States)

    Bragg-Sitton, Shannon M.

    2005-01-01

    The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

  14. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jennifer Lyons; Wade R. Marcum; Mark D. DeHart; Sean R. Morrell

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  15. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  16. Design of high temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Construction of High Temperature Engineering Test Reactor (HTTR) is now underway to establish and upgrade basic technologies for HTGRs and to conduct innovative basic research at high temperatures. The HTTR is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 850degC for rated operation and 950degC for high temperature test operation. It is planned to conduct various irradiation tests for fuels and materials, safety demonstration tests and nuclear heat application tests. JAERI received construction permit of HTTR reactor facility in February 1990 after 22 months of safety review. This report summarizes evaluation of nuclear and thermal-hydraulic characteristics, design outline of major systems and components, and also includes relating R and D result and safety evaluation. Criteria for judgment, selection of postulated events, major analytical conditions for anticipated operational occurrences and accidents, computer codes used in safety analysis and evaluation of each event are presented in the safety evaluation. (author)

  17. Conceptual design of the Clinch River Breeder Reactor spent-fuel shipping cask

    International Nuclear Information System (INIS)

    Details of a baseline conceptual design of a spent fuel shipping cask for the Clinch River Breeder Reactor (CRBR) are presented including an assessment of shielding, structural, thermal, fabrication and cask/plant interfacing problems. A basis for continued cask development and for new technological development is established. Alternates to the baseline design are briefly presented. Estimates of development schedules, cask utilization and cost schedules, and of personnel dose commitments during CRBR in-plant handling of the cask are also presented

  18. Present day design challenges exemplified by the Clinch River Breeder Reactor Plant

    International Nuclear Information System (INIS)

    The present day design challenges faced by the Clinch River Breeder Reactor Plant engineer result from two causes. The first cause is aspiration to achieve a design that will operate at conditions which are desirable for future LMFBRs in order for them to achieve low power costs and good breeding. The second cause is the licensing impact. Although licensing the CRBRP won't eliminate future licensing effort, many licensing questions will have been resolved and precedents set for the future LMFBR industry

  19. Management of New Production Reactor waste streams at Savannah River

    International Nuclear Information System (INIS)

    To ensure the adequacy of available facilities, the disposition of the several waste types generated in support of a heavy-water NPR operation at the Savannah River Site were projected through waste- treatment and disposal facilities after the year 2000. Volumes of high-level, low-level radioactive, TRU, hazardous, mixed and non-radioactive waste were predicted for early assessments of environmental impacts and to provide a baseline for future waste-minimization initiatives. Life-cycle unit costs for disposal of the waste, adjusted to reflect waste management capabilities in the NPR operating time frame, were developed to evaluate the economic effectiveness of waste-minimization activities in the NPR program

  20. Trends in large-scale testing of reactor structures

    International Nuclear Information System (INIS)

    Large-scale tests of reactor structures have been conducted at Sandia National Laboratories since the late 1970s. This paper describes a number of different large-scale impact tests, pressurization tests of models of containment structures, and thermal-pressure tests of models of reactor pressure vessels. The advantages of large-scale testing are evident, but cost, in particular limits its use. As computer models have grown in size, such as number of degrees of freedom, the advent of computer graphics has made possible very realistic representation of results - results that may not accurately represent reality. A necessary condition to avoiding this pitfall is the validation of the analytical methods and underlying physical representations. Ironically, the immensely larger computer models sometimes increase the need for large-scale testing, because the modeling is applied to increasing more complex structural systems and/or more complex physical phenomena. Unfortunately, the cost of large-scale tests is a disadvantage that will likely severely limit similar testing in the future. International collaborations may provide the best mechanism for funding future programs with large-scale tests. (author)

  1. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8∼2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4∼2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure

  2. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baek, W. P.; Song, C. H.; Kim, Y. S. and others

    2005-02-15

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

  3. Design and Status of RERTR Irradiation Tests in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Daniel M. Wachs; Richard G. Ambrosek; Gray Chang; Mitchell K. Meyer

    2006-10-01

    Irradiation testing of U-Mo based fuels is the central component of the Reduced Enrichment for Research and Test Reactors (RERTR) program fuel qualification plan. Several RERTR tests have recently been completed or are planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory in Idaho Falls, ID. Four mini-plate experiments in various stages of completion are described in detail, including the irradiation test design, objectives, and irradiation conditions. Observations made during and after the in-reactor RERTR-7A experiment breach are summarized. The irradiation experiment design and planned irradiation conditions for full-size plate test are described. Progress toward element testing will be reviewed.

  4. Evaluation of the Pseudalert/Quanti-Tray MPN Test for the Rapid Enumeration of Pseudomonas aeruginosa in Swimming Pool and Spa Pool Waters.

    Science.gov (United States)

    Sartory, David P; Brewer, Megan; Beswick, Agnieszka; Steggles, Darron

    2015-12-01

    This study assessed the performance of a new most probable number test (Pseudalert/Quanti-Tray) for the enumeration of Pseudomonas aeruginosa from swimming pool and spa pool waters by comparing it to the international and national membrane filtration-based culture methods for P. aeruginosa: ISO 16266:2006 and UK The Microbiology of Drinking Water-Part 8 (MoDW Part 8) which both use Pseudomonas CN agar. The comparison was based on the calculation of mean relative differences between the two methods conducted according to ISO 17994:2014. Using both routine pool water samples (149 from 8 laboratories) and artificially contaminated samples (309 from 7 laboratories), paired counts from each sample and enumeration method were analysed. For routine samples, there were insufficient data for a conclusive assessment, but the data do indicate at least equivalent performance of Pseudalert/Quanti-Tray to the reference methods. For the artificially contaminated samples, the data also did not result in a statistically conclusive assessment but did indicate potentially better performance of Pseudalert/Quanti-Tray. Combining the data from the routine samples and artificially contaminated samples resulted in an ISO 17994 outcome that the two methods were not statistically significantly different. Thus, the Pseudalert/Quanti-Tray method is an acceptable alternative to ISO 16266 and MoDW Part 8. The Pseudalert/Quanti-Tray method has the advantage in that it does not require confirmation testing, and of providing confirmed counts within 24-28 h incubation compared to 40-48 h or longer for the ISO 16266 and MoDW Part 8 methods.

  5. Advanced Test Reactor National Scientific User Facility Partnerships

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Todd R. Allen; Jeff B. Benson; James I. Cole; Mary Catherine Thelen

    2012-03-01

    In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of

  6. Reactor vessel integrity analysis based upon large scale test results

    International Nuclear Information System (INIS)

    The fracture mechanics analysis of a nuclear reactor pressure vessel is discussed to illustrate the impact of knowledge gained by large scale testing on the demonstration of the integrity of such a vessel. The analysis must be able to predict crack initiation, arrest and reinitiation. The basis for the capability to make each prediction, including the large scale test information which is judged appropriate, is identified and the confidence in the applicability of the experimental data to a vessel is discussed. Where there is inadequate data to make a prediction with confidence or where there are apparently conflicting data, recommendations for future testing are presented. 15 refs., 6 figs.. 1 tab

  7. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform the tests for design, operation, and safety regulation of pressurized water reactors. In the first phase of this project (1997.8∼2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished. In the second phase (2002.4∼2005.2), an optimized design of the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) was established and the construction of the facility was almost completed. In the third phase (2005.3∼2007.2), the construction and commission tests of the ATLAS are to be completed and some first-phase tests are to be conducted

  8. New results from pulse tests in the CABRI reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schmitz, F.; Papin, J.; Haessler, M. [Institut de Proterction et de Surete Nucleaire, Saint Paul Lez Durance (France)] [and others

    1996-03-01

    At the 21st and 22nd WRSM (1,2), the motivation and objectives of the French program on the behaviour of high burnup PWR fuel under RIA conditions in the CABRI test reactor has been presented. The major results of the three first tests of the test matrix were presented and in particular REP-Na1, which failed at an unexpected low level of fuel enthalpy, was exposed to the community of nuclear safety research. At this time, no final understanding was reached for the origin of the failure. This objective is reached now. Two further tests, REP-Na4 and 5, have been performed in 1995, they demonstrated a satisfactory and safe behaviour by resisting to the early phase of severe loading during the RIA pulse test. Further examination work and analytical testing is in progress and the next tests with MOX fuel are being prepared.

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

    Energy Technology Data Exchange (ETDEWEB)

    Stahl, D.

    1982-12-01

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

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

    International Nuclear Information System (INIS)

    A thorough review is provided on nuclear fuels for steady-state thermal research and test reactors. The review was conducted to provide a documented data base in support of recent advances in research and test reactor fuel development, manufacture, and demonstration in response to current US policy on availability of enriched uranium. The review covers current fabrication practice, fabrication development efforts, irradiation performance, and properties affecting fuel utilization, including thermal conductivity, specific heat, density, thermal expansion, corrosion, phase stability, mechanical properties, and fission-product release. The emphasis is on US activities, but major work in Europe and elsewhere is included. The standard fuel types discussed are the U-Al alloy, UZrH/sub x/, and UO2 rod fuels. Among new fuels, those given major emphasis include H3Si-Al dispersion and UO2 caramel plate fuels

  11. Acceptance test of graphite components in nuclear reactor

    International Nuclear Information System (INIS)

    The HTTR is the first high temperature gas-cooled reactor in Japan. It is a test reactor with thermal power of 30 MW and coolant outlet temperature of 950degC at maximum. To achieve high temperature coolant core internals were made of graphite and carbon materials due to their excellent thermal resistivity. After fabrication of graphite and carbon components at works they were installed in the HTTR, and now it is in the power up testing stage. Concerning the inspection standard of the graphite and carbon components, nondomestic standard exists as main components in the nuclear reactor. It is necessary, therefore, to prescribe the inspection standards for the HTTR graphite components. Many research and developments in relation to the inspection standard, e.g. in the research field of nondestructive examination of the graphite material, had been performed, and then the JAERI established the inspection standard. The acceptance test of the graphite and carbon components was carried out based on the inspection standard. This paper prescribes the outline of the established inspection standard. (author)

  12. Variation in dengue virus plaque reduction neutralization testing: systematic review and pooled analysis

    Directory of Open Access Journals (Sweden)

    Rainwater-Lovett Kaitlin

    2012-09-01

    Full Text Available Abstract Background The plaque reduction neutralization test (PRNT remains the gold standard for the detection of serologic immune responses to dengue virus (DENV. While the basic concept of the PRNT remains constant, this test has evolved in multiple laboratories, introducing variation in materials and methods. Despite the importance of laboratory-to-laboratory comparability in DENV vaccine development, the effects of differing PRNT techniques on assay results, particularly the use of different dengue strains within a serotype, have not been fully characterized. Methods We conducted a systematic review and pooled analysis of published literature reporting individual-level PRNT titers to identify factors associated with heterogeneity in PRNT results and compared variation between strains within DENV serotypes and between articles using hierarchical models. Results The literature search and selection criteria identified 8 vaccine trials and 25 natural exposure studies reporting 4,411 titers from 605 individuals using 4 different neutralization percentages, 3 cell lines, 12 virus concentrations and 51 strains. Of 1,057 titers from primary DENV exposure, titers to the exposure serotype were consistently higher than titers to non-exposure serotypes. In contrast, titers from secondary DENV exposures (n = 628 demonstrated high titers to exposure and non-exposure serotypes. Additionally, PRNT titers from different strains within a serotype varied substantially. A pooled analysis of 1,689 titers demonstrated strain choice accounted for 8.04% (90% credible interval [CrI]: 3.05%, 15.7% of between-titer variation after adjusting for secondary exposure, time since DENV exposure, vaccination and neutralization percentage. Differences between articles (a proxy for inter-laboratory differences accounted for 50.7% (90% CrI: 30.8%, 71.6% of between-titer variance. Conclusions As promising vaccine candidates arise, the lack of standardized assays among

  13. Effective-stress-law behavior of Austin chalk rocks for deformation and fracture conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Warpinski, N.R.; Teufel, L.W.

    1994-08-01

    Austin chalk core has been tested to determine the effective law for deformation of the matrix material and the stress-sensitive conductivity of the natural fractures. For deformation behavior, two samples provided data on the variations of the poroelastic parameter, {alpha}, for Austin chalk, giving values around 0.4. The effective-stress-law behavior of a Saratoga limestone sample was also measured for the purpose of obtaining a comparison with a somewhat more porous carbonate rock. {alpha} for this rock was found to be near 0.9. The low {alpha} for the Austin chalk suggests that stresses in the reservoir, or around the wellbore, will not change much with changes in pore pressure, as the contribution of the fluid pressure is small. Three natural fractures from the Austin chalk were tested, but two of the fractures were very tight and probably do not contribute much to production. The third sample was highly conductive and showed some stress sensitivity with a factor of three reduction in conductivity over a net stress increase of 3000 psi. Natural fractures also showed a propensity for permanent damage when net stressed exceeded about 3000 psi. This damage was irreversible and significantly affected conductivity. {alpha} was difficult to determine and most tests were inconclusive, although the results from one sample suggested that {alpha} was near unity.

  14. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors AGENCY: Nuclear...-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling- Water Reactors.'' This... testing features of emergency core cooling systems (ECCSs) for boiling-water reactors (BWRs)....

  15. Conceptual design study of a scyllac fusion test reactor

    International Nuclear Information System (INIS)

    The report describes a conceptual design study of a fusion test reactor based on the Scyllac toroidal theta-pinch approach to fusion. It is not the first attempt to describe the physics and technology required for demonstrating scientific feasibility of the approach, but it is the most complete design in the sense that the physics necessary to achieve the device goals is extrapolated from experimentally tested MHD theories of toroidal systems,and it uses technological systems whose engineering performance has been carefully calculated to ensure that they meet the machine requirements

  16. Conceptual design study of a scyllac fusion test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Thomassen, K.I. (comp.)

    1975-07-01

    The report describes a conceptual design study of a fusion test reactor based on the Scyllac toroidal theta-pinch approach to fusion. It is not the first attempt to describe the physics and technology required for demonstrating scientific feasibility of the approach, but it is the most complete design in the sense that the physics necessary to achieve the device goals is extrapolated from experimentally tested MHD theories of toroidal systems,and it uses technological systems whose engineering performance has been carefully calculated to ensure that they meet the machine requirements.

  17. Advanced Test Reactor National Scientific User Facility Progress

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Todd R. Allen; James I. Cole; Jeff B. Benson; Mary Catherine Thelen

    2012-10-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is one of the world’s premier test reactors for studying the effects of intense neutron radiation on reactor materials and fuels. The ATR began operation in 1967, and has operated continuously since then, averaging approximately 250 operating days per year. The combination of high flux, large test volumes, and multiple experiment configuration options provide unique testing opportunities for nuclear fuels and material researchers. The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected highly-enriched uranium fueled, reactor with a maximum operating power of 250 MWth. The ATR peak thermal flux can reach 1.0 x1015 n/cm2-sec, and the core configuration creates five main reactor power lobes (regions) that can be operated at different powers during the same operating cycle. In addition to these nine flux traps there are 68 irradiation positions in the reactor core reflector tank. The test positions range from 0.5” to 5.0” in diameter and are all 48” in length, the active length of the fuel. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. Goals of the ATR NSUF are to define the cutting edge of nuclear technology research in high temperature and radiation environments, contribute to improved industry performance of current and future light water reactors, and stimulate cooperative research between user groups conducting basic and applied research. The ATR NSUF has developed partnerships with other universities and national laboratories to enable ATR NSUF researchers to perform research at these other facilities, when the research objectives

  18. Reactor power cutback system test experience at YGN 4

    International Nuclear Information System (INIS)

    YGN 3 and 4 are the nuclear power plants having System 80 characteristics with a rated thermal output of 2815 MWth and a nominal net electrical output of 1040 MWe. YGN 3 achieved commercial operation on March 31, 1995 and YGN 4 completed Power Ascension Test (PAT) at 20%, 50%, 80% and 100% power by September 23, 1995. YGN 3 and 4 design incorporates the Reactor POwer Cutback System (RPCS) which reduces plant trips caused by Loss of Load (LOL)/ Turbine Trip and Loss of One Main Feedwater Pump (LOMFWP). The key design objective of the RPCS is to improve overall plant availability and performance, while minimizing challenges to the plant safety systems. The RPCS is designed to rapidly reduce reactor power by dropping preselected Control Element Assemblies (CEAs) while other NSSS control systems maintain process parameters within acceptable ranges. Extensive RPCS related tests performed during the initial startup of YGN 4 demonstrated that the RPCS can maintain the reactor on-line without opening primary or secondary safety valves and without actuating the Engineered Safety Features Actuation System (ESFAS). It is expected that use of the RPCS at YGN will increase the overall availability of the units and reduce the number of challenges to plant safety systems

  19. Advanced Gas Reactor (AGR)-5/6/7 Fuel Irradiation Experiments in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; David A. Petti; S. Blaine Grover

    2014-04-01

    The United States Department of Energy’s Very High Temperature Reactor (VHTR) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which each consist of at least five separate capsules, are being irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gases also have on-line fission product monitoring the effluent from each capsule to track performance of the fuel during irradiation. The first two experiments (designated AGR-1 and AGR-2), have been completed. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. The design of the fuel qualification experiment, designated AGR-5/6/7, is well underway and incorporates lessons learned from the three previous experiments. Various design issues will be discussed with particular details related to selection of thermometry.

  20. The Jules Horowitz reactor (JHR), a European material testing reactor (MTR), with extended experimental capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A.; Bergamaschi, Y.; Bouilloux, Y.; Bravo, X.; Guigon, B.; Rommens, M.; Tremodeux, P. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, 13 - Saint-Paul-lez-Durance (France)]|[CEA Saclay Dir. de l' Energie Nucleaire DEN, 91 - Gif sur Yvette (France)

    2003-07-01

    The Jules Horowitz Reactor (JHR) is the European MTR (Material Testing Reactor) designed to provide, after 2010, the necessary knowledge for keeping the existing power plants in operation and to design innovative reactors types with new objectives such as: minimizing the radioactive waste production, taking into account additional safety requirements, preventing risks of nuclear proliferation... To achieve such an ambitious objective. The JHR is designed with a high flexibility in order to satisfy the current demand from European industry, research and to be able to accommodate future requirements. The JHR will offer a wide range of performances and services in gathering, in a single site at Cadarache, all the necessary functionalities and facilities for an effective production of results: e.g. fuel fabrication laboratories, preparation of the instrumented devices, interpretation of the experiments, modelling. The JHR must rely on a top level scientific environment based on experts teams from CEA and EC and local universities. With a thermal flux of 7,4.10{sup 14} ncm{sup -2} s{sup -1} and a fast flux of 6,4.10{sup 14} ncm{sup -2}s{sup -1}, it is possible to carry out irradiation experiments on materials and fuels whatever the reactor type considered. It will also be possible to carry out locally, fast neutron irradiation to achieve damage effect up to 25 dpa/year. (dpa = displacement per atom.) The study of the fuels behavior under accidental conditions, from analytical experiments, on a limited amount of irradiated fuel, is a major objective of the project. These oriented safety tests are possible by taking into account specific requirements in the design of the facility such as the tightness level of the containment building, the addition of an alpha hot cell and a laboratory for on line fission products measurement. (authors)

  1. Chalk: composition, diagenesis and physical properties

    DEFF Research Database (Denmark)

    Fabricius, Ida Lykke

    2007-01-01

    entry pressure, and elastic moduli are consequences of primary sediment composition and of subsequent diagenetic history as caused by microbial action, burial stress, temperature, and pore pressure. Porosity is a main determining factor for other properties. For a given porosity, the specific surface......Chalk is a sedimentary rock of unusually high homogeneity on the scale where physical properties are measured, but the properties fall in wide ranges. Chalk may thus be seen as the ideal starting point for a physical understanding of rocks in general. Properties as porosity, permeability, capillary...... of the sediment controls permeability and capillary entry pressure. As diagenesis progresses, the specific surface is less and less due to the calcite component and more and more due to the fine-grained silicates, as a reflection of the coarsening and cementation of the calcite crystals. The elastic moduli, which...

  2. Reactor Testing and Qualification: Prioritized High-level Criticality Testing Needs

    Energy Technology Data Exchange (ETDEWEB)

    S. Bragg-Sitton; J. Bess; J. Werner; G. Harms; S. Bailey

    2011-09-01

    Researchers at the Idaho National Laboratory (INL) were tasked with reviewing possible criticality testing needs to support development of the fission surface power system reactor design. Reactor physics testing can provide significant information to aid in development of technologies associated with small, fast spectrum reactors that could be applied for non-terrestrial power systems, leading to eventual system qualification. Several studies have been conducted in recent years to assess the data and analyses required to design and build a space fission power system with high confidence that the system will perform as designed [Marcille, 2004a, 2004b; Weaver, 2007; Parry et al., 2008]. This report will provide a summary of previous critical tests and physics measurements that are potentially applicable to the current reactor design (both those that have been benchmarked and those not yet benchmarked), summarize recent studies of potential nuclear testing needs for space reactor development and their applicability to the current baseline fission surface power (FSP) system design, and provide an overview of a suite of tests (separate effects, sub-critical or critical) that could fill in the information database to improve the accuracy of physics modeling efforts as the FSP design is refined. Some recommendations for tasks that could be completed in the near term are also included. Specific recommendations on critical test configurations will be reserved until after the sensitivity analyses being conducted by Los Alamos National Laboratory (LANL) are completed (due August 2011).

  3. Final Physics Report for the Engineering Test Reactor

    International Nuclear Information System (INIS)

    This report is a summary of the physics design work performed on the Engineering Test Reactor. The ETR presents computational difficulties not found in other reactors because of the large number of experimental holes in the core. The physics of the ETR depends strongly upon the contents of the in-core experimental facilities. In order to properly evaluate the reactor' taking into account the experiments in the core, multi-region, two-dimensional calculations are required. These calculations require the use of a large computer such as the Remington Rand Univac and are complex and expensive enough to warrant a five-stage program: 1. In the early stages of design, only preliminary two-dimensional calculations were performed .in order to obtain a rough idea of the general behavior of the reactor and its critical mass with tentative experiments in place. 2. A large amount of work was carried out in which the reactor was approximated as one with a uniform homogeneous core. With this model, detailed studies were carried out to investigate the feasibility and to obtain general design data on such points as the design and properties of the gray and black control rods, the design of the beryllium reflector, gamma and neutron heating, the use of burnable poisons, etc. In performing these calculations, use was made of the IBM 650 PROD code obtained from KAPL. 3. With stages 1 and 2 carried out, two-dimensional calculations of the core at start-up conditions were performed on the Univac computer. 4. Detailed two-dimensional calculations of the properties of the ETR with a proposed first set of experiments in place were carried out. 5. A series of nuclear tests were performed at the reactivity measurements facility at the MTR site in order to confirm the validity of the analytical techniques in physics analysis. In performing the two-dimensional Univac calculations, the MUG code developed by KAPL and the Cuthill code developed at the David Taylor Model Basin were utilized. In

  4. Optimisation of safety parameters in fast breeder test reactor

    International Nuclear Information System (INIS)

    Full text: Optimisation of safety parameters is an important aspect to be considered in the design of nuclear power plant and also becomes extremely important activity to be followed up during the commissioning and operating phases of the plant taking into account the operational feed back and review of incidental situations and available diversity and reliability. Otherwise, the spurious/ superfluous trips on the reactor besides affecting the availability of the plant, initiate plant transients causing stress for the plant equipment resulting in reduction of plant life. This activity has a significant role to play in attaining the maximum availability of the plant, without compromising safety. The study and evolution of optimisation process in fast breeder test reactor (FBTR); at Kalpakkam has been an interesting and rewarding experience

  5. Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    E. Perry; J. Chrzanowski; K. Rule; M. Viola; M. Williams; R. Strykowsky

    1999-11-01

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. The Decontamination and Decommissioning (D and D) of the TFTR is scheduled to occur over a period of three years beginning in October 1999. This is not a typical Department of Energy D and D Project where a facility is isolated and cleaned up by ''bulldozing'' all facility and hardware systems to a greenfield condition. The mission of TFTR D and D is to: (a) surgically remove items which can be re-used within the DOE complex, (b) remove tritium contaminated and activated systems for disposal, (c) clear the test cell of hardware for future reuse, (d) reclassify the D-site complex as a non-nuclear facility as defined in DOE Order 420.1 (Facility Safety) and (e) provide data on the D and D of a large magnetic fusion facility. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The record-breaking deuterium-tritium experiments performed on TFTR resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 75 mRem/hr. These radiological hazards along with the size and shape of the Tokamak present a unique and challenging task for dismantling.

  6. Field test of wireless sensor network in the nuclear environment

    International Nuclear Information System (INIS)

    Wireless sensor networks (WSNs) are appealing options for the health monitoring of nuclear power plants due to their low cost and flexibility. Before they can be used in highly regulated nuclear environments, their reliability in the nuclear environment and compatibility with existing devices have to be assessed. In situ electromagnetic interference tests, wireless signal propagation tests, and nuclear radiation hardness tests conducted on candidate WSN systems at AECL Chalk River Labs are presented. The results are favourable to WSN in nuclear applications. (author)

  7. Field test of wireless sensor network in the nuclear environment

    Energy Technology Data Exchange (ETDEWEB)

    Li, L., E-mail: lil@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Wang, Q.; Bari, A. [Univ. of Western Ontario, London, Ontario (Canada); Deng, C.; Chen, D. [Univ. of Electronic Science and Technology of China, Chengdu, Sichuan (China); Jiang, J. [Univ. of Western Ontario, London, Ontario (Canada); Alexander, Q.; Sur, B. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2014-06-15

    Wireless sensor networks (WSNs) are appealing options for the health monitoring of nuclear power plants due to their low cost and flexibility. Before they can be used in highly regulated nuclear environments, their reliability in the nuclear environment and compatibility with existing devices have to be assessed. In situ electromagnetic interference tests, wireless signal propagation tests, and nuclear radiation hardness tests conducted on candidate WSN systems at AECL Chalk River Labs are presented. The results are favourable to WSN in nuclear applications. (author)

  8. In-reactor optical dosimetry in high-temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    The applicability of fused silica core optical fibres to in-reactor dosimetry was demonstrated at elevated temperatures and a special irradiation rig was developed for realizing high-temperature optical dosimetry in a high-temperature test reactor (HTTR) at the Oarai Research Establishment of JAERI (Japan Atomic Energy Research Institute). The paper will describe the present status of preparation for the high-temperature dosimetry in HTTR, utilising radiation-resistant optical fibres and radioluminescent materials. Temperature measurement with a high-speed response is the main target for the present optical dosimetry, which could be applied for monitoring transient behaviours of the HTTR. This could be realised by measuring the intensity of thermoluminescence and black body radiation in the infrared region. For monitoring reactor powers, optical measurements in the visible region are essential. At present, the measurement of the intensity of Cerenkov radiation is the most promising area of study. Other possibilities with radioluminescent materials having luminescent peaks in the visible region are under consideration. One of the candidates will be silica, which has a robust radioluminescent peak at 450 nm. (author)

  9. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    International Nuclear Information System (INIS)

    The regulatory requirement to develop an upgraded safety basis for a DOE Nuclear Facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830). Subpart B of 10 CFR 830, ''Safety Basis Requirements,'' requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements. 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, ''Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants'' as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830

  10. Completion summary for borehole USGS 136 near the Advanced Test Reactor Complex, Idaho National Laboratory, Idaho

    Science.gov (United States)

    Twining, Brian V.; Bartholomay, Roy C.; Hodges, Mary K.V.

    2012-01-01

    In 2011, the U.S. Geological Survey, in cooperation with the U.S. Department of Energy, cored and completed borehole USGS 136 for stratigraphic framework analyses and long-term groundwater monitoring of the eastern Snake River Plain aquifer at the Idaho National Laboratory. The borehole was initially cored to a depth of 1,048 feet (ft) below land surface (BLS) to collect core, open-borehole water samples, and geophysical data. After these data were collected, borehole USGS 136 was cemented and backfilled between 560 and 1,048 ft BLS. The final construction of borehole USGS 136 required that the borehole be reamed to allow for installation of 6-inch (in.) diameter carbon-steel casing and 5-in. diameter stainless-steel screen; the screened monitoring interval was completed between 500 and 551 ft BLS. A dedicated pump and water-level access line were placed to allow for aquifer testing, for collecting periodic water samples, and for measuring water levels. Geophysical and borehole video logs were collected after coring and after the completion of the monitor well. Geophysical logs were examined in conjunction with the borehole core to describe borehole lithology and to identify primary flow paths for groundwater, which occur in intervals of fractured and vesicular basalt. A single-well aquifer test was used to define hydraulic characteristics for borehole USGS 136 in the eastern Snake River Plain aquifer. Specific-capacity, transmissivity, and hydraulic conductivity from the aquifer test were at least 975 gallons per minute per foot, 1.4 × 105 feet squared per day (ft2/d), and 254 feet per day, respectively. The amount of measureable drawdown during the aquifer test was about 0.02 ft. The transmissivity for borehole USGS 136 was in the range of values determined from previous aquifer tests conducted in other wells near the Advanced Test Reactor Complex: 9.5 × 103 to 1.9 × 105 ft2/d. Water samples were analyzed for cations, anions, metals, nutrients, total organic

  11. Geologic setting of the New Production Reactor within the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Price, V. [Westinghouse Savannah River Co., Aiken, SC (United States); Fallaw, W.C. [Furman Univ., Greenville, SC (United States). Dept. of Geology; McKinney, J.B. [Exploration Resources, Inc., Athens, GA (United States)

    1991-12-31

    The geology and hydrology of the reference New Production Reactor (NPR) site at Savannah River Site (SRS) have been summarized using the available information from the NPR site and areas adjacent to the site, particularly the away from reactor spent fuel storage site (AFR site). Lithologic and geophysical logs from wells drilled near the NPR site do not indicate any faults in the upper several hundred feet of the Coastal Plain sediments. However, the Pen Branch Fault is located about 1 mile south of the site and extends into the upper 100 ft of the Coastal Plain sequence. Subsurface voids, resulting from the dissolution of calcareous portions of the sediments, may be present within 200 ft of the surface at the NPR site. The water table is located within 30 to 70 ft of the surface. The NPR site is located on a groundwater divide, and groundwater flow for the shallowest hydraulic zones is predominantly toward local streams. Groundwater flow in deeper Tertiary sediments is north to Upper Three Runs Creek or west to the Savannah River Swamp. Groundwater flow in the Cretaceous sediments is west to the Savannah River.

  12. Apparatus for irradiating in a magnetic field in a swimming pool type reactor, at high or low temperatures (1962)

    International Nuclear Information System (INIS)

    An apparatus for irradiation in a swimming-pool with a magnetic field of 5000 oersteds is described. An aluminium coil with a very high current density is water cooled. A relatively great Volume can be used in the coil center, and two furnaces can be introduced, or a liquid nitrogen cryostat. (author)

  13. Local stability tests in Dresden 2 boiling water reactor

    International Nuclear Information System (INIS)

    This report presents the results of a local stability test performed at Dresden Unit 2 in May 1983 to determine the effect of a new fuel element design on local channel stability. This test was performed because the diameter of the new fuel rods increases the heat transfer coefficient, making the reactor more responsive and, thus, more susceptible to instabilities. After four of the new fuel elements with a 9 x 9 array of fuel rods were loaded into Dresden 2, the test was performed by inserting an adjacent control rod all the way in and then withdrawing it to its original position at maximum speed. At the moment of the test, reactor conditions were 52.7% power and 38.9% flow. Both the new 9 x 9 fuel elements and the standard 8 x 8 ones proved to be locally stable when operating at minimum pump speed at the beginning of cycle in Dresden 2, and no significant difference was found between the behavior of the two fuel types. Finally, Dresden 2 showed a high degree of stability during control rod and normal noise type perturbations

  14. Fracture toughness test methods and examples for fusion reactor materials

    International Nuclear Information System (INIS)

    This paper introduces the importance of the evaluation of fracture toughness in nuclear fusion reactor structural materials, and the fracture toughness evaluation methods that are used as the standards and their actual examples. It also discusses the problems involved in the standardized approach and the efforts for the technology improvement. To evaluate the material life under nuclear fusion reactor environment, fracture toughness measurement after neutron irradiation is indispensable. Due to a limitation in the irradiation area size of an irradiation reactor, and to avoid the temperature difference in a specimen, the size of the specimen is required to be minimized, which is different from the common standards. As for the size effect of the test specimen, toughness value tends to decrease when ligament length is 7 mm or below. The main problems and challenges are as follows. (1) As for the tendency that fracture toughness value decreases along with the miniaturization of the ligament length, it is necessary to elucidate the mechanism of size effects, and to develop the correction method for size effects. (2) As for the issues of the curve shape and application to irradiation time in the master curve method, it is necessary to review the data checking method and plastic constraint conditions for crack tip M = 30 that is stipulated in ASTM E1921, and to elucidate the material dependence of master curve shape. (A.O.)

  15. STS-45 MS Foale in EMU is lowered into JSC's WETF pool for underwater test

    Science.gov (United States)

    1991-01-01

    STS-45 Atlantis, Orbiter Vehicle (OV) 104, Mission Specialist (MS) C. Michael Foale, fully suited in an extravehicular mobility unit (EMU) and standing on a platform, is lowered into a 25 ft deep pool for an underwater simulation of contingency extravehicular activity (EVA) procedures. The pool is located in JSC's Weightless Environment Training Facility (WETF) Bldg 29. Weights are added around Foale's ankles so he will be neutrally buoyant during the simulation. SCUBA-equipped divers (swimmers) assist during the exercise.

  16. Program plan for decontamination and decommissioning the Materials Testing Reactor at the INEL

    International Nuclear Information System (INIS)

    A discussion is presented of a program plan developed for the dismantling of the Materials Testing Reactor located in the Testing Reactor Area (TRA) of the Idaho National Engineering Laboratory. Included are the scope of work, dismantling problems resulting from the nature of construction of the MTR, and a program plan for physically dismantling the reactor

  17. Reactor recirculation pump seal lifetime improvement efforts at River Bend Station

    International Nuclear Information System (INIS)

    River Bend Station is a single-unit Boiling Water Reactor (BWR/6) with a Mark III containment. Initial criticality was achieved on October 31, 1985, and the unit entered Commercial Operation in June, 1986. Two model 20 X 20 X 33 type RV centrifugal pumps manufactured by Bingham-Willamette (now Sulzer-Bingham) are used for Reactor Recirculation Pump (RRP) service. The two pumps described in this paper as RRP-A and RRP-B are single suction, single stage, vertical, double volute pumps driven by 6,300 horsepower GE electric motors. Rated flow for each pump is 32,500 gallons per minute. Sulzer-Bingham mechanical shaft seals, model 750A are employed for RRP service. The seal assembly consists of a tandem seal cartridge (dual seals) with carbon stationary seal faces and tungsten-carbide rotating seal faces

  18. Knowledges and abilities catalog for nuclear power plant operators: Savannah River Site (SRS) production reactors

    International Nuclear Information System (INIS)

    The Knowledges and Abilities Catalog for Nuclear Power Plant Operations: Savannah River Site (SRS) Production Reactors, provides the basis for the development of content-valid certification examinations for Senior Reactor Operators (SROs) and Central Control Room Supervisors (SUP). The position of Shift Technical Engineer (STE) has been included in the catalog for completeness. This new SRS reactor operating shift crew position is held by an individual holding a CCR Supervisor Certification who has received special engineering and technical training. Also, the STE has a Bachelor of Science degree in engineering or a related technical field. The SRS catalog contains approximately 2500 knowledge and ability (K/A) statements for SROs and SUPs at heavy water moderated production reactors. Each K/A statement has been rated for its importance to the safe operation of the plant in a manner ensuring the health and safety of the public. The SRS K/A catalog is presently organized into five major sections: Plant Systems grouped by Safety Function, Plant Wide Generic K/As, Emergency Plant Evolutions, Theory and Components (to be developed)

  19. How burial diagenesis of chalk sediments controls sonic velocity and porosity

    DEFF Research Database (Denmark)

    Fabricius, Ida Lykke

    2003-01-01

    Based on P-wave velocity and density data, a new elastic model for chalk sediments is established. The model allows the construction of a series of isoframe (IF) curves, each representing a constant part of the mineral phase contributing to the solid frame. The IF curves can be related to the pro......Based on P-wave velocity and density data, a new elastic model for chalk sediments is established. The model allows the construction of a series of isoframe (IF) curves, each representing a constant part of the mineral phase contributing to the solid frame. The IF curves can be related......, whereby IF increases and chalk forms. Rock mechanical tests show that when compaction requires more than in-situ stress, porosity reduction is arrested. During subsequent burial, crystals and pores grow in size as a consequence of the continuing recrystallization. ne lack of porosity loss during......, and depending on pore-water chemistry and temperature, pore-filling cementation may occur over a relatively short depth interval. Limestone and mixed sedimentary rock form, and porosity may be reduced to less than 20%. Isoframe increases to more than 0.6. In hydrocarbon reservoirs in North Sea chalk, relatively...

  20. Alpha Particle Physics Experiments in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Budny, R.V.; Darrow, D.S.; Medley, S.S.; Nazikian, R.; Zweben, S.J.; et al.

    1998-12-14

    Alpha particle physics experiments were done on the Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium (DT) run from 1993-1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single-particle confinement model in magnetohydrodynamic (MHD) quiescent discharges. Also, the observed alpha particle interactions with sawteeth, toroidal Alfvén eigenmodes (TAE), and ion cyclotron resonant frequency (ICRF) waves were roughly consistent with theoretical modeling. This paper reviews what was learned and identifies what remains to be understood.