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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. Decommissioning Experience: Chalk River, Canada

    International Nuclear Information System (INIS)

    Full text: Atomic Energy of Canada Limited has reported that work has continued on the decommissioning of old structures on the Chalk River laboratory site. An environmental assessment was approved in 2006 for the decommissioning of the NRX reactor fuel bays (A and B). The regulator approved two work packages for the removal of water and the wooden structure over the bays. The A bays were cleaned as far as possible and were emptied in 2007. Decontamination work will continue. Sections of the B bays were filled with sand and other parts filled with water. NRX is currently in storage (i.e. a dormant state) with surveillance. (author)

  4. 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...

  5. Simulations of Two-Well Tracer Tests in Stratified Aquifers at the Chalk River and the Mobile Sites

    Science.gov (United States)

    Huyakorn, Peter S.; Andersen, Peter F.; Molz, Fred J.; Güven, Oktay; Melville, Joel G.

    1986-07-01

    A simulation study of two-well injection-withdrawal tracer tests in stratified granular aquifers at two widely separated sites is presented. The first site is located near the Chalk River Nuclear Laboratories in Canada, and the second site is located in Mobile, Alabama. Field data and test conditions at these sites are substantially different in terms of vertical distributions of hydraulic conductivity, well spacings, flow rates, test durations, and tracer travel distances. Furthermore, the test at the Chalk River site was conducted in a recirculating mode, whereas the test at the Mobile site was conducted in a nonrecirculating mode. Simulations of these tests were performed in three dimensions using the curvilinear finite element model developed in the previous paper of this series. The simulations incorporated measured vertical variations in relative hydraulic conductivity and local dispersivity values that are small fractions (between 1/1000 and 1/100) of the spacing between the injection and the withdrawal wells. The local dispersivities are used to account for local hydrodynamic dispersion and are chosen independently so that they are not affected by the scales of the tests. Simulation results obtained from the model are presented. Interpretation of these results is made in conjunction with measured breakthrough curves at the withdrawal well and multilevel observation wells. For the Chalk River site, predicted and measured breakthrough curves at the withdrawal well are in good agreement over the earlier part of the test duration. Deviation of the field data from the model prediction occurs over the second part, where the predicted breakthrough curves show a declining trend but the field data plot does not. For the Mobile site, predicted and measured breakthrough curves at the withdrawal well show similar trends throughout the entire test duration and are in good agreement overall. Model predictions of the effect of hydraulic conductivity stratification on

  6. Summary of loops in the Chalk River NRX and NRU reactors

    International Nuclear Information System (INIS)

    The design and operating parameters of the high pressure, high temperature light water loops in the NRX and NRU reactors are presented to assist experimenters reviewing these facilities for their experiments. The NRX and NRU reactor design and operating data of interest to the experimenters are also presented. (author)

  7. Advanced fuel cycle development at Chalk River Laboratories

    International Nuclear Information System (INIS)

    Chalk River Laboratories (CRL) has a mandate from the Canadian government to develop nuclear technologies that support generation of clean, safe energy. This includes the development of advanced nuclear fuel technologies to ensure sustainable energy sources for Canadians. The Fuel Development Branch leads CRL's development of advanced nuclear-reactor fuels. CRL capabilities include fuel fabrication development, irradiation testing, post-irradiation examination (PIE), materials characterization and code development (modeling). This paper provides an overview of these capabilities and describes recent development activities that support fuel-cycle flexibility in heavy-water reactors. This includes a review of irradiation testing and PIE for mixed-oxide, thoria, high-burnup UO2 and low-void reactivity fuels and burnable neutron absorbers. Fabrication development, material characterizations and modeling associated with these tests are also described. (author)

  8. 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)

  9. 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)

  10. The Chalk River Tritium Extraction Plant

    International Nuclear Information System (INIS)

    The Chalk River Tritium Extraction Plant for removal of tritium from heavy water is described. Tritium is present in the heavy water from research reactors in the form of DTO at a concentration in the range of 1-35 Ci/kg. It is removed by a combination of catalytic exchange to transfer the tritium from DTO to DT, followed by cryogenic distillation to separate and concentrate the tritium to T2. The tritium product is reacted with titanium and packaged for transportation and storage as titanium tritide. The plant processes heavy water at a rate of 25 kg/h and removes 80% of the tritium and 90% of the protium per pass. Catalytic exchange is carried out in the liquid phase using a proprietary wetproofed catalyst. The plant serves two roles in the Canadian fusion program: it produces pure tritium for use in fusion research and development, and it demonstrates on an industrial scale many of the tritium technologies that are common to the tritium systems in fusion reactors (author)

  11. 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)

  12. 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)

  13. 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

  14. 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.

  15. 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)

  16. 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

  17. 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)

  18. Inverse modeling of Chalk River block

    International Nuclear Information System (INIS)

    Within the framework of the international project HYDROCOIN, a block of fractured monzonitic gneiss within the facilities of Chalk River National Laboratories, Canada, was selected as a test case to study and develop strategies for the calibration and validation of groundwater flow models. Adopting a quasi-three dimensional formulation, the fractures were simulated by two-dimensional finite elements and the rock mass was simulated by strings of line elements. The models were calibrated using, first, steady-state data and, second, transient data. Model calibration involved both identification of model parameters and model structure. Model parameters were obtained by automatic estimation based on measures of the model response and prior information about the model parameters. Excellent agreement between measured and computed heads was obtained for the transient runs. However, such match was only fair in steady-state. Model Structures Identification criteria were used to rank the performance of several model structures. In the steady state the model structure identification criteria did not strongly support increasing the model complexity. However, it is also believed that the information content of the steady state data was quite poor. In contrast, the transient data being both more numerous and more informative than steady-state data, allowed the model structure identification criteria to suggest more complex models. The validation runs were performed on data corresponding to interference pump tests different from the ones used for calibration. The prediction errors in these runs were relatively small and consistent with the calibration uncertainty. Furthermore, the ranking of the models performances during validation runs was the same as the one obtained at the calibration stage, using Model Structure Identification Criteria. (author) 26 figs., 17 tabs., 39 refs

  19. Mortality among long-term Chalk River employees

    International Nuclear Information System (INIS)

    Mortality among Chalk River Nuclear Laboratory (CRNL) employees who died during employment or after retirement has been updated to 1985 December 31. Data in tabular form are presented for overall mortality for male and female employees, for the participants in the clean-up for the NRX and NRU reactor accidents and for a group of CRNL staff with lifetime accumulative doses in excess of 0.2 Sv. Data are also presented on the different types of cancer causing death among male employees. No statistically significant increases in cancer deaths were found in any of the groups analyzed. 25 refs

  20. 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

  1. Dealing with Historical Discrepancies: The Recovery of National Research Experiment (NRX) Reactor Fuel Rods at Chalk River Laboratories (CRL) - 13324

    International Nuclear Information System (INIS)

    Following the 1952 National Research Experiment (NRX) Reactor accident, fuel rods which had short irradiation histories were 'temporarily' buried in wooden boxes at the 'disposal grounds' during the cleanup effort. The Nuclear Legacy Liabilities Program (NLLP), funded by Natural Resources Canada (NRCan), strategically retrieves legacy waste and restores lands affected by Atomic Energy of Canada Limited (AECL) early operations. Thus under this program the recovery of still buried NRX reactor fuel rods and their relocation to modern fuel storage was identified as a priority. A suspect inventory of NRX fuels was compiled from historical records and various research activities. Site characterization in 2005 verified the physical location of the fuel rods and determined the wooden boxes they were buried in had degraded such that the fuel rods were in direct contact with the soil. The fuel rods were recovered and transferred to a modern fuel storage facility in 2007. Recovered identification tags and measured radiation fields were used to identify the inventory of these fuels. During the retrieval activity, a discrepancy was discovered between the anticipated number of fuel rods and the number found during the retrieval. A total of 32 fuel rods and cans of cut end pieces were recovered from the specified site, which was greater than the anticipated 19 fuel rods and cans. This discovery delayed the completion of the project, increased the associated costs, and required more than anticipated storage space in the modern fuel storage facility. A number of lessons learned were identified following completion of this project, the most significant of which was the potential for discrepancies within the historical records. Historical discrepancies are more likely to be resolved by comprehensive historical record searches and site characterizations. It was also recommended that a complete review of the wastes generated, and the total affected lands as a result of this historic

  2. A description of the tritium facility at the Chalk River Laboratories

    International Nuclear Information System (INIS)

    AECL's Tritium Facility is located at its Chalk River Laboratories (CRL). The Tritium Facility was originally built to support the tritium technology needs for CANDU reactors and Canadian fusion program. The Tritium Facility commenced its operation in 1979. Since its inception, it has been involved in the development of heavy water detritiation and upgrading processes, development and testing of tritium-breeder materials and design and testing of fusion-fuel cleanup systems for fusion reactor applications, investigation of tritium-materials interactions, tritium storage getters etc. The Tritium Facility also contributed to the design, construction and commissioning activities of the Combined Electrolysis and Catalytic Exchange Upgrading and Detritiation (CECE-UD) Facility at CRL and the Wolsong Tritium Removal Facility (WTRF) in Korea. This paper describes the general set-up of the laboratory, its capabilities and the current tritium-related activities. (author)

  3. Province of Ontario nuclear emergency plan part V - Chalk River

    International Nuclear Information System (INIS)

    The aim of Part 5 of the Provincial Nuclear Emergency Plan is to describe the measures that shall be undertaken to deal with a nuclear emergency caused by the Chalk River Laboratories. This plan deals mainly with actions at the Provincial level and shall by supplemented by the appropriate Municipal Plan. The Townships of Rolph, Buchanan, Wylie, and McKay, the Town of Deep River and the Village of Chalk River are the designated municipalities with respect to CRL. 2 tabs., 5 figs

  4. 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)

  5. 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

  6. 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

  7. Initial field measurements on the Chalk River superconducting cyclotron

    International Nuclear Information System (INIS)

    The midplane magnetic field of the Chalk River superconducting cyclotron has been mapped in detail over the full operating range of 2.5 to 5 tesla. The field measuring apparatus is described and results given include measurements of the field stability, reproducibility and harmonic content. (author)

  8. 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...

  9. 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)

  10. Widespread methanotrophic primary production in lowland chalk rivers.

    Science.gov (United States)

    Shelley, Felicity; Grey, Jonathan; Trimmer, Mark

    2014-05-22

    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 rivers. In 15 of these, the potential for methane oxidation (methanotrophy) was also compared with photosynthesis. In addition, we performed detailed concurrent measurements of photosynthesis and methanotrophy in one large chalk river over a complete annual cycle, where we found methanotrophy to be active to at least 15 cm into the riverbed and to be strongly substrate limited. The seasonal trend in methanotrophic activity reflected that of the riverine methane concentrations, and thus the highest rates were measured in mid-summer. At the sediment surface, photosynthesis was limited by light for most of the year with heavy shading induced by dense beds of aquatic macrophytes. Across 15 rivers, in late summer, we conservatively calculated that net methanotrophy was equivalent to between 1% and 46% of benthic net photosynthetic production within the gravel riverbed, with a median value of 4%. Hence, riverbed chemosynthesis, coupled to the oxidation of methane, is widespread and significant in English chalk rivers. PMID:24695425

  11. 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.

  12. 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

  13. Shipment of 255 DIDO fuel elements to the Savannah River Site to empty the storage and reactor pools at Risoe National Laboratory

    International Nuclear Information System (INIS)

    The DR-3 reactor, owned and operated by the Danish National Laboratory, was built in the late 1950's and initiated operation in January 1960. At that time the DR-1 and DR-2 reactors were already in operation. The main purpose if of Danish research reactor DR-3 was material and fuel testing. Until 1989 the reactor utilized HEU fuel elements. Conversion to the LEU fuel cycle was accomplished in 1990. DOE restarted the return program of for Foreign Research Reactor fuel elements to the United States in 1994. From that time, through 1998, three IUO4 casks (one cask in 1994) operated by Transnucleaire (now named Cogema Logistics, ACL) were used to transport Risoe's fuel to the Savannah River Site (SRS) near Aiken, SC in the USA. In 1999, Risoe elected to issue a request for proposal to transport DR-3 the DIDO fuel elements to SRS with a new licensed cask designed to replace the IUO4 cask. ACL was awarded the contract to transport the irradiated fuel from DR-3 to SRS for the remainder of the FRR Fuel Return program (2009). However, on September 28, 2000, the Board of Governors of Risoe National Laboratory decided to shut down the Danish research reactor of DR3. There had been of R2 technical problems (corrosion on the aluminum reactor tank) and, due to anticipated increasing operational expenses, the Board elected to close the reactor facility. Shortly thereafter, the Danish Government asked the National Laboratory to empty the reactor and its reactor and storage pools containing a total of 255 Dido irradiated fuel elements and ship them to Savannah Rive Site. At that time, ACL was in the process of licensing the new TN-MTR package in the USA. The early shut down of the DR-3 reactor and consequently the resultant new shipping schedule was not compatible with ACL's equipment and licensing schedule for the cask. (author)

  14. Recent developments in waste characterization at Chalk River Nuclear Laboratories

    International Nuclear Information System (INIS)

    The waste characterization program (WCP) at Chalk River Nuclear Laboratories (CRNL) was initiated in 1982 to determine the physical, chemical and radiological properties of wastes intended for disposal in IRUS (Intrusion Resistant Underground Structure), a belowground vault to be constructed at CRNL. During the last year, work on the WCP has centered on determining the radionuclide inventories in candidate wastes for IRUS by gamma-ray monitoring and destructive radiochemical analysis. This paper presents the technical problems associated with monitoring various waste forms (geometry considerations, shielding problems, operating environment, etc.) and also presents details of the destructive radiochemical analysis program

  15. 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)

  16. Canadian fusion breeder blanket program: Irradiation facilities at chalk river*1

    Science.gov (United States)

    Hastings, I. J.; Burton, D. G.; Celli, A.; Delaney, R. D.; Fehrenbach, P. J.; Howe, L. M.; Larson, L. L.; MacEwen, S. R.; Miller, J. M.; Naeem, T. A.; Sawicki, J. A.; Swanson, M. L.; Verrall, R. A.; Zee, R. H.

    1986-11-01

    The major irradiation facility at Chalk River Nuclear Laboratories (CRNL) is the NRU research reactor. Both unvented and vented capsule experiments on candidate blanket ceramics can be performed. In the unvented tests, tritium release data (HT-to-HTO ratio, tritium retention) are obtained by post-irradiation heating of the breeder ceramic in the presence of a sweep gas. Four tests have been completed on Li 2O and LiAlO 2. Effects of sweep gas composition, extraction vessel material and ceramic properties have been determined. Two unvented irradiations under the BEATRIX international breeder exchange program have been completed; analysis is underway. The vented tests involve long-term irradiation of candidate blanket materials. CRITIC-I, scheduled for mid-1986 under BEATRIX, will examine ANL-fabricated Li 2O in a six-month irradiation at 700-1200 K, varying sweep gas composition, with on-line HT/HTO measurement. Additionally, accelerator simulation techniques are available, using 70 kV and 2.0 MV mass separators, a 2.5 MV Van de Graaff accelerator and a tandem accelerator super-conducting cyclotron, the latter allowing irradiation with protons, deuterons or helium at 18-20 MeV.

  17. 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...

  18. Performance of the Chalk River 36Cl AMS system

    International Nuclear Information System (INIS)

    The MP Tandem Injector of the Chalk River TASCC (Tandem Accelerator Superconducting Cyclotron) Facility is being used for 36Cl determinations in studies relating to hydrology and low and high level nuclear waste management. In addition to the accelerator, the computer controlled system comprises a multiple-sample, medium-current ion source, a high resolution injector, a low resolution velocity filter, a gas filled magnet and a Bragg-type particle detection/identification system. Accuracies of 5--10% have been achieved with good suppression of 36S and background levels as low as 5x10-1536Cl/Cl. Following a brief overview of the system, detailed results are presented for the performance of the gas-filled magnet and particle detector as well as for sources of background including ion source memory effects

  19. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D.R.; Chaput, T.; Miller, A.; Wills, C.A., E-mail: leed@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-12-15

    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 ({sup 137}Cs), strontium-90 ({sup 90}Sr), mercury (Hg), and selected organo-chlorine compounds. Concentrations of {sup 137}Cs in the flesh and {sup 90}Sr 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

  1. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Nascimento, Francisco R.T. do; Lima Junior, Carlos A.S.; Oliveira, Andre F.S. de; Affonso, Renato R.W.; Faccini, Jose L.H.; Moreira, Maria L., E-mail: rogerio.tdn@gmail.com, E-mail: souzalima_ca@ien.gov.br, E-mail: oliveira.afelipe@gmail.com, E-mail: raoniwa@yahoo.com.br, E-mail: faccini@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    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)

  3. 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

  4. Field burial results and SIMS analysis of the Chalk River glass blocks

    International Nuclear Information System (INIS)

    In 1959, 25 2-kg hemispherical blocks of aluminosilicate glass, each containing ∼90 MBq/g of mixed fission products, were buried in a sandy soil aquifer in the waste management area at the Chalk River Nuclear Laboratories. A second set of blocks, containing ∼260 MBq/g mixed fission products, was buried in 1960. One block from each test was retrieved in 1978 to undergo chemical and surface analysis. This report reviews the migration of the 90Sr and 137Cs plume in the soil and presents the results of SIMS depth profiling of the surface of a glass block. (author)

  5. Field burial results and SIMS analysis of the Chalk River glass blocks

    International Nuclear Information System (INIS)

    In 1959, 25 2-kg hemispherical blocks of aluminosilicate glass, each containing ∼90 MBq/g of mixed fission products, were buried in a sandy soil aquifer in the waste management area at the Chalk River Nuclear Laboratories. A second set of blocks, containing ∼260 MBq/g mixed fission products, was buried in 1960. One block from each test was retrieved in 1978 to undergo chemical and surface analysis. This report reviews the migration of the /sup 90/Sr and /sup 137/Cs plume in the soil and presents the results of SIMS depth profiling of the surface of a glass block

  6. Livermore pool-type reactor

    International Nuclear Information System (INIS)

    The Livermore Pool-Type Reactor (LPTR) has served a dual purpose since 1958--as an instrument for fundamental research and as a tool for measurement and calibration. Our early efforts centered on neutron-diffraction, fission, and capture gamma-ray studies. During the 1960's it was used for extensive calibration work associated with radiochemical and physical measurements on nuclear-explosive tests. Since 1970 the principal applications have been for trace-element measurements and radiation-damage studies. Today's research program is dominated by radiochemical studies of the shorter-lived fission products and by research on the mechanisms of radiation damage. Trace-element measurement for the National Uranium Resource Evaluation (NURE) program is the major measurement application today

  7. 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

  8. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Malkawi, S.R.; Ahmad, N. E-mail: nasir@pieas.edu.pk

    2002-01-01

    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.

  10. 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

  11. 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

  12. 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

  13. Waste management activities at Chalk River Nuclear Laboratories

    International Nuclear Information System (INIS)

    Low-level radioactive waste-management operations at the Chalk River Nuclear Laboratories (CRNL) of Atomic Energy of Canada Limited began in 1946 and currently include waste processing and interim storage in engineered facilities built in unsaturated sandy overburden. In addition, an R and D program has been underway for about ten years directed at preparations for a transition from the current storage mode to one of permanent disposal for the management of about 5000 m3/a (as-generated volume) of low- and intermediate-level solid wastes generated on the CRNL site or shipped there from the nuclear industry, radioisotope producers and users across Canada. The first phase of the disposal program was the development and demonstration of selected waste processing methods for the volume reduction and immobilization of solid and liquid low-level wastes. This phase is now nearing completion with the construction, commissioning and operation of the CRNL Waste Treatment Centre. The Centre consists of a controlled-air incinerator for combustible solid and liquid wastes, ultrafiltration, reverse-osmosis, and evaporator systems for aqueous wastes, and wipe-film and ribbon-blender bituminizers for immobilizing the ash and waste concentrates. The second phase of the program is directed at further advances in waste characterization and processing, and at the development of two disposal concepts potentially suitable for the local geological situation - Intrusion-resistant shallow land burial and excavated rock cavities at shallow depth. Also included is the preparation of safety-assessment methodologies for the two concepts. The intent is to carry one or both disposal concepts through the constriction and operation of prototype facilities at CRNL as a qualified component of an evolving integrated disposal strategy for the current inventory and future arisings of wastes to be managed

  14. 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)

  15. Analysis of Removal Alternatives for the Heavy Water Components Test Reactor at the Savannah River Site

    International Nuclear Information System (INIS)

    This engineering study was developed to evaluate different options for decommissioning of the Heavy Water Components Test Reactor (HWCTR) at the Savannah River Site. This document will be placed in the DOE-SRS Area reading rooms for a period of 30 days in order to obtain public input to plans for the demolition of HWCTR

  16. Analysis of Removal Alternatives for the Heavy Water Components Test Reactor at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Owen, M.B. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1996-08-01

    This engineering study was developed to evaluate different options for decommissioning of the Heavy Water Components Test Reactor (HWCTR) at the Savannah River Site. This document will be placed in the DOE-SRS Area reading rooms for a period of 30 days in order to obtain public input to plans for the demolition of HWCTR.

  17. ACE - an algebraic compiler and encoder for the Chalk River datatron computer

    International Nuclear Information System (INIS)

    ACE is a program written for the Chalk River Datatron (Burroughs 205) Computer to enable the machine to compile a program for solving a problem from instructions supplied by the user in a notation related much more closely to algebra than to the machine's own code. (author)

  18. WIMS-CRNL: A user's manual for the Chalk River version of WIMS

    International Nuclear Information System (INIS)

    This report describes the preparation of the input for WIMS-CRNL, the Chalk River version of the WIMS lattice code. Also included are notes on the operation of the code, contents of the associated libraries, and the relation of WIMS-CRNL to other versions of the code

  19. Organically bound tritium (OBT) in soil at different depths around Chalk River Laboratories (CRL), Canada

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) Chalk River Laboratories (CRL) is a large nuclear research and test establishment with nuclear and non-nuclear facilities located in Chalk River, Ontario. The CRL Environmental Monitoring Program is designed to demonstrate that radiological exposure resulting from releases from the CRL site remain below the public dose limit specified in the regulations (1 mSv/year). This study was conducted to consolidate environmental effects following a continuous atmospheric tritium release observed at CRL. Soil samples were collected at depths of up to 20 cm using soil probes at the CRL site and surrounding areas. The samples were sectioned at 5 cm intervals, and HTO and OBT concentrations were measured in the samples. Prevailing winds at CRL are from NW and SE, which was suggested to be in close relationship with tritium distribution in environmental samples such as soils and plant leaves. The HTO concentration was the highest in surface soil water and plant leaves at a given sampling point. This result suggests that the concentration of tritium in surface soil water and in plants tissue free water essentially reflects the surrounding atmospheric tritium concentration. OBT concentrations in soil were measured at the historical HT release site, Plant Road, Mattawa Road and three background sites near CRL. The top layer of soil generally had the highest OBT concentration among collected soil samples. This result suggests that OBT concentrations are different from HTO concentrations at the same site and can be representative of previously released environmental tritium at the sampling point. The relationship between the OBT concentration in soil and the amount of tritium released into the environment will be useful for the evaluation of environmental tritium effects and the fate of tritium in the terrestrial ecosystem. The study points out that HTO shows shorter-term dynamic conditions, whereas OBT shows longer-term steady-state conditions

  20. A parameter identifiability study of two chalk tracer tests

    Directory of Open Access Journals (Sweden)

    S. A. Mathias

    2006-08-01

    Full Text Available As with most fractured rock formations, Chalk is highly heterogeneous. Therefore, meaningful estimates of model parameters must be obtained at a scale comparable with the process of concern. These are frequently obtained by calibrating an appropriate model to observed concentration-time data from radially convergent tracer tests (RCTT. Arguably, an appropriate model should consider radially convergent dispersion (RCD and Fickian matrix diffusion. Such a model requires the estimation of at least four parameters. A question arises as to whether or not this level of model complexity is supported by the information contained within the calibration data. Generally modellers have not answered this question due to the calibration techniques employed. A dual-porosity model with RCD was calibrated to two tracer test datasets from different UK Chalk aquifers. A multivariate sensitivity analysis, which assumed only a priori upper and lower bounds for each model parameter, was undertaken. Rather than looking at measures of uncertainty, the shape of the multivariate objective function surface was used to determine whether a parameter was identifiable. Non-identifiable parameters were then removed and the procedure was repeated until all remaining parameters were identifiable.

    It was found that the single fracture model (SFM (which ignores mechanical dispersion obtained the best mass recovery, excellent model performance and best parameter identifiability in both the tests studied. However, there was no objective evidence suggesting that mechanical dispersion was negligible. Moreover, the SFM (with just two parameters was found to be good at approximating the Single Fracture Dispersion Model SFDM (with three parameters when different, and potentially erroneous parameters, were used. Overall, this study emphasises the importance of adequate temporal sampling of breakthrough curve data prior to peak concentrations, to ensure adequate characterisation of

  1. 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)

  2. 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

  3. Overview of research in physics and health sciences at the Chalk River Nuclear Laboratories

    International Nuclear Information System (INIS)

    Toxicology research was a logical extension of existing program at Chalk River. Research in radiotoxicology has been going on there since the early forties. An overview of the existing physics and health sciences research programs operating at the Research Company of Atomic Energy of Canada Limited was presented. Programs in nuclear physics, heavy ion nuclear physics, astrophysical neutrino physics, condensed matter physics, fusion, biology, dosimetry, and environmental sciences were briefly described. In addition, a description of the research company organization was provided

  4. 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

  5. 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

  6. Lithology, fracture intensity, and fracture filling of drill core from Chalk River research area, Ontario

    International Nuclear Information System (INIS)

    In 1977, 1978, and 1979, nine inclined cored boreholes, ranging in length from 113 to 704 m, were drilled in the Chalk River Research Area in order to define the geological subsurface characteristics of the rock mass at several selected test areas. A total of 2,458 metres of NQ-3 and HQ-3 core was obtained from the nine boreholes. Orthogneiss was the most predominant rock type intersected by the boreholes. Pyroxenite, amphibolite, metagabbro and dykes of diabase, pegmatite and aplite were also encountered. The crosscutting relationships and textures within the rocks indicate that the relative ages of the rock units, from youngest to oldest, are diabase; aplite and pegmatite dykes with no defined fabric; pyroxenite; meta-ferrogabbro; amphibolite; aplite and pegmatite dykes and pegmatite pods with a defined fabric; and orthogneiss. Textural characteristics and mineral assemblages indicate that the orthogneisses in the Chalk River Area are a product of regional, medium to high-grade metamorphism and belong to the upper amphibilite to granulite facies. A total of 35,597 fractures (an average of 14.5 fractures per metre) was observed in the core. Brecciated zones and open fractures were noted in the core from all of the boreholes, and major faults were identified in four of the nine boreholes. Nearly all of the fractures have a thickness between 0.4 and 1.2 mm and contain one or more types of filling. Chlorite and calcite are the most common types of filling. Epidote, hematite, clays, sulphides, talc, sericite, and rock fragments also occur in the fractures. The crosscutting relationships between fractures and the sequence of filling layers within the fractures indicate that several episodes of fracturing have occurred and that fractures containing more than one filling have probably been reactivated. A comparison of the geological logs from one of the boreholes with natural gamma, neutron-neutron and magnetic susceptibility logs indicates that certain rock types and

  7. TRR-1/M1 reactor pool refurbishment

    International Nuclear Information System (INIS)

    The pool refurbishment of the TRR-1/M1 is intended to maintain the pool and irradiation facilities in the operable condition prior to the next decade before making decision whether the reactor will be shutdown and decommissioning or used for other purposes. What ever reason the TRR-1/M1 will serve as a training tool for scientists or engineers, and isotopes production or other analytical works for a period of time until the new research reactor will be established. (orig.)

  8. 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)

  9. Stade NPP. Dismantling of the reactor pool

    Energy Technology Data Exchange (ETDEWEB)

    Scharf, Daniel; Dziwis, Joachim [E.ON Anlagenservice GmbH Nukleartechnik, Gelsenkirchen (Germany); Kemp, Lutz-Hagen [KKW Stade GmbH und Co. oHG, Stade (Germany)

    2012-11-01

    Within the scope of the 4{sup th} partial decommissioning permission of Stade NPP the activated and contaminated structures of the reactor pool had to be dismantled in order to gain a completely non-radioactive reactor pool area for the subsequent clearance measurement of the reactor building. In order to achieve the aim it was intended to remove the activated pool liner sheets, its activated framework and several contaminated ventilation channels made of stainless steel, the concrete walls of the reactor pool entirely or in parts depending on their activation level, as well as the remaining activated carbon steel structures of the reactor pool bottom. Embedded in the concrete walls there were several highly contaminated excore tubes and the contaminated pool top edge, which were intended to be removed to its full extent. The contract of the Stade NPP initiated reactor pool dismantling project had been awarded to E.ON Anlagenservice GmbH (EAS) and its subsupplier sat. Kerntechnik GmbH for the concrete dismantling works and was performed as follows. In order to minimize the radiation level in the main working area in accordance with the ALARA principle, the liner sheets and middle parts of its framework were removed by means of angle grinders first, as they were the most dose rate relevant parts. As a result the primary average radiation level in the reactor pool (measured in a distance of 500 mm from the walls) was lowered from 40 {mu}Sv/h to less than 2 {mu}Sv/h. After the minimization of the radiation level in the working area the main dismantling step started with the cutting of the reactor pool walls in blocks by means of diamond rope cutters. Once a concrete block was cut out, it was transported into the fuel pool by means of a crane and crane fork, examined radiologically, marked area by area and segmented to debris by means of an electrical excavator with a hydraulic chisel. Afterwards the debris and carbon steel parts were fractioned and packed for further

  10. Stade NPP. Dismantling of the reactor pool

    International Nuclear Information System (INIS)

    Within the scope of the 4th partial decommissioning permission of Stade NPP the activated and contaminated structures of the reactor pool had to be dismantled in order to gain a completely non-radioactive reactor pool area for the subsequent clearance measurement of the reactor building. In order to achieve the aim it was intended to remove the activated pool liner sheets, its activated framework and several contaminated ventilation channels made of stainless steel, the concrete walls of the reactor pool entirely or in parts depending on their activation level, as well as the remaining activated carbon steel structures of the reactor pool bottom. Embedded in the concrete walls there were several highly contaminated excore tubes and the contaminated pool top edge, which were intended to be removed to its full extent. The contract of the Stade NPP initiated reactor pool dismantling project had been awarded to E.ON Anlagenservice GmbH (EAS) and its subsupplier sat. Kerntechnik GmbH for the concrete dismantling works and was performed as follows. In order to minimize the radiation level in the main working area in accordance with the ALARA principle, the liner sheets and middle parts of its framework were removed by means of angle grinders first, as they were the most dose rate relevant parts. As a result the primary average radiation level in the reactor pool (measured in a distance of 500 mm from the walls) was lowered from 40 μSv/h to less than 2 μSv/h. After the minimization of the radiation level in the working area the main dismantling step started with the cutting of the reactor pool walls in blocks by means of diamond rope cutters. Once a concrete block was cut out, it was transported into the fuel pool by means of a crane and crane fork, examined radiologically, marked area by area and segmented to debris by means of an electrical excavator with a hydraulic chisel. Afterwards the debris and carbon steel parts were fractioned and packed for further treatment

  11. Post-Construction Testing of the Elk River, Hallam and Piqua Power Reactor Plants

    International Nuclear Information System (INIS)

    Actual experience gained during the post-construction testing of three nuclear power plants, under the USAEC Power Reactor Demonstration Program, may permit some generalizations concerning this phase of plant construction and operation. The three plants, Elk River Reactor (ERR), Hallam Nuclear Power Facility (HNPF), and the Piqua Nuclear Power Facility (PNPF), represent three different reactor concepts: natural-circulation boiling water, sodiumgraphite, and organic cooled and moderated, respectively. The post-construction testing period included the time between the end of construction (erection of structures and installation of equipment) and the beginning of power operation (generation of significant net electrical power). The tests were intended to: (a) verify the performance characteristics of the as-installed equipment; (b) obtain initial criticality and reactivity coefficient measurements; and (c) determine reactor physics and plant performance characteristics at a sequence of increasing power levels. .The experience gained can be reported in six separate but interrelated categories: (1) schedule; (2) costs; (3) staffing requirements; (4) procedures; (5) equipment performance (including malfunctions); and (6) actual, as compared to predicted, system performance characteristics. The average project staffing, including craftsmen, operators, supervisors, technical support and trainees, was approximately 50 for ERR, 115 for HNPF, and 60 for PNPF. Detailed written Pre-operational Test Procedures were prepared for each major component and system. To the maximum possible extent, all tests were performed before fuel loading and operation of the integrated plant. Authorization procedures (duplicates of the licensing procedures for non-USAEC-owned plants) were in progress during almost all of the post-construction testing periods. The time required for post-construction testing of each of these plants significantly exceeded the original estimates. The tests disclosed

  12. 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. PMID:26531711

  13. Facilities for Waste Management at Chalk River, Canada

    International Nuclear Information System (INIS)

    The waste disposal areas used by the Atomic Energy of Canada Limited are situated in a rock basin filled with glacial till and sand, draining into the Ottawa River. Low-activity liquid effluent is run into pits in the sand, which are filled with small rocks to prevent contact of liquid with the air. Medium- level liquid is mixed with cement in drums which are stacked and totally enclosed in concrete trenches; medium-level solids are buried in concrete-lined trenches; high-level solids are placed in holes lined with steel or concrete piping. Special facilities are provided for organic liquids and bottled wastes. Details will be given of the structural work and procedures, with an outline of the results of environmental monitoring. (author)

  14. 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

  15. 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.

  16. Advection dispersion modeling of tritium and chloride migration in a shallow sandy aquifer at the Chalk River Laboratories

    International Nuclear Information System (INIS)

    Elevated tritium, helium-3 and chloride concentrations have been measured in groundwaters in a shallow sandy aquifer draining a small lake at the Chalk River Laboratories (CRL), Ontario, Canada. The chloride in the lakewater recharge is 25 times greater than precipitation recharge and forms a continuous, concentrated source of contamination to the aquifer. Tritium (3H) concentrations in both lake and precipitation recharge are elevated owing to the operation of a research reactor on the CRL site and form a continuous spatially distributed source of contamination. The transport of tritium and chloride over the 600 m groundwater flowpath from the lake to the discharge zone are simulated using a 3-D advection-dispersion model. The model requires information on the contaminant input concentrations, the velocity field, dispersion parameters, hydrostratigraphy and boundary conditions. The two independent sets of concentration data provide complementary information to minimize problems associated with the unknown input concentration. The velocity field was estimated from a 3-D simulation of the groundwater flow system; dispersion parameters were estimated from analysis of a controlled natural-gradient tracer test performed previously at the site. The hydrostratigraphy and boundary geometry was characterized by visual logging of borehole sediments, grain size analyses and ground penetrating radar surveys. The abundance of hydrogeologic and geophysical information allowed simulation of the spatial distribution of chloride concentrations with a remarkable degree of accuracy. Simulated and measured peak chloride concentrations differed by less than 15%. The excellent agreement between the simulated and observed chloride concentrations facilitated further modelling of the source and migrational behavior of 3H within this aquifer. We have solved the inverse problem for the 3H source function and successfully modelled the 3H source as a stepwise function. Estimates of

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

    International Nuclear Information System (INIS)

    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)

  18. 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)

  19. 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

  20. Control Rods in high-Flux Swimming-Pool Reactors

    International Nuclear Information System (INIS)

    Control-rod problems in open swimming-pool high-flux and high specific power research reactors are examined in the light of the calibrations and experiments made during the construction of the SILOE reactor. Control-rod operating experience for this reactor at 13 MW is also described. 2. The following are considered in turn: (a) Reactivity balances and reactivity values for the different types of rod tested (cadmium, B4C , rare earths and combinations of these different elements). (b) Flux peaks set up in the core by the presence of the control rods, their incidence on the specific power, the fast fluxes that can be obtained and means of increasing them. (c ) The technological problems involved in constructing the rods. (d) In-pile cooling, vibration, deformation and scram-time problems. 3. In conclusion, current studies on control rods in open swimming-pool reactors operating in the 10 - 30 1W range are briefly summarized. (author)

  1. Calculation and comparisons with measurement of fast neutron fluxes in the material testing facilities of the NRU research reactor

    International Nuclear Information System (INIS)

    The NRU reactor at Chalk River provides three irradiation facilities to study the effects of fast neutrons (E> 1 MeV) on reactor materials for assessing material damage and deformation. The facilities comprise two types of fast neutron rods (Mark 4 and Mark 7), and a Material Test Bundle (MTB) irradiated in a loop site. This paper describes the neutronic simulation of these testing facilities using the WIMS-AECL and TRIAD codes, and comparisons with the fast neutron flux measurements using iron-wire activation techniques. It also provides comparisons of flux levels, neutron spectra, and size limitations of the experimental cavities between these test facilities. (author)

  2. Calculation and comparisons with measurement of fast neutron fluxes in the material testing facilities of the NRU research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Leung, T.C. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    The NRU reactor at Chalk River provides three irradiation facilities to study the effects of fast neutrons (E> 1 MeV) on reactor materials for assessing material damage and deformation. The facilities comprise two types of fast neutron rods (Mark 4 and Mark 7), and a Material Test Bundle (MTB) irradiated in a loop site. This paper describes the neutronic simulation of these testing facilities using the WIMS-AECL and TRIAD codes, and comparisons with the fast neutron flux measurements using iron-wire activation techniques. It also provides comparisons of flux levels, neutron spectra, and size limitations of the experimental cavities between these test facilities. (author)

  3. Magnetic field related mechanical tolerances for the proposed Chalk River superconducting heavy-ion cyclotron

    International Nuclear Information System (INIS)

    A four sector azimuthally varying field cyclotron with superconducting main coils has been proposed as a heavy-ion post-accelerator for the Chalk River MP Tandem van de Graaff. The radial profile of the average axial field will be variable using movable steel trim rods. The field errors due to coil, trim rod and flutter pole imperfections are calculated. Those considered are errors in the axial field, first and second azimuthal harmonic axial fields, transverse field and first azimuthal harmonic transverse field. Such fields induce phase slip, axial or radial coherent oscillations and can result in axial or radial beam instability. The allowed imperfections (tolerances) required to retain stability and maintain acceptably small coherent oscillation amplitudes are calculated. (author)

  4. Pre-operational HTO/HT surveys in the vicinity of the Chalk River Laboratories tritium extraction plant

    International Nuclear Information System (INIS)

    Surveys of the concentrations of HT and HTO in the atmosphere downwind of the Chalk River Laboratories reactor facilities were carried out in 1986 November, and in 1989 March, April and September under different conditions of air temperature, wind direction, and snow or vegetative cover. HT usually amounted to 1-5% of total tritium, but values up to 20% were observed, probably resulting from preferential removal of HTO. In all of the surveys, the greater persistence in the atmosphere of HT than of HTO was evident. The existing levels of HT are such that they will not be augmented significantly by chronic releases from the Tritium Extraction Plant (TEP) when it comes into operation. Hence, operation of the TEP will not facilitate studies of the environmental behaviour of chronically released HT. However, longer term studies of the distribution of HT from the existing facilities would be worthwhile. Soil and vegetation HTO levels in the study area are reported. Further studies of the distribution of tritium between the air, soil and vegetation in areas subjected to chronic exposure would be valuable

  5. Management of Legacy Spent Nuclear Fuel Wastes at the Chalk River Laboratories: The Challenges and Innovative Solutions Implemented - 13301

    International Nuclear Information System (INIS)

    AECL's Fuel Packaging and Storage (FPS) Project was initiated in 2004 to retrieve, transfer, and stabilize an identified inventory of degraded research reactor fuel that had been emplaced within in-ground 'Tile Hole' structures in Chalk River Laboratories' Waste Management Area in the 1950's and 60's. Ongoing monitoring of the legacy fuel storage conditions had identified that moisture present in the storage structures had contributed to corrosion of both the fuel and the storage containers. This prompted the initiation of the FPS Project which has as its objective to design, construct, and commission equipment and systems that would allow for the ongoing safe storage of this fuel until a final long-term management, or disposition, pathway was available. The FPS Project provides systems and technologies to retrieve and transfer the fuel from the Waste Management Area to a new facility that will repackage, dry, safely store and monitor the fuel for a period of 50 years. All equipment and the new storage facility are designed and constructed to meet the requirements for Class 1 Nuclear Facilities in Canada. (authors)

  6. Clinch River Breeder Reactor

    International Nuclear Information System (INIS)

    Mr. Baron says the administration's effort to terminate the Clinch River Breeder Reactor (CRBR) project is symptomatic; they have also placed restrictions on fusion, coal, solar, and other areas of energy development in which technological advances are held back in order to force conservation. Because the breeder reactor, unlike solar and fusion energy, is both economically and technically feasible, a demonstration plant is needed. The contentions that the CRBR design is obsolete, that its proposed size is inappropriate, or that plutonium can be diverted for weapons proliferation are argued to be invalid. Failure to complete the CRBR will have both economic and national security repercussions

  7. 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.

  8. Development of an Integrated Waste Plan for Chalk River Laboratories - 13376

    International Nuclear Information System (INIS)

    To further its Strategic Planning, the Atomic Energy of Canada Limited (AECL) required an effective approach to developing a fully integrated waste plan for its Chalk River Laboratories (CRL) site. Production of the first Integrated Waste Plan (IWP) for Chalk River was a substantial task involving representatives from each of the major internal stakeholders. Since then, a second revision has been produced and a third is underway. The IWP remains an Interim IWP until all gaps have been resolved and all pathways are at an acceptable level of detail. Full completion will involve a number of iterations, typically annually for up to six years. The end result of completing this process is a comprehensive document and supporting information that includes: - An Integrated Waste Plan document summarizing the entire waste management picture in one place; - Details of all the wastes required to be managed, including volume and timings by waste stream; - Detailed waste stream pathway maps for the whole life-cycle for each waste stream to be managed from pre-generation planning through to final disposition; and - Critical decision points, i.e. decisions that need to be made and timings by when they need to be made. A waste inventory has been constructed that serves as the master reference inventory of all waste that has been or is committed to be managed at CRL. In the past, only the waste that is in storage has been effectively captured, and future predictions of wastes requiring to be managed were not available in one place. The IWP has also provided a detailed baseline plan at the current level of refinement. Waste flow maps for all identified waste streams, for the full waste life cycle complete to disposition have been constructed. The maps identify areas requiring further development, and show the complexities and inter-relationships between waste streams. Knowledge of these inter-dependencies is necessary in order to perform effective options studies for enabling

  9. 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)

  10. Safety re-assessment of AECL test and research reactors

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited currently has four operating engineering test/research reactors of various sizes and ages; a new isotope-production reactor Maple-X10, under construction at Chalk River Nuclear Laboratories (CRNL), and a heating demonstration reactor, SDR, undergoing high-power commissioning at Whiteshell Nuclear Research Establishment (WNRE). The company is also performing design studies of small reactors for hot water and electricity production. The older reactors are ZED-2, PTR, NRX, and NRU; these range in age from 42 years (NRX) to 29 years (ZED-2). Since 1984, limited-scope safety re-assessments have been underway on three of these reactors (ZED-2, NRX AND NRU). ZED-2 and PTR are operated by the Reactor Physics Branch; all other reactors are operated by the respective site Reactor Operations Branches. For the older reactors the original safety reports produced were entirely deterministic in nature and based on the design-basis accident concept. The limited scope safety re-assessments for these older reactors, carried out over the past 5 years, have comprised both quantitative probabilistic safety-assessment techniques, such as event tree and fault analysis, and/or qualitative techniques, such as failure mode and effect analysis. The technique used for an individual assessment was dependent upon the specific scope required. This paper discusses the types of analyses carried out, specific insights/recommendations resulting from the analysis, and the plan for future analysis. In addition, during the last four years safety assessments have been carried out on the new isotope-, heat-, and electricity-producing reactors, as part of the safety design review, commissioning and licensing activities

  11. An Innovative Hybrid Loop-Pool Design for Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    The existing sodium cooled fast reactors (SFR) have two types of designs--loop type and pool type. In the loop type design, such as JOYO (Japan) [1] and MONJU (Japan), the primary coolant is circulated through intermediate heat exchangers (IHX) external to the reactor tank. The major advantages of loop design include compactness and easy maintenance. The disadvantage is higher possibility of sodium leakage. In the pool type design such as EBR-II (USA), BN-600M(Russia), Superphenix (France) and European Fast Reactor [2], the reactor core, primary pumps, IHXs and direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) all are immersed in a pool of sodium coolant within the reactor vessel, making a loss of primary coolant extremely unlikely. However, the pool type design makes primary system large. In the latest ANL's Advanced Burner Test Reactor (ABTR) design [3], the primary system is configured in a pool-type arrangement. The hot sodium at core outlet temperature in hot pool is separated from the cold sodium at core inlet temperature in cold pool by a single integrated structure called Redan. Redan provides the exchange of the hot sodium from hot pool to cold pool through IHXs. The IHXs were chosen as the traditional tube-shell design. This type of IHXs is large in size and hence large reactor vessel is needed

  12. 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)

  13. An overview of the waste characterization program at Chalk River Nuclear Laboratories

    International Nuclear Information System (INIS)

    In the last five years, Chalk River Nuclear Laboratories (CRNL) placed 17,000 m3 of wastes into storage (excluding contaminated soil and fill). Almost half of the waste was generated off-site. CRNL is now developing IRUS, an Intrusion Resistant Underground Structure, and the IST, an Improved Sand Trench, to replace storage with safe, permanent disposal. IRUS will be used to dispose of wastes with radiologically hazardous lifetimes between 150 and 500 years duration and the IST will be used for wastes with radiologically hazardous lifetimes of less than 150 years. A comprehensive Waste Characterization Program (WCP) is in place to support disposal projects. The WCP is responsible for (1) specifying the manifests for waste shipments; (2) developing and maintaining central databases for waste inventories and analytical data; and (3) developing the technologies and procedures to characterize the radiological and the physical/chemical properties of wastes. WCP work is being performed under the umbrella of a newly developed waste management quality assurance (QA) program. This paper gives an overview of the WCP with an emphasis on the requirements for determining radionuclide inventories in wastes, for implementing record-keeping systems and for maintaining a QA program for disposal operations

  14. The reduction of sample memory effects in the Chalk River AMS ion source

    International Nuclear Information System (INIS)

    The mechanism underlying Cl and I sample-to-sample interference in the new Chalk River AMS ion source has been studied and compared with the interference observed in an earlier ion source of different internal geometry. The distribution of sputtered material and its degree of migration was measured with the radioactive tracer, 82Br. The temperature dependence of the surface constituents was measured with the elastic recoil detection (ERD) technique and the effect of sample geometry and ion source cleaning was studied with elevated (5 x 10-10) 36Cl/Cl and 129I/I samples. These measurements indicate that a hot (> 350oC) aperture plate ahead of the sample can prevent the sputtering of contaminated regions near the sample. The plate itself remains relatively free of Cl or I itself since these elements or their Cs-gettered compounds are desorbed at this temperature. A small, fixed quantity of Cl or I on this surface is observed, which if sputtered by Cs+ ions, may contribute to ion source memory. Relative sample-to-sample interference for both Cl and I is about 10-3 after 20 min or l0-4 after 60 min. (author)

  15. 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.

  16. MAGS low level waste storage at Atomic Energy of Canada Limited's Chalk River Laboratories

    International Nuclear Information System (INIS)

    The recent introduction of Modular Above Ground Storage (MAGS) constitutes a substantial improvement in the way solid Low-Level Radioactive Waste (LLRW) is handled and stored at AECL's Chalk River Laboratories (CRL). The LLRW generally contains items such as lightly contaminated clothing, paper towels, glassware, used equipment and building materials produced at CRL, or received from Canadian hospitals, universities and other waste generators. These materials, previously stored in unlined sand trenches, are now being stored in a dry, monitored and more easily retrievable state in steel containers in MAGS storage buildings. The MAGS project involved design and construction of three elements: a Waste Handling Building, the first of a series of pre-engineered steel storage buildings, and a new Waste Management Area to house the storage buildings and containers of bulk waste. This project was well received by the local municipalities during the public consultation conducted as a part of the licensing process. The MAGS system entered service in 2002 and has operated satisfactorily since then. A second storage building was completed in 2003. (author)

  17. Determining a pool - type reactor fuel policy

    International Nuclear Information System (INIS)

    Refuelling the 10 to 15 MW pool type reactor considered here will occur frequently (some 10 elements every 3 to 4 weeks). It is therefore necessary to determine the most economic fuel policy. This study proposes to define a strategy that will make it possible to decide on the number and characteristics of the shipment containers, as well as on the means of storage, so as to reduce the risks as much as possible should the basic parameters of the study vary. Among these parameters, the respective influence of which is investigated, chemical reprocessing costs play a vital part. Two examples of optimum fuel management are given according to whether the reprocessing charges applied are those of the old or of the 1961 U.S. AEC base charges for reprocessing highly enriched irradiated fuel. (authors)

  18. Non-electric applications of pool-type nuclear reactors

    International Nuclear Information System (INIS)

    This paper recommends the use of pool-type light water reactors for thermal energy production. Safety and reliability of these reactors were already demonstrated to the public by the long-term operation of swimming pool research reactors. The paper presents the design experience of two projects: Apatity Underground Nuclear Heating Plant and Nuclear Sea-Water Desalination Plant. The simplicity of pool-type reactors, the ease of their manufacturing and maintenance make this type of a heat source attractive to the countries without a developed nuclear industry. (author). 6 figs, 1 tab

  19. Magnetic susceptibility of rocks from boreholes CR-1 to CR-9 at Chalk River

    International Nuclear Information System (INIS)

    Magnetic susceptibility measurements made on rock cores obtained from boreholes at the Chalk River Research Area indicate that the foliated, granitic to granodioritic gneisses are weakly magnetic. Susceptibility values are about 5 x 10-4 S.I., two orders of magnitude less than average values for the Atikokan or Lac du Bonnet granites. Interpretation of the variations recorded in the gneisses in all cores is difficult because the average magnetic susceptibility level is near the limit of resolution of the measuring instrument used. However, in CR-6 and CR-9, mafic units intersect the boreholes and high magnetic susceptibility zones are seen. In CR-9 susceptibility values of the order of 5 x 10-2 S.I. characterize a dyke at depths of 40 to 60 m. A second high-susceptibility zone, distinctly different from the shallower one, is recorded at depths of 580 to 670 m with susceptibility of the orders of 5 x 10-3 S.I. This difference in susceptibility suggests mineralogical differences between the two units. The distinctive susceptibility signatures of these two units are better differentiated than signatures obtained from the other geophysical logs. In CR-6 only one high-susceptibility zone (of the order of 5 x 10-2 S.I.) is recorded in CR-6, at depths of 200 to 290 m. Its signature is similar in shape and intensity to the shallower unit observed in CR-9. Preliminary interpretation suggest continuity between these two zones. In CR-2 and CR-5, significant correlations exist among magnetic susceptibility, temperature anomalies and fracture occurrences. Contrary to observations at other research areas, fracture signatures in these two holes correspond to slight increases in susceptibility values. Alteration products associated with the fractures have higher susceptibility than the gneisses. Even though most of the recorded variations of the magnetic susceptibility are near the detection limit of the measuring instrument, significant features were observed and are discussed in

  20. A compartment model for 90Sr contamination in a wetland at the Chalk River Laboratories

    International Nuclear Information System (INIS)

    Radioactive wastes originating from Canada's nuclear research and development program have been managed at the Atomic Energy of Canada Limited (AECL) Chalk River Laboratories (CRL) since 1946. In 1953 an area called Waste Management Area 'B' (WMA B) was developed to contain low and intermediate level solid waste (LLW and ILW respectively). Initially, all of the wastes were buried in unlined sand trenches or in asphalt lined trenches. These early trenches have been releasing strontium-90 (90Sr) to groundwater since 1954, resulting in an underground contaminant plume. A treatment system was constructed in 1994 and as a result the plume is being intercepted and treated for removal of 90SR. Prior to the establishment of the treatment system the plume extended south and discharged into a watercourse called 'Spring B', then into a wetland area called 'West Swamp'. Routine monitoring of Spring B and the West Swamp outflows for 90Sr has been conducted since the 1960s. A compartment model of the West Swamp was developed and validated against monitoring data for Spring B. The purpose of developing the model was to determine if a standard compartment model could describe 90Sr dynamics in a wetland to support environmental decision making. The model employed mass balance calculations to describe the movement/distribution of 90Sr between the primary system compartments: water/peat, sediment, vegetation and litter. This paper describes how the compartment model was developed and validated. The model can be used as a tool to evaluate remediation alternatives and to provide input to CRL site decommissioning plans. (author)

  1. Stabilization of reactor fuel storage pool-TTP

    International Nuclear Information System (INIS)

    The proposed work includes evaluating standard and improved technologies an designing an integrated demonstration system to clean the water and sludge the fuel storage pools. The water released would meet drinking water standards and tritium standards. The volume of radioactive sludge would be reduced by partial separation of the sludge and radionuclides and eventual solidification of the hazardous and radioactive waste. The scope of the wo includes a survey of needs and applicable technologies, system engineering evaluation, conceptual design, detailed design, fabrication of the integrat demonstration system, and testing of the system. The survey task will locate potential specific customers within the DOE complex, and outside of the DOE complex throughout the United States, that be able to utilize the narrowly focused technology to stabilize/shutdown reactor fuel storage pools, responsible parties will be located and asked respond to a survey about their specific process requirements. Literature searches will be run through technical and scientific databases to locate technologies that may be an improvement over the standard baselined technol for cleanup of radioactively-contaminated pools. Systems engineering will provide decision analysis support for the development, evaluation, design, test functions of the treatment of pool water and sludge

  2. Evaluation of five surface EM techniques for fracture detection and mapping at the Chalk River research area, Ontario

    International Nuclear Information System (INIS)

    Experimental field surveys were carried out with five surface electromagnetic (EM) systems at the Chalk River research area in 1980 and 1981. The purpose of the surveys was to test the usefulness of some new and a few existing EM systems for mapping fracture and shear zones in the highly resistive monzonite gneiss in the survey area. Fractures in the gneiss are often filled with water rich in ions, making the fractures medium to poor-grade conductors. The detection of these fractures by surface electromagnetic surveys depends on their conductance values (product of conductivity (s) and thickness (t)), the conductivity of the host rock, depth of the conductors, and strength, orientation, and frequency of the exciting field. The five EM systems used in the test surveys were: the local loop VLF-EM system; Max-Min II horizontal loop EM system; geonics EM-34 ground-mapping system; geonics EM-37; and, Maxi-Probe systems. Analysis of the field results shows that: the local loop VLF transmitter is a suitable substitute for Navy VLF stations; VLF and Max-Min II systems can detect poor conductors coincident with geologically defined lineaments and fracture zones, however, it is difficult at the moment to distinguish between near-vertical conductors in the bedrock and conductive overburden filling depressions in the bedrock surface; geonics EM-34 is not suitable in highly resistive terrain or in the presence of near-vertical conductors; and, the geonics EM-37 and Maxi-Probe systems detected possible subhorizontal conductors at depths of over 200 m. The presence of these could not be verified because of the lack of borehole control in those areas. An evaluation of the five systems indicates that VLF-EM and Max-Min II systems are most useful for detection of vertical and near-vertical fracture zones in the top 100 m, while EM-37 and Maxi-Probe systems are useful for detecting fracture zones at greater depths

  3. Corrosion of aluminium alloy test coupons in the WWR-K reactor cooling pool and wet storage tank in Almaty, Kazakhstan

    International Nuclear Information System (INIS)

    The corrosion of a number of aluminium alloy coupons was studied. The coupons were assembled in racks and exposed to water in the storage pools for spent nuclear fuels of the WWR-K reactor in Almaty, Kazakhstan. The maximum duration of exposure of the racks was 921 days. Mass loss of the coupons, depth of pits and the average pit sizes on the coupons were determined. The data was evaluated and compared as a function of coupon position in the rack, time of exposition and nature of contact between coupons (none, crevice or bimetallic). (author)

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

    International Nuclear Information System (INIS)

    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

  5. 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.

  6. Spatial analysis of Carbon-14 dynamics in a wetland ecosystem (Duke Swamp, Chalk River Laboratories, Canada)

    International Nuclear Information System (INIS)

    A detailed survey was conducted to quantify the spatial distribution of 14C in Sphagnum moss and underlying soil collected in Duke Swamp. This wetland environment receives 14C via groundwater pathways from a historic radioactive Waste Management Area (WMA) on Atomic Energy Canada Limited (AECL)'s Chalk River Laboratories (CRL) site. Trends in 14C specific activities were evaluated with distance from the sampling location with the maximum 14C specific activity (DSS-35), which was situated adjacent to the WMA and close to an area of groundwater discharge. Based on a spatial evaluation of the data, an east-to-west 14C gradient was found, due to the influence of the WMA on 14C specific activities in the swamp. In addition, it was possible to identify two groups of sites, each showing significant exponential declines with distance from the groundwater source area. One of the groups showed relatively more elevated 14C specific activities at a given distance from source, likely due to their proximity to the WMA, the location of the sub-surface plume originating from the WMA, the presence of marsh and swamp habitat types, which facilitated 14C transport to the atmosphere, and possibly, 14C air dispersion patterns along the eastern edge of the swamp. The other group, which had lower 14C specific activities at a given distance from the groundwater source area, included locations that were more distant from the WMA and the sub-surface plume, and contained fen habitat, which is known to act as barrier to groundwater flow. The findings suggest that proximity to source, groundwater flow patterns and habitat physical characteristics can play an important role in the dynamics of 14C being carried by discharging groundwater into terrestrial and wetland environments. - Highlights: • Groundwater represents an important source of volatile radionuclides to wetlands. • Habitat type influenced 14C transport from sub-surface to surface environments. • C-14 specific activity

  7. 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...

  8. 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.

  9. Reactor TRIGA PUSPATI (RTP) spent fuel pool conceptual design

    International Nuclear Information System (INIS)

    Reactor TRIGA PUSPATI (RTP) is the one and only research reactor in Malaysia that has been safely operated and maintained since 1982. In order to enhance technical capabilities and competencies especially in nuclear reactor engineering a feasibility study on RTP power upgrading was proposed to serve future needs for advance nuclear science and technology in the country with the capability of designing and develop reactor system. The need of a Spent Fuel Pool begins with the discharge of spent fuel elements from RTP for temporary storage that includes all activities related to the storage of fuel until it is either sent for reprocessed or sent for final disposal. To support RTP power upgrading there will be major RTP systems replacement such as reactor components and a new temporary storage pool for fuel elements. The spent fuel pool is needed for temporarily store the irradiated fuel elements to accommodate a new reactor core structure. Spent fuel management has always been one of the most important stages in the nuclear fuel cycle and considered among the most common problems to all countries with nuclear reactors. The output of this paper will provide sufficient information to show the Spent Fuel Pool can be design and build with the adequate and reasonable safety assurance to support newly upgraded TRIGA PUSPATI TRIGA Research Reactor. (author)

  10. Low-level radioactive river sediment particles originating from the Chalk river nuclear site carry a mixture of radionuclides and metals

    Energy Technology Data Exchange (ETDEWEB)

    Lind, Ole Christian; Cagno, Simone; Salbu, Brit [Norwegian University of Life Sciences - NMBU, Center of Excellence in Environmental Radioactivity - CERAD, P.O. Box 5003, N-1432 Aas (Norway); Falkenberg, Gerald [Photon science, DESY, Hamburg (Germany); Janssens, Koen; Nuyts, Gert; Vanmeert, Frederik [AXIL, Department of Chemistry, University of Antwerpen (Belgium); Jaroszewicz, Jakub [Faculty of Materials Science and Engineering, Warsaw University of Technology, Warsaw (Poland); Priest, Nicholas D.; Audet, Marc [Nuclear Science Division, AECL Chalk River Laboratories (Canada)

    2014-07-01

    The Chalk River Laboratory of Atomic Energy of Canada Ltd., site is located on the Ottawa River approximately 200 km northwest of Ottawa, Canada. The site has two large research reactors: NRX, which operated from 1947 to 1991 and NRU, which continues to operate and is used to produce a significant fraction of the world's supply of medical isotopes. During the course of the operation of the NRX reactor small quantities of radioactive particles were discharged to the Ottawa River through a process sewer discharge pipe. These are now located in river bed sediments within a 0.08 km² area close to the discharge pipe. In the present study, selected particles were isolated from riverbed sediments. These were then characterized by environmental scanning electron microscopy with energy dispersive micro X-ray analysis (ESEM-EDX). This was undertaken to obtain information on particle size, structure and the distribution of elements across particle surfaces. Based on the results of ESEM-EDX, particles were selected for X-ray absorption nano-tomography analysis, which provides videos showing the 3D density distribution of the particles. Furthermore, 2D and 3D Synchrotron Radiation based X-ray techniques (micro-X-ray fluorescence; micro-XRF, micro-X-ray absorption near edge spectroscopy; micro-XANES and micro-X-ray diffraction; micro-XRD) with submicron resolution (beam size 0.5 μm) were employed to investigate the elemental and phase composition (micro-XRF/XRD) and oxidation states (micro-XANES) of matrix elements with high spatial resolution and sensitivity. Results show that the particles investigated so far varied according to: 1) <~40 μm diameter sized U fuel particles similar in structure to particles observed from Chernobyl and Krasnoyarsk-26 and 2) larger particles with diameters up to several hundred μm. The larger particles comprised a matrix of low density, sediment material with high density inclusions that contained a range of metals including Cu, Cr, As

  11. 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.

  12. Assessing inventories of past radioactive waste arisings at Chalk River Laboratories

    International Nuclear Information System (INIS)

    Internationally, a great deal of progress has been made in improving the management of currently accumulating and anticipated future radioactive wastes. Progress includes improved waste collection, segregation, characterization and documentation in support of disposal facility licensing and operation. These improvements are not often very helpful for assessing the hazards of wastes collected prior to their implementation, since, internationally, historic radioactive wastes were not managed and documented according to today's methods. This paper provides an overview of Atomic Energy of Canada Limited's (AECL) unique approach to managing its currently accumulating, low-level radioactive wastes at Chalk River Laboratories (CRL) and it describes the novel method AECL-CRL has developed to assess its historic radioactive wastes. Instead of estimating the characteristics of current radioactive wastes on a package-by-package basis, process knowledge is used to infer the average characteristics of most wastes. This approach defers, and potentially avoids, the use of expensive analytical technologies to characterize wastes until a reasonable certainty is gained about their ultimate disposition (Canada does not yet have a licensed radioactive waste disposal facility). Once the ultimate disposition is decided, performance assessments determine if inference characterization is adequate or if additional characterization is required. This process should result in significant cost savings to AECL since expensive, resource-intensive, up-front characterization may not be required for low-impact wastes. In addition, as technological improvements take place, the unit cost of characterization usually declines, making it less expensive to perform any additional characterization for current radioactive wastes. The WIP-III data management system is used at CRL to 'warehouse' the average characteristics of current radioactive wastes. This paper describes how this 'warehouse of information

  13. 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…

  14. 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.

  15. 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.

  16. Convective cooling in a pool-type research reactor

    International Nuclear Information System (INIS)

    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

  17. 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

  18. Plenum separator system for pool-type nuclear reactors

    International Nuclear Information System (INIS)

    A pool-type liquid-metal-heat-transfer nuclear reactor is described which consists of: a vertically disposed generally cylindrical reactor vessel having a closed bottom portion, and a closure head atop the reactor vessel and closing the reactor vessel; the reactor vessel enclosing the major components of the nuclear reactor which include a reactor core supported at a centrally disposed lower portion of the reactor vessel; a bottom-supported gas-plenum-forming hollow cylindrical member closed at its upper end, the hollow cylindrical member sealed to and supported by the reactor vessel; a hollow cylindrically conformed neutron shield member spaced from and radially surrounding the reactor core; separate liquid-metal plena confining liquid metal during normal reactor operation and comprising a hot upper plenum, a cold lower plenum and intermediate temperature plena; the liquid metal intakes of the liquid metal pumps positioned in the cold lower plenum with the cooler liquid metal therein being pumped upwardly through the reactor core to be heated and exit therefrom in a turbulent fashion; and the liquid metal intakes of the heat exchangers positioned within the hot upper plenum and the liquid metal discharges of the heat exchangers positioned within the cold lower plenum to discharge cooled liquid metal into the cold lower plenum

  19. Test Area North Pool Stabilization Project: Environmental assessment

    International Nuclear Information System (INIS)

    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 fuelsclose 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

  20. 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)

  1. RELAP5 Analysis of the Hybrid Loop-Pool Design for Sodium Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hongbin Zhang; Haihua Zhao; Cliff Davis

    2008-06-01

    An innovative hybrid loop-pool design for sodium cooled fast reactors (SFR-Hybrid) has been recently proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to improve economics and safety of SFRs. In the hybrid loop-pool design, primary loops are formed by connecting the reactor outlet plenum (hot pool), intermediate heat exchangers (IHX), primary pumps and the reactor inlet plenum with pipes. The primary loops are immersed in the cold pool (buffer pool). Passive safety systems -- modular Pool Reactor Auxiliary Cooling Systems (PRACS) – are added to transfer decay heat from the primary system to the buffer pool during loss of forced circulation (LOFC) transients. The primary systems and the buffer pool are thermally coupled by the PRACS, which is composed of PRACS heat exchangers (PHX), fluidic diodes and connecting pipes. Fluidic diodes are simple, passive devices that provide large flow resistance in one direction and small flow resistance in reverse direction. Direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) are immersed in the cold pool to transfer decay heat to the environment by natural circulation. To prove the design concepts, especially how the passive safety systems behave during transients such as LOFC with scram, a RELAP5-3D model for the hybrid loop-pool design was developed. The simulations were done for both steady-state and transient conditions. This paper presents the details of RELAP5-3D analysis as well as the calculated thermal response during LOFC with scram. The 250 MW thermal power conventional pool type design of GNEP’s Advanced Burner Test Reactor (ABTR) developed by Argonne National Laboratory was used as the reference reactor core and primary loop design. The reactor inlet temperature is 355 °C and the outlet temperature is 510 °C. The core design is the same as that for ABTR. The steady state buffer pool temperature is the same as the reactor inlet

  2. Design guide for Category III reactors: pool type reactors

    International Nuclear Information System (INIS)

    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

  3. 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.

  4. 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...

  5. 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...

  6. 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...

  7. 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...

  8. 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...

  9. Comments on nuclear reactor safety in Ontario

    International Nuclear Information System (INIS)

    The Chalk River Technicians and Technologists Union representing 500 technical employees at the Chalk River Nuclear Laboratories of AECL submit comments on nuclear reactor safety to the Ontario Nuclear Safety Review. Issues identified by the Review Commissioner are addressed from the perspective of both a labour organization and experience in the nuclear R and D field. In general, Local 1568 believes Ontario's CANDU nuclear reactors are not only safe but also essential to the continued economic prosperity of the province

  10. Borehole television surveys and acoustic televiewer logging at the National Hydrology Research Institute's hydrogeological research area in Chalk River, Ontario

    International Nuclear Information System (INIS)

    This paper presents the results of studies of the fracture distribution and fracture orientation encountered in 23 boreholes at the National Hydrology Research Institute's Hydrogeological Research Area in Chalk River. Borehole television camera and acoustic televiewer data were used to determine: the fracture distribution as a function of depth; the aperture distribution as a function of depth; and, the predominant fracture sets. Fracture frameworks of the rock mass were constructed based on these data. The rock mass was found to be moderately to well fractured. Many open fractures were detected in the CR-series boreholes, especially in the upper 70 m. Three interconnecting, highly fractured zones are intersected by the CR-series boreholes between elevations of 96.80 m and 56.56 m, 34.91 m and -22.42 m, and -56.56 m and -61.31 m, while two interconnecting fracture zones are encountered by the FS-series boreholes between elevations 103.89 m and 96.64 m, and 108.20 m and 95.08 m. The predominant fracture sets strike east-west, northwest and indeterminably subhorizontal. The predominant set of veins strikes to the south. The subsurface fracture patterns and vein patterns were found to be similar to those observed on surface outcrops

  11. 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

  12. Hydrogeochemical processes affecting the migration of radionuclides in a fluvial sand aquifer at the Chalk River Nuclear Laboratories

    International Nuclear Information System (INIS)

    In the mid-1950's two experimental disposals of liquid radioactive waste containing about 700 curries of strontium-90 and cesium-137 were made into pits in sandy ground at one of the disposal areas at Chalk River Nuclear Laboratories. Since then, the wastes have migrated into two nearby aquifers and have chromatographically separated into strontium-90 and cesium-137 plumes moving at velocities less than that of the transporting groundwater. Analysis of radioactively contaminated aquifer sediments showed that most of the strontium-90 is exchangeably adsorbed, primarily to feldspars and layer silicates (mainly biotite); the rest is either specifically adsorbed to iron (III) and perhaps manganese (IV) oxhydroxides or fixed to unknown sinks. Less than one half of adsorbed cesium-137 is exchangeable with 0.5 m calcium chloride; the high levels of cesium-137 adsorption and fixation are probably due to its reaction with micaceous minerals. Complexation of strontium-90 and cesium-137 does not appear to be an important factor affecting their transport or adsorption. In studies of groundwater quality or pollution, dissolved oxygen and sulfide should be measured in addition to the redox potential since it allows independent assessment of the redox levels. The latter were found to affect the mobility of multivalent transition metals and nonmetals. (DN)

  13. Experience on Maintenance of Thai Research Reactor's 'Small-Section' Pool

    International Nuclear Information System (INIS)

    The reactor pool of TRR-1/M1 has been used since 1962 when the reactor building was constructed. Periodic maintenance of the reactor pool has been conducted by cleaning the pool surface and re-painting with epoxy coating. The TRR-1/M1 pool basically consists of two sections referred as 'large-section' and 'small-section'. The latest re-painting activity of the 'large-section' pool was performed in 2006 but the 'small-section' pool had not been re-painted for more than 10 years. Therefore, to assure that the 'small-section' pool can maintain leak-proof condition, the re-painting of the 'small-section' pool was performed in the early 2012. A project team was organized specially for this project and a detailed execution plan was developed. The project activities include removing foreign objects and highly activated materials from the pool section, cleaning, inspecting, re-painting the pool surface and testing for water leaks. Preparation of the repainting activities had begun 2 years in advance. During the time, the reactor core had been relocated to operate in the large-section pool away from the working area in order to minimize radioactivity. The challenge of this project was to handle 4 sets of highly radioactive bolts and nuts which support the weight of the 'void tank' irradiation facility. These bolts and nuts were made from stainless steel and had been in the flux region since the installation of the 'void tank' irradiation facility approximately 30 years ago. Dose rate measurement at the contacts of these bolts and nuts were found to be in the range of 10 . 20 R/hr. The strategy to minimize the dose rate of the workers to conduct the pool repainting in the area was to remove the bolts and nuts and replace with new ones before entering the area. Special tools were improvised in order to remove the bolts and nuts under water. During the execution of the project, close radiation monitoring was performed by the radiation protection team. The project was conducted

  14. 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

  15. Challenges and lessons-learned during the reactor pool repair at the Penn State Breazeale Reactor

    International Nuclear Information System (INIS)

    On October 10, 2007, operators of the Penn State Breazeale Reactor (PSBR) observed the reactor pool level had decreased more than expected over the weekend. Upon further investigation, the staff confirmed that a small leak of 10 gallons had developed in the 52 year old unlined pool. The staff immediately informed the appropriate regulatory authorities and set about finding a solution. Over the next six weeks, the reactor staff worked with University personnel, contractors and regulators to fix the leak and return the facility to normal operation. The Penn State reactor was the first university research reactor licensed by the Atomic Energy Commission. The facility has had several minor pool leaks since its construction in 1955, with the latest occurring in 1976. Each time, the leak was located and repaired with concrete and an epoxy coating. The 2007 leak repair was more extensive involving three-step process that required hydro-lazing the pool wall, removing old sealant, and covering the areas with an epoxy concrete. After these processes the entire pool wall surface was sealed with a polyurea coating. The response from the State of Pennsylvania and US-NRC regulatory authorities was much more involved than earlier leak events. Although public risk was never an issue, the US-NRC immediately dispatched inspectors to the facility so that senior officials could be knowledgeable and responsive to the public's information needs. Additionally, the University set up a team to provide the public and news organizations with ongoing status of the investigation and repair activities. This team allowed the reactor staff to remain focused on the technical aspects of the repair and interface with the regulators. The PSBR pool leak detection and repair will be discussed. Also, PSBR administration and staff, other PSBR functions, coordination with US-NRC and the state of Pennsylvania officials will be presented and experiences gained from this event will be shared with the other

  16. OE Management at Research Technology Operations, Chalk River Laboratories, AECL, Canada

    International Nuclear Information System (INIS)

    Brief description of nuclear facility. A nuclear installation consisting of a 130 MW research reactor and 13 licensed nuclear facilities, staffed by ∼2600 employees, on three distinct sites. Main activities include: (1) Reactor development; (2) CRL nuclear operations; (3) Research and development; (4) Isotope production; (5) Waste management and decommissioning. Overview of OE arrangements. A centralized OE group that is permanently resourced and trained to support the organization. The group is spread over two time zones and supported by a cadre of permanently dedicated OE coordinators and action tracking coordinators throughout the organization

  17. 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)

  18. Shielding Calculations for The New Spent Fuel Storage Pool of Etrr1 Reactor

    International Nuclear Information System (INIS)

    MCNP code was used to model and simulate the new spent fuel storage pool of Etrr1 research reactor. Shielding calculations for the pool were performed to calculate the radiation dose through different pool layers. Radiation sources for photons and neutrons inside the pool were determined under different conditions. Key parameters that affect the radiation dose outside the pool were studied. Comparison with the designer values was performed, agreement and disagreement were investigated. Radiation safety of the pool has been verified

  19. 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

  20. 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.

  1. 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.

  2. 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)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  5. Refurbishment of Pakistan research reactor (PARR-1) for stainless steel lining of the reactor pool

    International Nuclear Information System (INIS)

    Pakistan Research Reactor-1 (PARR-1) is a pool-type research reactor. Reactor aging has resulted in the increase of water seepage from the concrete walls of the reactor pool. To stop the seepage, it was decided to augment the existing pool walls with an inner lining of stainless steel. This could be achieved only if the pool walls could be accessed unhindered and without excessive radiation doses. For this purpose a partial decommissioning was done by removing all active core components including standard/control fuel elements, reflector elements, beam tubes, thermal shield, core support structure, grid plate and the pool's ceramic tiles, etc. An overall decommissioning program was devised which included procedures specific to each item. This led to the development of a fuel transport cask for transportation, and an interim fuel storage bay for temporary storage of fuel elements (until final disposal). The safety of workers and the environment was ensured by the use of specially designed remote handling tools, appropriate shielding and pre-planned exposure reduction procedures based on the ALARA principle. During the implementation of this program, liquid and solid wastes generated were legally disposed of. It is felt that the experience gained during the refurbishment of PARR-1 to install the stainless steel liner will prove useful and better planning and execution for the future decommissioning of PARR-1, in particular, and for other research reactors like PARR-2 (27 kW MNSR), in general. Furthermore, due to the worldwide activities on decommissioning, especially those communicated through the IAEA CRP on 'Decommissioning Techniques for Research Reactors', the importance of early planning has been well recognized. This has made possible the implementation of some early steps like better record keeping, rehiring of trained manpower, and creation of interim and final waste storage. (author)

  6. Comparison of pool/loop configurations in the JAEA fast reactor feasibility study

    International Nuclear Information System (INIS)

    JAEA conducted a feasibility study on commercialized fast breeder reactor cycle systems from 1999 to 2006 (FS). In the FS, various fast reactor concepts with various power levels, coolant materials and plant configurations were proposed and competed. In the large-scale sodium cooled reactor region, four nuclear industry vendors proposed each original advanced sodium cooled reactor concept in 1999. One was a loop concept which was named 'JSFR' later and the other three were pool concepts. The first competition among the four concepts showed that the economical competitiveness of JSFR is better than pool concepts. Therefore, in the 2000 study, one pool concept was selected and the selected pool concept was refined to compete with JSFR. In this appendix, the selected FS pool concept is described and the pool/loop comparative study in the FS is briefly summarized. Schematic illustration of the reactor vessel of the FS pool is shown. Relation of reactor vessel diameters and electric output of various sodium-cooled reactor concepts are compared. The comparison shows that the FS pool concept has a smaller reactor vessel than recent conceptual large reactors such as SPX-2, SNR-2, EFR and BN-1600. The material amount comparison shows that the reactor vessel and primary system material amount of the FS pool concept is heavier than that of the FS loop concept (JSFR) by approximately 250ton. Because the reactor vessel diameter comparison of various concepts shows that the FS pool concept is one of the most compact pool configurations, the FS pool/loop competition is thought to provide comparative information between the most economical loop and pool concepts

  7. Steam blowdown experiments with the condensation pool test rig

    International Nuclear Information System (INIS)

    During a possible loss-of-coolant accident (Local) a large amount of non-condensable (nitrogen) and condensable (steam) gas is blown from the upper drywell of the containment to the condensation pool through the blowdown pipes at the boiling water reactors (BWRs). The wet well pool serves as the major heat sink for condensation of steam. The blowdown causes both dynamic and structural loads to the condensation pool. There might also be a risk that the gas discharging to the pool could push its way to the emergency core cooling systems (ECCS) and undermine their performance. (author)

  8. Radiochemistry Lab Decommissioning and Dismantlement. AECL, Chalk River Labs, Ontario, Canada

    International Nuclear Information System (INIS)

    Atomic Energy of Canada (AECL) was originally founded in the mid 1940's to perform research in radiation and nuclear areas under the Canadian Defense Department. In the mid 50's The Canadian government embarked on several research and development programs for the development of the Candu Reactor. AECL was initially built as a temporary site and is now faced with many redundant buildings. Prior to 2004 small amounts of Decommissioning work was in progress. Many reasons for deferring decommissioning activities were used with the predominant ones being: 1. Reduction in radiation doses to workers during the final dismantlement, 2. Development of a long-term solution for the management of radioactive wastes in Canada, 3. Financial constraints presented by the number of facilities shutdown that would require decommissioning funds and the absence of an approved funding strategy. This has led to the development of a comprehensive decommissioning plan that is all inclusive of AECL's current and legacy liabilities. Canada does not have a long-term disposal site; therefore waste minimization becomes the driving factor behind decontamination for decommissioning before and during dismantlement. This decommissioning job was a great learning experience for decommissioning and the associated contractors who worked on this project. Throughout the life of the project there was a constant focus on waste minimization. This focus was constantly in conflict with regulatory compliance primarily with respect to fire regulations and protecting the facility along with adjacent facilities during the decommissioning activities. Discrepancies in historical documents forced the project to treat every space as a contaminated space until proven differently. Decommissioning and dismantlement within an operating site adds to the complexity of the tasks especially when it is being conducted in the heart of the plant. This project was very successful with no lost time accidents in over one hundred

  9. 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.

  10. Operational and research activities of Tsing Hua open pool reactor

    International Nuclear Information System (INIS)

    Tsing Hua Open Pool Reaction (THOR) is the first nuclear reactor to become operational in Taiwan. It reached its first critical on April 13, 1961. Until now, THOR has been operated successfully for 27 years. The major missions of THOR include radioisotope production, neutron activation analysis, nuclear science and engineering researches, education, and personnel training. The THOR was originally loaded with HEU MTR-type fuels. A gradual fuel replacing program using LEU TRIGA fuel to replace MTR started in 1977. By 1987, THOR was loaded with all TRIGA fuels. This paper gives a brief history of THOR, its current status, the core conversion work, some selected research topics, and its improvement plan. (author)

  11. The macroinvertebrate fauna of pools in the floodplain (fadama) of the Anambra River, Nigeria

    OpenAIRE

    Eyo, Joseph; Ekwonye, Uchenna

    1995-01-01

    The Anambra River is the largest tributary of the lower Niger River below Lukoja. Between the months of May and November the river is subject to seasonal flooding from heavy precipitation and land runoff into the drainage system. During the flood phase, pools form on the floodplains (known as the fadama) and these pools receive materials and biota from the main river channel. The biota often includes representatives of freshwater vertebrates (including fishes) and invertebrates. On this brief...

  12. 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.

  13. 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.

  14. Commissioning of the Open Pool Australian Lightwater (OPAL) research reactor - A health physics perspective

    International Nuclear Information System (INIS)

    During 2006 and 2007 the Australian Nuclear Science and Technology Organisation (ANSTO) commissioned OPAL, a 20 MW open pool Research Reactor. This commissioning involved three stages; Stage A, testing the reactors systems prior to fuel loading, Stage B, first loading of fuel, achieving first criticality, reactor characterisation and systems testing up to 400kW, and Stage C, raising the power of the reactor in steps, up to its full operating power of 20MW. Prior to and following fuel loading a series of radiation measurements were made throughout the plant. These included dose rates, radioactivity in air, on surfaces and in cooling and shielding liquids. Installed continuous monitoring and portable equipment were used. Health physics measurements were repeated at increasing reactor power levels to check engineering design features and design of plant for radiation protection aspects. A Radiation Protection Plan and associated monitoring programs were implemented, including establishing and maintaining area, task and personnel monitoring regimes in the facility. Health Physics assessments and advice at each stage, played a major role in this commissioning process. This paper discusses the health physics experience of commissioning the OPAL Research Reactor and describes health physics results, actions taken and lessons learned during commissioning. (author)

  15. Commissioning of the Open Pool Australian Light water (OPAL) research reactor: a health physics perspective

    International Nuclear Information System (INIS)

    During 2006 and 2007 the Australian Nuclear Science and Technology Organisation (ANSTO) commissioned OPAL, a 20 MW open pool Research Reactor. This commissioning involved three stages: Stage A: testing the reactors systems prior to fuel loading; Stage B: first loading of fuel, achieving first criticality, reactor characterisation and systems testing up to 400 kw; and Stage C: raising the power of the reactor in steps, up to its full operating power of 20 MW. Prior to and following fuel loading a series of radiation measurements were made throughout the plant. These included dose rates, radioactivity in air, on surfaces and in cooling and shielding liquids. Installed continuous monitoring and portable equipment were used. Health physics measurements were repeated at increasing reactor power levels to check engineering design features and design of plant for radiation protection aspects. A Radiation Protection Plan and associated monitoring programs were implemented, including establishing and maintaining area, task and personnel monitoring regimes in the facility. Health Physics assessments and advice at each stage, played a major role in this commissioning process. This paper discusses the health physics experience of commissioning the OPAL Research Reactor and describes health physics results, actions taken and lessons learned during commissioning. (author)

  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. Criticality calculations for the spent fuel storage pools for Etrr1 and Etrr2 reactors

    International Nuclear Information System (INIS)

    A criticality analysis of two spent fuel storage pools for Etrr1 and Etrr2 research reactors was performed. The multiplication factor for the pools was calculated as a function of relevant lattice physics parameters. Monte Carlo code MNCP-4A code was used in the criticality calculations. The results were compared with those given by CITATION code and results obtained formerly during the design phase of the pools with the MONK 6.3 code. Safety of the pools was confirmed. (author)

  18. Hot Water Layer and Thermal Stratification in an Open-pool type Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yong-Seok; Yoon, Hyun-Gi; Choi, Jeongwoon; Kim, Seong-Hoon; Chi, Dae-Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    In many open-pool type research reactors, a hot water layer is introduced in the upper part of the pool as a shielding layer to reduce the radiation level on the pool top. By maintaining the hot water layer in a properly higher temperature than the lower part of the pool, a thermally stratified region is developed below the hot water layer and the flows in the lower part of the pool is successfully isolated from the upper part of the pool. This reduces a mass transport from the lower part of the pool to the pool upper part and consequently the radioactivity level on the pool top is also diminished. In this study, the characteristics of the hot water layer and the thermally stratified region in the pool of the KIJANG Research Reactor (KJRR) are investigated. Numerical simulation on a 3D simplified model of the pool of KJRR is conducted using the commercial CFD software ANSYS FLUENT 13.0. The results show initial time evolutions of the temperatures and the flow velocities in the pool toward each quasi steady state. For the shielding analysis in the pool which is required to estimate the radioactivity of the hot water layer, the mixing rate between the hot water layer and the lower part of the pool is important variable as well as the thicknesses of the hot water layer and the stratified region. In further study, the mixing rate will be estimated by post processing the obtained results.

  19. 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

  20. 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.

  1. 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.

  2. Chalk as a reservoir

    DEFF Research Database (Denmark)

    Fabricius, Ida Lykke

    chalk intervals are to some extent cemented and cannot compact mechanically at realistic effective stresses and only deform elastically. All chalk intervals though, may react by fracturing to changes in shear stress. So where natural fractures are not prevalent, fractures may be generated hydraulically....... Fractures play a significant role in the production of hydrocarbons from chalk reservoirs....

  3. RUTA pool-type reactor for heat supply and the possibility for its application area expansion

    International Nuclear Information System (INIS)

    RUTA, a reactor facility with a pool-type reactor, has been designed for heat supply of residential districts. A relatively low potential of the heat generated by the reactor requires a special approach to building up heat supply systems with RUTA facilities. The application of the RUTA facility as a heat source for seawater thermal distillation has been considered. It is possible to use the reactor for neutron therapy. The reactor optimization provides for the improvement of the facility's consumer qualities. (author)

  4. 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)

  5. Full scale steady state component tests of the SWR 1000 fuel pool cooler at the INKA test facility

    International Nuclear Information System (INIS)

    The SWR 1000 fuel pool coolers are tubular heat exchangers. They are installed on the fuel pool wall around the spent fuel storage racks. Fuel pool water is cooled by means of natural convection. Forced circulation flow of closed-cooling water exists on the tube side of each heat exchanger. The penetrations of the cooling water supply lines through the fuel pool linear are all located above the pool water surface. This ensures that the fuel pool cannot lose water in the event of a pipe break. Integration of the cooling components inside the fuel pool ensures only non-contaminated piping within the reactor building. The fuel pool cooling system consists of two redundant cooling trains. Each cooling train comprises four heat exchangers connected in parallel. The system must ensure adequate heat removal both during normal plant operation and in the event of any postulated accident. To verify proper functioning of the component, full-scale, steady-state tests were performed at the INKA (Integral Teststand Karlstein) test facility in Karlstein Germany. The characteristic diagram for heat transfer capacity of the component as a function of cooling water temperature and fuel pool water temperature obtained from these experiments will be presented in this paper. (author)

  6. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ali Khan, Liaquat; Ahmad, Nasir E-mail: epg.piaas@dgcc.org.pk; Zafar, M.S.; Ahmad, Ayaz

    2000-07-01

    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.

  8. 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

  9. Development of the impression material for the replica of the reactor components in pool

    International Nuclear Information System (INIS)

    It is a very difficult and time-consuming work to remove and install the components because of its inherent characteristics and the physical interference with other components in the reactor core. We developed to easily inspect the wear marks or the deformation of the reactor components, for our purpose, a proper impression compound from the commercial material usually called vinylpolysiloxane which dentists are generally using. A proper mixing rate of the additional silica powder, the chemical catalyst and the commercial material was decided through various tests to ensure the good workability, appropriate hardening time as well as taking a good replication of the reactor components. To develop the compound satisfying applicable conditions in the reactor pool, we considered the tool handling time, the water temperature, deformations of the compound during tool handling, and radiation damages of the compound in the reactor core. We had finally developed the impression material for our purpose and successfully accomplished the inspection of the wear marks for a few fuel channels. Also, we have another plan to inspect the deformation of the spider pin of the fuel channel by using an impression material

  10. Twentieth anniversary of German nuclear reactor insurance pool

    International Nuclear Information System (INIS)

    When the peaceful utilization of nuclear energy was launched, insurance companies in the countries concerned faced a risk whose magnitude intially was unfathomable and for whose assessment they were not experienced enough, either with respect to the hazard probability or the potential extent of a damage. For this reason a pool, an association on a pro rata base, was organized on a national level by all insurance companies willing to offer coverage. In addition, to further spread the risk, worldwide mutual cooperation was agreed upon among the national pools. The 107 companies which presently make up the German pool cooperate with pools in twenty other countries. (orig.)

  11. Post-test simulations of BTF-107: an in-reactor loss-of-coolant test with flow blockage and rewet

    International Nuclear Information System (INIS)

    The Blowdown Test Facility (BTF) located in the NRU reactor at Chalk River Laboratories is the principal experimental tool for the Canadian in-reactor safety research program. This dedicated facility was designed for performing integrated 'all effects' tests on CANDU-type fuel to generate data for verifying and assessing Canadian safety analysis codes and models. This paper briefly describes the first BTF experiment, designated BTF-107, and presents the results of post-test thermalhydraulics and fuel behaviour simulations of this experiment. The thermalhydraulics simulations, performed using the CATHENA computer code, focus on analyzing the response of the BTF test section following blowdown, during dryout, and during the final rewet phase of the experiment. The fuel behaviour simulations, performed using the ELOCA Mk5 code, give estimates of the thermo-mechanical and fission-product release behaviour of the fuel during the course of the transient. The results of these simulations illustrate the capabilities of the CATHENA and ELOCA codes to model the processes involved in this severe high-temperature transient, and indicate possible areas for future improvement of these codes. (author). 7 refs., 2 tabs., 7 figs

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

    International Nuclear Information System (INIS)

    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

  13. 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.

  14. 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)

  15. Benchmarking the RELAP5/MOD2.5 r-Θ model of an SRS [Savannah River Site] reactor to the 1989 L-Reactor tests

    International Nuclear Information System (INIS)

    Benchmarking calculations utilizing RELAP5/MOD2.5 with a detailed multi-dimensional r-θ model of the SRS L-Reactor will be presented. This benchmarking effort has provided much insight into the two-component two-phase behavior of the reactor under isothermal conditions with large quantities of air ingested from the moderator tank to the external loops. Initial benchmarking results have illuminated several model weaknesses which will be discussed in conjunction with proposed modeling changes. The benchmarking work is being performed to provide a fully qualified RELAP5 model for use in computing the system response to a double ended large break LOCA. 5 refs., 14 figs

  16. Radiometric analysis of the spent fuel pool water and reactor coolant of ET-RR.1

    International Nuclear Information System (INIS)

    This work aims at analysis of radioactivity levels in the water of spent fuel pool and reactor core of the Egyptian 2MW research reactor (ET-RR.1 at Inshas). Gamma spectrometric and laser fluorimetric analysis have been used for carrying out this study. The fission product 137Cs and activation product 60Co are found with very high concentration in the spent fuel storage pool water. Thirteen isotopes; La-140, Cr-51, Ba-140, I-131, Cs-137, Ce-144, Nb-95, Ce-141, Zr-95, Ru-103, Cs-134, Nd-147 and Zn-65 are identified in the reactor core water. However no radiological hazard resulted because the fission products are contained within the shielded reactor pool. The radioactivity released into the reactor coolant water is mainly controlled by the diffusion mechanism. (orig.)

  17. Location and Roles of Deep Pools in Likangala River during 2012 Recession Period of Lake Chilwa Basin

    OpenAIRE

    Rodgers Makwinja; Mphatso Chapotera; Patrick Likongwe; John Banda; Asaf Chijere

    2014-01-01

    The ecological study focusing on Likangala River was conducted during the recent (2012) Lake Chilwa recession and aimed at identifying the important pools and the impact of indigenous ecological knowledge on the use and management of the aquatic biodiversity in the pools. An extensive georeferencing of the pools, field observations, and measurement of the pool depths was conducted to locate and map the deep pools along the river. Garmin Etrex Venture HC, GPS, and georeferencing were used to o...

  18. Closing Canada's ‘universal’ reactor

    Science.gov (United States)

    Asghar, M.; Rogge, R.

    2015-08-01

    In reply to a post on the physicsworld.com blog about the forthcoming closure of the National Research Universal reactor at Chalk River, Ontario, Canada (“Lament for ‘the reactor that can do everything’”, 16 June, http://ow.ly/On9VN).

  19. Thermohydraulic analysis of loss of forced flow accident in a pool type reactor

    International Nuclear Information System (INIS)

    This paper shows a calculation model for the fuel and pool water temperatures, and the internal building pressure of a 5 MW pool-type reactor, in the hypothetical event of the forced flow interruption. it is obtained the solution of the thermal energy and the momentum conservation equations in one dimension, which represent the heat conduction and natural convection in the coolant. The reactor building pressure increment due to the partial pool water evaporation is also calculated, using a homogenous model with thermal equilibrium of the phases (liquid water and steam) and the existing air. The heat loss to the building walls is also considered. (Author)

  20. 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

  1. 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)

  2. Criticality safety for the new spent fuel storage pool of Etrr-1 reactor

    International Nuclear Information System (INIS)

    Etrr-1 is the first egyptian research reactor, it is 36 years old. The decision is taken to equipped the reactor with a new spent fuel storage pool. The facility houses a storage pool besides a handling region. The safety regulations require that the pool when filled with full capacity of fuel assembly FA must be subcritical over a long period during normal and accidental conditions. The present work studies this issues at different pool temperatures. The key parameters affecting pool criticality as lattice pitch and U235 loads are investigated. WIMS and CITATION codes are used in the calculations. The results show good agreement when compared with design calculations which is performed by MCNP code

  3. 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)

  4. 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

  5. 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

  6. 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.)

  7. 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

  8. Sipping Test: Checking for Failure of Fuel Elements at the OPAL Reactor

    International Nuclear Information System (INIS)

    Sipping measurements were implemented at the Open Pool Australian Light water reactor (OPAL) to test for failure in reactor fuel elements. Fission product released by the fuel element into the pool water was measured using both High Purity Germanium (HPGe) detection via samples and a NaI(Tl) detection in-situ with the sipping device. Results from two fuel elements are presented

  9. 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

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

    International Nuclear Information System (INIS)

    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. Clinch River Breeder Reactor secondary control rod system

    International Nuclear Information System (INIS)

    The shutdown system for the Clinch River Breeder Reactor (CRBR) includes two independent systems--a primary and a secondary system. The Secondary Control Rod System (SCRS) is a new design which is being developed by General Electric to be independent from the primary system in order to improve overall shutdown reliability by eliminating potential common-mode failures. The paper describes the status of the SCRS design and fabrication and testing activities. Design verification testing on the component level is largely complete. These component tests are covered with emphasis on design impact results. A prototype unit has been manufactured and system level tests in sodium have been initiated

  12. Combined effects experiments with the condensation pool test facility

    International Nuclear Information System (INIS)

    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

  13. 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

  14. 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)

  15. 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)

  16. Preliminary results of the US pool-boiling coils from the IFSMTF full-array tests

    International Nuclear Information System (INIS)

    The Large Coil Task to develop superconducting magnets for fusion reactors, is now in the midst of full-array tests in the International Fusion Superconducting Magnet Test Facility at Oak Ridge National Laboratory. Included in the test array are two pool-boiling coils designed and fabricated by US manufacturers, General Dynamics/Convair Division and General Electric/Union Carbide Corporation. So far, both coils have been energized to full design currents in the single-coil tests, and the General Dynamics coil has reached the design point in the first Standard-I full-array test. Both coils performed well in the charging experiments. Extensive heating tests and the heavy instrumentation of these coils have, however, revealed some generic limitations of large pool-boiling superconducting coils. Details of these results and their analyses are reported

  17. Development of system design and seismic performance evaluation for reactor pool working platform of a research reactor

    International Nuclear Information System (INIS)

    Highlights: • Design of reactor pool working platform (RPWP) is newly proposed for an open-tank-in-pool type research reactor. • Main concept of RPWP is to minimize the pool top radiation level. • Framework for seismic performance evaluation of nuclear SSCs in a deterministic and a probabilistic manner is proposed. • Structural integrity, serviceability, and seismic margin of the RPWP are evaluated during and after seismic events. -- Abstract: The reactor pool working platform (RPWP) has been newly designed for an open-tank-in-pool type research reactor, and its seismic response, structural integrity, serviceability, and seismic margin have been evaluated during and after seismic events in this paper. The main important concept of the RPWP is to minimize the pool top radiation level by physically covering the reactor pool of the open-tank-in-pool type research reactor and suppressing the rise of flow induced by the primary cooling system. It is also to provide easy handling of the irradiated objects under the pool water by providing guide tubes and refueling cover to make the radioisotopes irradiated and protect the reactor structure assembly. For this concept, the new three dimensional design model of the RPWP is established for manufacturing, installation and operation, and the analytical model is developed to analyze the seismic performance. Since it is submerged under and influenced by water, the hydrodynamic effect is taken into account by using the hydrodynamic added mass method. To investigate the dynamic characteristics of the RPWP, a modal analysis of the developed analytical model is performed. To evaluate the structural integrity and serviceability of the RPWP, the response spectrum analysis and response time history analysis have been performed under the static load and the seismic load of a safe shutdown earthquake (SSE). Their stresses are analyzed for the structural integrity. The possibility of an impact between the RPWP and the most

  18. 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

  19. Detection of fission products release in the research reactor 'RA' spent fuel storage pool

    International Nuclear Information System (INIS)

    Spent fuel resulting from 25 years of operating the 6.5/10 MW thermal heavy water moderated and cooled research reactor RA at the VINCA Institute is presently all stored in the temporary spent fuel storage pool in the basement of the reactor building. In 1984, the reactor was shut down for refurbishment, which for a number of reasons has not yet been completed. Recent investigations show that independent of the future status of the research reactor, safe disposal of the so far irradiated fuel must be the subject of primary concern. The present status of the research reactor RA spent fuel storage pool at the VINCA Institute presents a serious safety problem. Action is therefore initiated in two directions. First, safety of the existing spent fuel storage should be improved. Second, transferring spent fuel into another, presumably dry storage space should be considered. By storing the previously irradiated fuel of the research reactor RA in a newly built storage space, sufficient free space will be provided in the existing spent fuel storage pool for the newly irradiated fuel when the reactor starts operation again. In the case that it would be decided to decommission the research reactor RA, the newly built storage space would provide safe disposal for the fuel irradiated so far

  20. Testing linkage disequilibrium from pooled DNA: a contingency table perspective.

    Science.gov (United States)

    Xu, Jinfeng; Yang, Yaning; Ying, Zhiliang; Ott, Jurg

    2008-12-10

    Pooling DNA samples of multiple individuals has been advocated as a method to reduce genotyping costs. Under such a scheme, only the allele counts at each locus, not the haplotype information, are observed. We develop a systematic way for handling such data by formulating the problem in terms of contingency tables, where pooled allele counts are expressed as the margins and the haplotype counts correspond to the unobserved cell counts. We show that the cell frequencies can be uniquely determined from the marginal frequencies under the usual Hardy-Weinberg equilibrium (HWE) assumption and that the maximum likelihood estimates of haplotype frequencies are consistent and asymptotically normal as the number of pools increases. The limiting covariance matrix is shown to be closely related to the extended hypergeometric distribution. Our results are used to derive Wald-type tests for linkage disequilibrium (LD) coefficient using pooled data. It is discovered that pooling is not efficient in testing weak LD despite its efficiency in estimating haplotype frequencies. We also show by simulations that the proposed LD tests are robust to slight deviation from HWE and to minor genotype error. Applications to two real angiotensinogen gene data sets are also provided. PMID:18712782

  1. The Fluid Dynamics Analysis of the RSG GAS Reactor's Pool with FLUENT 6

    International Nuclear Information System (INIS)

    The RSG-GAS reactor has been operating for eighteen years, and as long as the operation there were many changes on its characteristics. Therefore, some safety analysis must be recalculated and reviewed to ensuring the safety of reactor operation. Safety analysis is carried out by modeling the system and virtually simulation of the severe accident in the model. Accuracy of the analysis is strongly depending on the similarity of the model to actual system to be modeled. One of the data that required in the safety analyses is the fluid flow pattern of reactor pool where the core is placed inside. The data is very useful when the modeling in the comprehensive safety analyses of RSG-GAS is carried out. The fluid flow pattern analysis was tried unsuccessfully, since the modeling was inappropriate. In this research, computational fluid dynamic analysis of the reactor pool is conducted utilizing FLUENT 6. The software solves three dimensionally mass, momentum and energy conservation equations, and also considers the turbulence and the boundaries condition with many model provided on it. The analysis resulted in an appropriate model of the reactor's pool for FLUENT 6 and fluid flow pattern of the RSG-GAS reactor pool. The calculation was converged easily and the resulted flow pattern has strong correlation with the actual condition, therefore the results of the analysis are acceptable. And this successful analysis results have not been achieving previously in RSG-GAS. (author)

  2. 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

  3. 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)

  4. Distribution of 16N and 19O in the reactor pool water of the THOR facility

    International Nuclear Information System (INIS)

    Radioactive 16N and 19O in the Tsing Hua Open-Pool Reactor, produced from 16O(n,p)16N and 18O(n,γ)19O reactions, respectively, have been measured using a rapid sampling device and gamma-ray spectroscopic systems. The radioactivity of the 7-s half-life 16N and 27-s half-life 19O in the pool water are monitored in the power range from 1 W to 1 MW. The three-dimensional concentration of these radionuclides in the water coolant is also contour mapped down to the detection limit of 10 Bq/l. The spatial distribution of the short-lived radionuclides in the reactor pool, resulting from both the neutron flux distribution and heat transfer characteristics external to the core, is discussed for reactor operation at various power levels

  5. 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

  6. Basic CFD investigation of decay heat removal in a pool type research reactor

    International Nuclear Information System (INIS)

    Safety is one of the most important and desirable characteristic in a nuclear plant. Natural circulation cooling systems are noted for providing passive safety. These systems can be used as mechanism for removing the residual heat from the reactor, or even as the main cooling system for heated sections, such as the core. In this work, a computational fluid-dynamics (CFD) code is used to simulate the process of natural circulation in an open pool research reactor after its shutdown. The physical model studied is similar to the Open Pool Australian Light water reactor (OPAL), and contains the core, cooling pool, reflecting tank, circulation pipes and chimney. For best computing performance, the core region was modeled as a porous media, where the parameters were obtained from a separately detailed CFD analysis. (author)

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

    International Nuclear Information System (INIS)

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

  8. 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...

  9. 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...

  10. 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...

  11. 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...

  12. 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...

  13. 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...

  14. 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...

  15. 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...

  16. 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...

  17. 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...

  18. 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...

  19. 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...

  20. 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...

  1. 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...

  2. 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...

  3. 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...

  4. 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...

  5. 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...

  6. 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...

  7. 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...

  8. 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...

  9. 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...

  10. 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...

  11. 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...

  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. 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...

  14. 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...

  15. 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...

  16. 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...

  17. The IBR-2 test reactor

    International Nuclear Information System (INIS)

    Major design criteria, specifications and potential fields of application of the IBR-2 pulsed test reactor (now under construction in Dubna, USSR) are described. The pulsed power bursts will be due to fast periodic reactivity changes by a rotating reflector. The frequency of approximately 100 μs pulsed may be 5, 12.5 or 50 Hz. The IBR-2 reactor will be mostly profitable for slow neutron experiments when investigating solids, nuclei or neutrons themselves using spectroscopic methods. Due to the high peak flux of thermal neutrons (1016-1017 n/cm2xs) the reactor will be superior (for the sort of experiments) to the currently operating SM-2 and HFR high flux steady-state test reactors for many times

  18. Cobalt-60 production in CANDU power reactors

    International Nuclear Information System (INIS)

    MDS Nordion has been supplying cobalt-60 sources to industry for industrial and medical purposes since 1946. These cobalt-60 sources are used in many market and product segments. The major application is in the health care industry where irradiators are used to sterilize single use medical products. These irradiators are designed and built by MDS Nordion and are used by manufacturers of surgical kits, gloves, gowns, drapes and other medical products. The irradiator is a large shielded room with a storage pool for the cobalt-60 sources. The medical products are circulated through the shielded room and exposed to the cobalt-60 sources. This treatment sterilizes the medical products which can then be shipped to hospitals for immediate use. Other applications for this irradiation technology include sanitisation of cosmetics, microbial reduction of pharmaceutical raw materials and food irradiation. The cobalt-60 sources are manufactured by MDS Nordion in their Cobalt Operations Facility in Kanata. More than 75,000 cobalt-60 sources for use in irradiators have been manufactured by MDS Nordion. The cobalt-60 sources are double encapsulated in stainless steel capsules, seal welded and helium leak tested. Each source may contain up to 14,000 curies. These sources are shipped to over 170 industrial irradiators around the world. This paper will focus on the MDS Nordion proprietary technology used to produce the cobalt-60 isotope in CANDU reactors. Almost 55 years ago MDS Nordion and Atomic Energy of Canada developed the process for manufacturing cobalt-60 at the Chalk River Labs, in Ontario, Canada. A cobalt-59 target was introduced into a research reactor where the cobalt-59 atom absorbed one neutron to become cobalt-60. Once the cobalt-60 material was removed from the research reactor it was encapsulated in stainless steel and seal welded using a Tungsten Inert Gas weld. The first cobalt-60 sources manufactured using material from the Chalk River Labs were used in cancer

  19. 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.)

  20. Effects of nuclear island connected buildings on seismic behaviour of reactor internals in a pool type fast breeder reactor

    International Nuclear Information System (INIS)

    The seismic analysis of reactor assembly housing the primary circuit of a typical 500 MWe capacity pool type fast breeder reactor (PFBR) is reported. The reactor assembly is supported on the reactor vault within the nuclear island connected buildings (NICB). The seismic responses, viz. critical displacements, sloshing heights, stresses and strain energy values in the vessels are determined for the reactor assembly by detailed finite element analysis including the fluid-structure interaction and sloshing effects. Analysis is carried out to quantify the effects of inter-connection of the reactor vault with the adjacent buildings under the assumptions that the reactor vault along with reactor assembly is: (1) an isolated structural system from the adjacent buildings within reactor containment building (RCB) and (2) connected with the adjacent civil structures through floor slabs. Analysis indicates that, by inter-connecting the vault with the NICB, there are overall increases of all the governing parameters which decide the seismic design criteria. The significant effects are increases of: (1) radial and axial displacements of core top and absorber rods and vertical accelerations of core subassemblies which are of concern to reactor safety, (2) primary membrane stress intensities for the inner vessel and (3) strain energies developed at the critical portions which can enhance the buckling risks of main vessel, inner vessel and thermal baffles. Hence, it is preferable to isolate the reactor vault, directly constructing from the base raft without inter-connecting it with the NICB, from the seismic loading considerations

  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. 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

  3. Characterization of radioactive contaminants and water treatment trials for the Taiwan Research Reactor's spent fuel pool

    International Nuclear Information System (INIS)

    Highlights: ► Deal with a practical radioactive contamination in Taiwan Research Reactor spent fuel pool water. ► Identify the properties of radioactive contaminants and performance test for water treatment materials. ► The radioactive solids were primary attributed by ruptured spent fuels, spent resins, and metal debris. ► The radioactive ions were major composed by uranium and fission products. ► Diatomite-based ceramic depth filter can simultaneously removal radioactive solids and ions. - Abstract: There were approximately 926 m3 of water contaminated by fission products and actinides in the Taiwan Research Reactor's spent fuel pool (TRR SFP). The solid and ionic contaminants were thoroughly characterized using radiochemical analyses, scanning electron microscopy equipped with an energy dispersive spectrometer (SEM-EDS), and inductively coupled plasma optical emission spectrometry (ICP-OES) in this study. The sludge was made up of agglomerates contaminated by spent fuel particles. Suspended solids from spent ion-exchange resins interfered with the clarity of the water. In addition, the ionic radionuclides such as 137Cs, 90Sr, U, and α-emitters, present in the water were measured. Various filters and cation-exchange resins were employed for water treatment trials, and the results indicated that the solid and ionic contaminants could be effectively removed through the use of <0.9 μm filters and cation exchange resins, respectively. Interestingly, the removal of U was obviously efficient by cation exchange resin, and the ceramic depth filter composed of diatomite exhibited the properties of both filtration and adsorption. It was found that the ceramic depth filter could adsorb β-emitters, α-emitters, and uranium ions. The diatomite-based ceramic depth filter was able to simultaneously eliminate particles and adsorb ionic radionuclides from water.

  4. 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)

  5. Development of thermal-hydraulic system analysis code SSC-K for pool-type liquid metal reactor

    International Nuclear Information System (INIS)

    The Supper System Code of KAERI (SSC-K) is a best-estimate system code for analyzing an variety of off-normal or accident of a pool type design. It is developed at KAERI on the basis of SSC-L developed at BNL to analyze pool-type LMR transients. Because of inherent difference between th pool and loop design, the major modefications of SSC-L is required for the safety analysis of KALIMER. The major difference between KALIMER and general loop type LMRs exists in the primary heat transport system. In KALIMER, all of the essential components consisted of the primary heat transport system are located within the reactor vessel. This is contrast to the loop type LMRs, in which all the primary components are connected via piping to form loops attached externally to the reactor vessel. KALIMER has only one cover gas space. This eliminates the need for separate cover gas systems over liquid level in pump tanks and upper plenum. Since the sodium in hot pool is separated from cold pool by insulated barrier in KALIMER, The liquid level in hot pool is different from that in the cold pool mainly due to hydraulic losses and pump suction heads occuring during flow through the circulation pathes. In some accident conditions the liquid in the hot pool is flooded into cold pool and forms the natural circulation flow path. During the loss of heat sink transients, this will provided as a major heat rejection mechanism with the passive decay heat removal system. Since the pipes in the primary system exist only between pump discharge and core inlet plenum and are submerged in cold pool, a pipe rupture accident becomes less severe due to a constant back pressure exerted against the coolant flow from break. The intermediate and steam generator systems of both are generally identical. To adapt SSC-K to KALIMER design, the major modification of SSC-L has been made for the safety analysis of KALIMER. Test runs have been performed for the qualitative verification of the developed models. The

  6. 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

  7. Summary record of the twenty-third meeting, (technical sessions), Chalk River, Canada, 27 Sep - 1 Oct 1982

    International Nuclear Information System (INIS)

    The technical sessions deal with advances in nuclear data measurements and newer facilities (with special attention paid to data for fission and fusion reactors); advances in nuclear data evaluations (regional activities and joint evaluated file); activities of nuclear data centres; report on recent NEA meetings; subcommittee reports

  8. Chalk as a reservoir

    DEFF Research Database (Denmark)

    Fabricius, Ida Lykke

    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, or be......, and the best reservoir properties are typically found in mudstone intervals. Chalk mudstones vary a lot though. The best mudstones are purely calcitic, well sorted and may have been redeposited by traction currents. Other mudstones are rich in very fine grained silica, which takes up pore space and...... thus 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...

  9. 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.

  10. The NRU blowdown test facility commissioning program

    International Nuclear Information System (INIS)

    A major experimental program has been established at the Chalk River Nuclear Laboratories (CRL) that will provide essential data on the thermal and mechanical behaviour of nuclear fuel under abnormal reactor operating conditions and on the transient release, transport and deposition of fission product activity from severely degraded fuel. A number of severe fuel damage (SFD) experiments will be conducted within the Blowdown Test Facility (BTF) at CRL. A series of experiments are being conducted to commission this new facility prior to the SFD program. This paper describes the features and the commissioning program for the BTF. A development and testing program is described for critical components used on the reactor test section. In-reactor commissioning with a fuel assembly simulator commenced in 1989 June and preliminary results are given. The paper also outlines plans for future all-effects, in-reactor tests of CANDU-designed fuel. (author). 11 refs., 3 tabs., 7 figs

  11. U–Pb, Rb–Sr, and U-series isotope geochemistry of rocks and fracture minerals from the Chalk River Laboratories site, Grenville Province, Ontario, Canada

    International Nuclear Information System (INIS)

    Highlights: • AECL evaluates Chalk River Laboratories site as potential nuclear waste repository. • Isotope-geochemical data for rocks and fracture minerals at CRL site are reported. • Zircons from gneiss and granite yielded U–Pb ages of 1472 ± 14 and 1045 ± 6 Ma. • WR Rb–Sr and Pb–Pb systems do not show substantial large-scale isotopic mobility. • U-series and REE data do not support oxidizing conditions at depth in the past 1 Ma. - Abstract: As part of the Geologic Waste Management Facility feasibility study, Atomic Energy of Canada Ltd. (AECL) is evaluating the suitability of the Chalk River Laboratories (CRL) site in Ontario, situated in crystalline rock of the southwestern Grenville Province, for the possible development of an underground repository for low- and intermediate-level nuclear waste. This paper presents petrographic and trace element analyses, U–Pb zircon dating results, and Rb–Sr, U–Pb and U-series isotopic analyses of gneissic drill core samples from the deep CRG-series characterization boreholes at the CRL site. The main rock types intersected in the boreholes include hornblende–biotite (±pyroxene) gneisses of granitic to granodioritic composition, leucocratic granitic gneisses with sparse mafic minerals, and garnet-bearing gneisses with variable amounts of biotite and/or hornblende. The trace element data for whole-rock samples plot in the fields of within-plate, syn-collision, and volcanic arc-type granites in discrimination diagrams used for the tectonic interpretation of granitic rocks. Zircons separated from biotite gneiss and metagranite samples yielded SHRIMP-RG U–Pb ages of 1472 ± 14 (2σ) and 1045 ± 6 Ma, respectively, in very good agreement with widespread Early Mesoproterozoic plutonic ages and Ottawan orogeny ages in the Central Gneiss Belt. The Rb–Sr, U–Pb, and Pb–Pb whole-rock errorchron apparent ages of most of the CRL gneiss samples are consistent with zircon U–Pb age and do not indicate

  12. 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.

  13. 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.

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

    International Nuclear Information System (INIS)

    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∼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

  15. 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

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

    International Nuclear Information System (INIS)

    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 previously unknown locations of waste

  18. 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)

  19. 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)

  20. 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

  1. 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

  2. Benefit of chromium in reducing the rates of flow accelerated corrosion of carbon steel outlet feeders in CANDU reactors

    International Nuclear Information System (INIS)

    In the mid 1990's, wall thinning of outlet feeders due to flow accelerated corrosion (FAC) was recognized as an active mechanism in the outlet feeders of CANDU reactors. To address wall thinning of outlet feeders in new reactor construction and refurbishment projects, AECL introduced a minimum Cr concentration in its specification for the SA-106 carbon steel feeder pipe. The effectiveness of Cr in reducing FAC was subsequently demonstrated in in-reactor and out-reactor loops at AECL's Chalk River Laboratories. More recently, wall-thinning rates have been determined from wall thickness data collected from outlet feeders, containing a specified minimum Cr concentration, installed in the Point Lepreau Generating Station in 2001. This paper presents the FAC rates determined from in-service outlet feeders and compares the rates with data from previous in-reactor and out-reactor test loops, highlighting the consistency observed in results from the three sources. (author)

  3. PITR: Princeton Ignition Test Reactor

    International Nuclear Information System (INIS)

    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

  4. 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.

  5. Heavy Water Components Test Reactor Decommissioning

    International Nuclear Information System (INIS)

    The Heavy Water Components Test Reactor (HWCTR) Decommissioning Project was initiated in 2009 as a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Removal Action with funding from the American Recovery and Reinvestment Act (ARRA). This paper summarizes the history prior to 2009, the major D and D activities, and final end state of the facility at completion of decommissioning in June 2011. The HWCTR facility was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In the early 1990s, DOE began planning to decommission HWCTR. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. In 2009 the $1.6 billion allocation from the ARRA to SRS for site footprint reduction at SRS reopened the doors to HWCTR - this time for final decommissioning. Alternative studies concluded that the most environmentally safe, cost effective option for final decommissioning was to remove the reactor vessel, both steam generators, and all equipment above grade including the dome. The transfer coffin, originally above grade, was to be placed in the cavity vacated by the reactor vessel and the remaining below grade spaces would be grouted. Once all above equipment

  6. A calculational model for the NRU reactor

    International Nuclear Information System (INIS)

    A new computer model to calculate neutronic properties of the NRU research reactor is being implemented at the Chalk River Nuclear Laboratories (CRNL). The model is founded on numerous theoretical studies, on analysis of NRU support experiments done in a zero power reactor, and on comparison with measurements in the NRU reactor. This paper examines the elements of the new calculational model, concentrating on the unique features of NRU and their influences on neutron behaviour

  7. A calculational model for the NRU reactor

    International Nuclear Information System (INIS)

    A new computer model to calculate neutronic properties of the NRU research reactor is being implemented at the Chalk River Nuclear Laboratories (CRNL). The model is founded on numerous theoretical studies, on analysis of NRU support experiments done in a zero power reactor, and on comparison with measurement in the NRU reactor. This paper examines the elements of the new calculational model, concentrating on the unique features of NRU and their influences on neutron behaviour

  8. Compaction of microfossil and clay-rich chalk sediments

    DEFF Research Database (Denmark)

    Fabricius, Ida Lykke

    2001-01-01

    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 ar...

  9. 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

  10. Pools and rapids as spawning and nursery areas for fish in a river stretch without floodplains

    Directory of Open Access Journals (Sweden)

    Sunshine de Ávila-Simas

    2014-09-01

    Full Text Available This study aimed to evaluate the importance of two environments situated in the main channel of the Peixe River (a tributary of the upper Uruguay River on fish reproduction and initial growth. Ichthyoplankton, macrozooplankton, and zoobenthos collections were taken on a monthly basis from October 2011 to March 2012, sampling a rapids and a pool environment. The instrument used for the capture of the ichthyoplankton in both environments was a light trap. In total, 795 eggs and 274 larvae were captured. The species that presented higher abundance and occurrence frequency out of the total captured in both environments were Leporinus obtusidens, Bryconamericus iheringii, and Bryconamericus stramineus. The evaluation of the feeding activity reveals a major repletion degree of the larvae in more advanced stages in the pool. The pool environment presented a higher abundance of larvae in more advanced development stages. We conclude that the channel of the Peixe River is important for the reproduction and initial growth of fish and that each river environment seems to fulfill a different role in the life cycle of the ichthyoplankton community.

  11. A model for the analysis of loss of decay heat removal accident in MTR pool type research reactors

    International Nuclear Information System (INIS)

    During a loss of coolant accident leading to total emptying of the reactor pool, the decay heat could be removed through air natural convection. However, under partial pool emptying the core is partially submerged and the coolant circulation inside the fuel element could no more be possible. In such conditions, a core overheat take place, and the heat is essentially diffused from the core to its periphery by combined thermal radiation and conduction. In order to predict fuel element temperature evolution under such conditions a mathematical model is performed. The model is based on a three dimensional geometry and takes into account a variety of core configurations including fuel elements (standard and control), reflector elements and grid plates. The homogeneous flow model is used and the time and space dependent non-linear partial differential fluid conservation equations are solved using a semi-implicit finite difference method. Preliminary tests of the developed model were made by considering a series of hypothetical accidents. In the current framework a loss of decay heat removal accidents in the IAEA benchmark open pool MTR-type research reactor is considered. It is shown that in the case of a low core immersion height no water boiling is observed and the fuel surface temperature rise remains below the melting point of the aluminium cladding. (author)

  12. Testing the reactor charging machine

    International Nuclear Information System (INIS)

    One of the main objective of the R - D technological engineering program devoted to the Fuel Handling System is domestic production of equipment and technology for testing the ends of the reactor charging machine (MID) destined to Cernavoda NPP, beginning with Unit 2. To achieve the objective based on an own design, a bench-scale testing stand of MIDs which can simulate the pressure, flow-rate, and temperature conditions proper to fuel channels in operating CANDU 600 reactors. The main components of this testing facility are: - fuel channels, cold also test sections, allowing the coupling of MID end upwardly and downwardly, corresponding to the direction of the water flow through the channel; - technological installation feeding with light water the testing sections of the facility in thermohydraulic conditions, similar to those in the reactor, allowing the cold and hot testings, respectively, of the MID end; - cold testing installation, water supply and oil control panel, feeding the hydraulic drives of the MID's end during the testings; - fixed bridge and mobile carrier for MID's end positioning against testing sections; - installation for functional testing of MID thrusters, before pre-admission and reception tests; - dedicated tools and devices; - raising and transport mechanical devices for handling and positioning the MID's end upon the carrier; - automation panel for controlling the stand equipment and MID's end; - process computer for conducting on-line tests. MID's end testing implies mainly the following operations: - regulation, calibration and functional testing of the MID thrusters carried out independently on a specialised stand; - regulation and calibration of MID's end sub-assemblages; - carrying out the cold and hot pre-admission tests consisting in automatic performing, without operator intervention, of 12 fuel changes, two of which being successive; - performing the cold and hot reception tests, consisting in automatic accomplishment of 4

  13. Aquifer properties of the Chalk of England

    OpenAIRE

    MacDonald, Alan M.; Allen, David J

    2001-01-01

    Aquifer properties data from 2100 pumping tests carried out in the Chalk aquifer have been collated as part of a joint British Geological Survey/Environment Agency project. The dataset is highly biased: most pumping tests have been undertaken in valley areas where the yield of the Chalk is highest. Transmissivity values from measured sites give the appearance of log-normality, but are not truly log-normal. The median of available data is 540 m2/d and the 25th and 75th percentiles 190 m2/d and...

  14. 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)

  15. 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

  16. Surface pool dynamics and climate in boreal peatland of the La Grande River watershed

    International Nuclear Information System (INIS)

    A study was conducted to investigate the influence of climate on the morphological changes of boreal peatlands of the La Grande (LG) River watershed. In particular, changes in 100 pools from 3 boreal peatlands of different trophic status were examined in relation to their regional meteorological data. Aerial photographs from 1977 and 1999 were used to measure area, perimeter, length and width axes of 35 pools in each of the 3 peatlands from LG1, LG2 and LG3. Results show changes in size and shape of pools during the studied periods. The total area covered by pools decreased between 1977 and 1999. At LG1, a decrease of 246 m2 was measured between 1986 and 1990. At LG2, the total area covered by pools decreased by 3320 m2 between 1984 and 1999, and at LG3 the area decreased by 2634 m2 between 1979 and 1988. The changes appear to be related to seasonal or inter-annual variations of precipitation and evapotranspiration regime. Spatial variability response of pools to changes in hydrological condition indicates their different level of sensitivity

  17. 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.

  18. 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

  19. 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)

  20. 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)

  1. Simulation of the Gamma Dose Rate in Loss of Pool Water Accident of the Second Egyptian Research Reactor ETRR-2

    International Nuclear Information System (INIS)

    The Second Egyptian Research Reactor ETRR-2, is a pool type reactor, a sudden loss of pool water resulting of leaving the core region un-covered. The reactor core is surrounded by chimney chambers whose water is isolated from pool water. This accident would lead to significant external dose. A model is developed and is used to calculate the dose rates for key access and traffic plans from indirect line of sight of the core have a maximum dose rate. The model developed uses the discrete ordinate method as implemented in the code DOT 3.5

  2. 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)

  3. 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)

  4. 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)

  5. Conceptual design of reactor TRIGA PUSPATI (RTP) spent fuel pool cooling system

    International Nuclear Information System (INIS)

    After undergo about 30 years of safe operation, Reactor TRIGA PUSPATI (RTP) was planned to be upgraded to ensure continuous operation at optimum safety condition. In the meantime, upgrading is essential to get higher flux to diversify the reactor utilization. Spent fuel pool is needed for temporary storage of the irradiated fuel before sending it back to original country for reprocessing, reuse after the upgrading accomplished or final disposal. The irradiated fuel elements need to be secure physically with continuous cooling to ensure the safety of the fuels itself. The decay heat probably still exist even though the fuel elements not in the reactor core. Therefore, appropriate cooling is required to remove the heat produced by decay of the fission product in the irradiated fuel element. The design of spent fuel pool cooling system (SFPCS) was come to mind in order to provide the sufficient cooling to the irradiated fuel elements and also as a shielding. The spent fuel pool cooling system generally equipped with pumps, heat exchanger, water storage tank, valve and piping. The design of the system is based on criteria of the primary cooling system. This paper provides the conceptual design of the spent fuel cooling system. (author)

  6. 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.

  7. 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.

  8. MAPLE-X10 reactor digital control system

    International Nuclear Information System (INIS)

    The MAPLE-X10 reactor, currently under construction at the Chalk River Laboratories of Atomic Energy of Canada Limited, is a 10 MWt, pool-type, light-water reactor. It will be used for radioisotope production and silicon neutron transmutation doping. The reactor is controlled by a Digital Control System (DCS) and protected against abnormal process events by two independent safety systems. The DCS is an integrated control system used to regulate the reactor power and process systems. The safety philosophy for the control system is to minimize unsafe events arising from system failures and operational errors. this is achieved through redundancy, fail-safe design, automatic fault detection, and the selection of highly reliable components. The DCS provides both computer-controlled reactor regulation from the shutdown state to full power and automated reactor shutdown if safe limits are exceeded or critical sensors malfunction. The use of commercially available hardware with enhanced quality assurance makes the system cost effective while providing a high degree of reliability

  9. 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.

  10. 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

  11. Dose-effects relationships in wild populations of the aquatic snail Campeloma decisum at Chalk River Laboratories

    International Nuclear Information System (INIS)

    In the last decade regulatory bodies worldwide have implemented standards to protect populations of non-human biota (NHB) from the consequences of radiation exposure. This is a departure from previous regulatory frameworks, which were concerned only with protecting man. The implementation of these new standards initiated an ongoing discussion concerning appropriate dose-rate limits for NHB. For the most part, the data utilized for estimating appropriately protective dose-rate limits has come from data collected via the irradiation of NHB in a laboratory setting. While some dose-effects studies have been performed under field conditions, such experiments represent a minority of the available data. This deficit in the literature has resulted in challenges to the established paradigm, with researchers reporting increased radiosensitivity in NHB under field conditions. However, many such studies overlook critical components of dose-effects analysis: lacking either robust ecological technique or dosimetric rigor. The study cited herein provides rigorous analysis of factors affecting populations of aquatic snails and is intended as a framework for identifying those factors statistically indicative of snail population. These benchmarks (e.g., number of snails, mass of individuals) were employed as proxies for snail population health, and how it was impacted by over two dozen environmental variables. Dose-rates were calculated via a novel voxel model, developed for this study to estimate internal dose rates for the species of interest. A linear regression model was employed to tease out the relationship between individual snails, their environment, and radiation dose rate. There was no evidence that snail population health was influenced by radiation exposure (p=0.70) at the observed dose rates. Of the environmental variables tested, water concentration of Ca was well correlated with snail mass size (p<0.001), while water concentration of P was well correlated with the

  12. 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.

  13. 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.

  14. 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)

  15. 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.)

  16. 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...

  17. 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

  18. 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.

  19. 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)

  20. Hazards in Choosing Between Pooled and Separate- Variances t Tests

    Directory of Open Access Journals (Sweden)

    Bruno D. Zumbo

    2009-01-01

    Full Text Available If the variances of two treatment groups are heterogeneous and, at the same time, sample sizes are unequal, the Type I error probabilities of the pooledvariances Student t test are modified extensively. It is known that the separate-variances tests introduced by Welch and others overcome this problem in many cases and restore the probability to the nominal significance level. In practice, however, it is not always apparent from sample data whether or not the homogeneity assumption is valid at the population level, and this uncertainty complicates the choice of an appropriate significance test. The present study quantifies the extent to which correct and incorrect decisions occur under various conditions. Furthermore, in using statistical packages, such as SPSS, in which both pooled-variances and separate-variances t tests are available, there is a temptation to perform both versions and to reject H0 if either of the two test statistics exceeds its critical value. The present simulations reveal that this procedure leads to incorrect statistical decisions with high probability.

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

    International Nuclear Information System (INIS)

    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)

  2. 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)

  3. 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)

  4. 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)

  5. 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.

  6. 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

  7. Ground test facility for nuclear testing of space reactor subsystems

    International Nuclear Information System (INIS)

    Two major reactor facilities at the INEL have been identified as easily adaptable for supporting the nuclear testing of the SP-100 reactor subsystem. They are the Engineering Test Reactor (ETR) and the Loss of Fluid Test Reactor (LOFT). In addition, there are machine shops, analytical laboratories, hot cells, and the supporting services (fire protection, safety, security, medical, waste management, etc.) necessary to conducting a nuclear test program. This paper presents the conceptual approach for modifying these reactor facilities for the ground engineering test facility for the SP-100 nuclear subsystem. 4 figs

  8. 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)

  9. 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)

  10. modeling of a total loss pool water accident in mtr reactor

    International Nuclear Information System (INIS)

    in this study , it is intended to analyze early phases of a protected loss of coolant accident (LOCA)for MTR reactor. and to show the applicability of the presented model to the other similar types of research reactors. the transient situation since the time when coolant is beginning to be lost throughout one or more of the main coolant pipes which were supposed to be broken guillotine-like to the time when the core is totally uncovered is investigated. the modeling of the problem was separated into two phases; in the first phase when the water level of the pool is being decreased in a pre-estimated time -dependent way calculated by using modified Bernoulli equation, the conservation equation are solved by using shooting method. the later phase, when water level reaches the top level of fuel plates and begins to decrease until bottom of the core, and the fuel plates are being cooled by air.

  11. Research and manufacture of Rossi-α measurement device for pool thermal reactor based on virtual instrument technology

    International Nuclear Information System (INIS)

    This paper designs a set of Rossi-α measurement device for pool thermal reactor based on virtual instrument technology. This device makes PXI-6602 high-speed synchronization counter as the hardware measurement platform. By using Labview 8.5 program, the application program of data acquirement and data process analysis of the measurement device have been designed. In addition, By using Fluke-282 arbitrary waveform generator, one channel square signal which frequency and voltage are respectively defined as 200 Hz and 5 V have been produced. By using application analysis program, the pulse counter simulation test has been completed for this channel square signal. According to the test results, the measurement device reaches the measurement use requirements. (authors)

  12. Startup operational tests of fast reactors

    International Nuclear Information System (INIS)

    This paper is mainly concerned with the experiences of the two main phases of startup operational tests of fast reactors: (1) The general tests and Sodium filling before core loading. (2) The core loading,approach to criticality and power build up operational tests, taking for example a large and middle demonstrating integrated-type fast reactor. (author)

  13. Recent reactor testing and experience with gamma thermometers

    International Nuclear Information System (INIS)

    Recent experience with gamma thermometers for light water reactors has primarily been in the Framatome reactors operated by Electricite de France. Other recent testing has taken place at Oak Ridge National Laboratory and the Otto Hahn ship reactor. Earlier experience with gamma thermometers was in heavy water reactors at Savannah River and Halden. This paper presents recent data from the light water reactor (LWR) programs. The principles of design and operation of the Radcal gamma thermometer were presented in ''Gamma Thermometer Developments for Light Water Reactors'', Leyse and Smith1. Observations from LWRs confirm the earlier experience from heavy water reactors that the gamma thermometer units give signals which are proportional to the power of surrounding fuel rods and virtually independent of exposure, surrounding poison and other conditions which affect signals of neutron sensitive devices. After 200 sensor-years in EdF reactors, there has been no change in the sensitivity of the devices. Nonetheless, the Radcal units can be recalibrated in-reactor by the introduction of electrical heating via a heater cable imbedded in the device. Algorithms and signal processing software have been developed to interpret and display the gamma thermometer signals. The results of this processing are illustrated here

  14. 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.

  15. Sipping test of fuel assemblies in LVR-15 reactor

    International Nuclear Information System (INIS)

    The LVR-15 reactor is a light water research type which is situated at NRI in Rez near Prague. The poster describes the procedure and methodology used for sipping test of the fuel assemblies. These tests are designed to evaluate the leakage of fuel and fission products from the tested fuel assembly. From 1995 to 2003 there have been performed about 200 tests. Examples of results of sipping water activity measurements are presented. The values of activities of 137Cs and 134Cs are used for decision if the fuel assembly can be used in reactor core, transported to storage pool or if it is necessary to put the fuel assembly into the special protective can. The used limits of activities are discussed. (author)

  16. 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.

  17. 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)

  18. Current activities of chemical applications of Tsing Hua Open-pool Reactor

    International Nuclear Information System (INIS)

    Tsing Hua Open-pool Reactor (THOR) is a swiming pool type research reactor operated and maintained by the University. The functions of THOR are two folds, i.e. (1) teaching and training of nuclear scientists and engineers, and (2) promotion of peaceful uses of atomic energy in Taiwan. Thus, besides the educational program THOR has been offering neutron irradiation service on regular schedule (30 hrs. per week full power operation at 1 MW with neutron flux of ca. 2 x 1012 n cm-2 sec.-1 since 1963). Among many other projects chemical application of THOR has been one of the important program implimented in early days. Following two projects have been set up with rather limited budget: (1) Production and supply of short-lived radioisotopes (both general purpose and medical use) for domestic use. (2) Application of neutron activation analysis in the fields of environmental, geological, biological and material science. It turned out that quite fruitful results have been obtained. It is the purpose of this paper to describe in detail some of the characteristic aspects on the utilization of THOR in the field of chemical application. (author)

  19. A Versatile Cobalt-60 Irradiation Facility within a Swimming-Pool Research Reactor

    International Nuclear Information System (INIS)

    The IRR-1 enriched-fuel swimming-pool-type reactor incorporates a concrete-shielded gamma cell for using the radiation from spent fuel elements. A versatile 60Co irradiation facility was added at relatively low cost. The 30 000 Ci 60Co source runs on a carriage at the bottom of the reactor pool. The source plaque completely covers the aluminium window between the pool and the irradiation cell. This geometry allows for a ''source-to-target'' overlap; therefore, dose-rate homogeneity within ±20% is attained inside two commercial cases (43 x 31 x 31 cm) or two commercial sacks (80 x 45 x 45 cm) and within ±10% in two flat boxes (40 x 20 x 5 cm) irradiated simultaneously. A set of steel and aluminium screens attached to two rotating turn-tables permits irradiation of commercial cases at any desired dose-rate smaller than 100 000 R/h without the need for turning over at half-time. Two special underwater canisters allow long-term irradiation of flat specimens at dose-rates of less than 600 000 R/h, while the source is used for normal short-term irradiations in the gamma cell. Safety is ensured by a visible and audible indicator and alarm system and by an elaborate interlock system. A system of ionizing gauges and recorders permits measurement of dose-rates over the range 0. 001 to 1 000 000 R/h. Isodose curves for the irradiation chamber have been determined. The cell is soon to be modified to include a refrigeration plant and a timing system for automatic control of source movement. The disadvantages of low source utilization inherent in required source-to-target overlap and of one sided utilization of the radiation are more than compensated for by the possibility of pilot-scale irradiation of commercial cases and by the greater versatility and low cost. This installation is therefore recommended for all similar swimming-pool reactors. It is especially valuable for countries desiring to embark on a food irradiation programme at minimum cost but with maximum

  20. Savannah River Site production reactor safety analysis report. Vol. VIII

    International Nuclear Information System (INIS)

    The Savannah River Site (SRS) production reactors are unique in their methods of charging and discharging fuel assemblies, target assemblies, and other components. All components are charged and discharged in the air using remotely operated precision cranes. The following sections describe the systems used to store, charge, discharge, and disassemble reactor components and assemblies, and the redundant systems designed to ensure the reliability of crane cooling and control systems

  1. Measurement of the axial distribution of thermal neutron flux beside NRU loop fuel test sites

    International Nuclear Information System (INIS)

    At Chalk River Laboratories, fuel bundle tests for the CANDU power reactor fuel development program are performed in the fuel test sites of the NRU reactor loops. At present, calculated axial neutron flux profiles from neutronics modeling of the NRU reactor are used to distribute the total measured powers of the loop fuel test sites to determine the relative fuel bundle powers and burnups of the test bundles. In order to provide data for validating the calculated fluxes, measurements of the axial neutron flux distributions adjacent to the loop fuel test sites were also performed. This paper describes how the axial thermal neutron flux distributions were measured using in-core flux detectors and presents the results of comparisons between the measured fluxes and the calculated fluxes predicted by the neutronics simulation code. (author)

  2. Full scale steady state component tests of the SWR 1000 Fuel Pool Cooler at the INKA test facility

    Energy Technology Data Exchange (ETDEWEB)

    Maisberger, Fabian; Leyer, Stephan; Schaub, Bernd; Brettschuh, Werner; Wagner, Thomas; Doll, Mathias; Wich, Michael; Schaefer, Heinrich [AREVA NP, Offenbach (Germany); Unger, Jochem [TU Darmstadt (Germany)

    2009-07-01

    The SWR 1000 is a medium-capacity boiling water reactor. It combines proven design active safety systems with innovative passive safety systems. The passive systems utilizes basic physical laws, such as gravity or natural convection, enabling them to function without electrical power supply or actuation by powered instrumentation and control (I and C) systems. They are designed to bring the plant in a secure and stable state without the help of any active system. Furthermore the passive safety features partially replace the active systems leading to a significant cost reduction and provide a reliable, safe and economically competitive alternative to standard plant design /1/. For further simplification of the plant design and additional cost reduction, the fuel pool cooling system has been modified in comparison to the currently running German BWR plants. This new system was tested in the Pressure Suppression Pool Vessel (PSPV) of the INKA test facility (Integral Teststand Karlstein) in Germany, which was originally build for the full scale testing of the key elements of the SWR 1000 passive safety concept /2/. The PSPV of INKA was chosen because it provides enough space for the cooler and its attached chimney (total height 11.5m). In this work the setup and the execution of the tests will be described. A characteristic diagram of the heat transfer capacity of the component as a function of cooling water temperature and fuel pool water temperature obtained form these experiments will be presented. In parallel CFD calculations, simulating the tests will be made. The results of these calculations and the comparison between the experimental and calculated results will be presented elsewhere and will serve furthermore to validate the CFD-code. (orig.)

  3. Full scale steady state component tests of the SWR 1000 Fuel Pool Cooler at the INKA test facility

    International Nuclear Information System (INIS)

    The SWR 1000 is a medium-capacity boiling water reactor. It combines proven design active safety systems with innovative passive safety systems. The passive systems utilizes basic physical laws, such as gravity or natural convection, enabling them to function without electrical power supply or actuation by powered instrumentation and control (I and C) systems. They are designed to bring the plant in a secure and stable state without the help of any active system. Furthermore the passive safety features partially replace the active systems leading to a significant cost reduction and provide a reliable, safe and economically competitive alternative to standard plant design /1/. For further simplification of the plant design and additional cost reduction, the fuel pool cooling system has been modified in comparison to the currently running German BWR plants. This new system was tested in the Pressure Suppression Pool Vessel (PSPV) of the INKA test facility (Integral Teststand Karlstein) in Germany, which was originally build for the full scale testing of the key elements of the SWR 1000 passive safety concept /2/. The PSPV of INKA was chosen because it provides enough space for the cooler and its attached chimney (total height 11.5m). In this work the setup and the execution of the tests will be described. A characteristic diagram of the heat transfer capacity of the component as a function of cooling water temperature and fuel pool water temperature obtained form these experiments will be presented. In parallel CFD calculations, simulating the tests will be made. The results of these calculations and the comparison between the experimental and calculated results will be presented elsewhere and will serve furthermore to validate the CFD-code. (orig.)

  4. 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.

  5. Demolition of the Karlstein Superheated Steam Reactor Spent Fuel Pool, Germany

    International Nuclear Information System (INIS)

    The Karlstein superheated steam reactor started operation in October 1969. A number of failures in the fuel elements, associated with a concept design problem, resulted in final shutdown in April 1971. The reactor was then used as a test bed for reactor safety experiments from 1974 until the end of 1991. The spectrum of experiments simulating accidents under design and beyond design conditions included simulation of aeroplane crashes, earthquake experiments, material parameter studies, loss of coolant accidents (LOCAs), hydrogen distribution, hydrogen deflagration and burning experiments

  6. The Importance of the Regional Species Pool, Ecological Species Traits and Local Habitat Conditions for the Colonization of Restored River Reaches by Fish

    OpenAIRE

    Stoll, Stefan; Kail, Jochem; Lorenz, Armin W.; Sundermann, Andrea; Haase, Peter

    2014-01-01

    It is commonly assumed that the colonization of restored river reaches by fish depends on the regional species pools; however, quantifications of the relationship between the composition of the regional species pool and restoration outcome are lacking. We analyzed data from 18 German river restoration projects and adjacent river reaches constituting the regional species pools of the restored reaches. We found that the ability of statistical models to describe the fish assemblages established ...

  7. The Maple reactor project

    International Nuclear Information System (INIS)

    MDS Nordion supplies the majority of the world's reactor-produced medical isotopes. These isotopes are currently produced in the NRU reactor at AECL's Chalk River Laboratories (CRL). Medical isotopes and related technology are relied upon around the world to prevent, diagnose and treat disease. The NRU reactor, which has played a key role in supplying medical isotopes to date, has been in operation for over 40 years. Replacing this aging reactor has been a priority for MDS Nordion to assure the global nuclear medicine community that Canada will continue to be a dependable supplier of medical isotopes. MDS Nordion contracted AECL to construct two MAPLE reactors dedicated to the production of medical isotopes. The MDS Nordion Medical Isotope Reactor (MMIR) project started in September 1996. This paper describes the MAPLE reactors that AECL has built at its CRL site, and will operate for MDS Nordion. (author)

  8. 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 tested 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 US and for an undetermined number of nuclear plants located in foreign countries. This same cost-effective approach 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 evaluation program. This effort marks the first complete (non-trial) application of this state-of-the-art USI A-46 resolution methodology

  9. 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

  10. 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.

  11. 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

  12. Radiological consequences of a postulated cooling channel blockage incident at a pool-type research reactor

    International Nuclear Information System (INIS)

    An assessment of the radiological consequences of a postulated coolant flow blockage incident at the Hoger Onderwijs Reactor (HOR) is being presented. The HOR is a swimming-pool type research reactor with a maximum licensed power of 3 MW. Assuming a sudden blockage of cooling channels in the high power density region of the core, the source term for the release of radioactivity into the environment was calculated. The magnitude of this source term is required by actions of the HOR protection system as well as by physical processes acting on the fission products. Hence, almost 99% of the calculated release from the containment consists of noble gases; most of the aerosol-type activity set free in the environment results from the decay of these noble gases. The deposit of long-living radionuclides outside the reactor building is very low. Radio-iodine will be the main contributor to the environmental radiation dose, ingestion of contaminated food being the critical pathway. Despite the conservativeness of most assumptions used, the calculated thyroid dose for critical individuals at all distances from the site boundary remains well below the emergency reference levels recommended by national and international organizations and the national dose limit for members of the public

  13. Pressure drop calculation in a fuel element of a pool type reactor

    International Nuclear Information System (INIS)

    Even with the advances of hardware in computer sciences, sometimes it is necessary to simplify the simulation in order to optimize the results given the same calculation runtime. The object of this study is a thermodynamic analysis of the core of a pool type research reactor, focusing on natural circulation. Due to the high geometrical complexity of the core, the scale transfer process becomes an essential step to the thermodynamic study of the reactor. This process takes place by determining the effective equivalent properties obtained from a detailed simulation of the core and transferring them to a porous medium having a coarse mesh while preserving the overall characteristics. In this way, it will be able to obtain the quadratic resistance coefficient KQ by calculating the pressure drop inside the fuel element. To observe in detail the behavior of this flow, longitudinal and transversal cross sections will be made in different points, thereby observing the velocity and pressure distributions. The analysis will provide detailed data on the fluid flow between the fuel plates enabling the observation of possible critical points or undesired behavior. The whole analysis was made by using the commercial code ANSYS CFX ver. 12.1. This is study will provide data, as a first step to enable future simulations which will consider the entire reactor. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Iqbal, Masood [Nuclear Engineering Division, PINSTECH, P.O. Nilore, Islamabad (Pakistan)], E-mail: masiqbal@hotmail.com; Mahmood, Tayyab; Pervez, Showket [Nuclear Engineering Division, PINSTECH, P.O. Nilore, Islamabad (Pakistan)

    2008-03-15

    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 {beta}{sub eff} decreases with 1.33 x 10{sup -6}/% burnup whereas prompt neutron generation time increases with 6.42 x 10{sup -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.

  15. 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

  16. 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.

  17. 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

  18. Whole core analysis of an open pool research reactor under the most severe loss of coolant accident conditions

    International Nuclear Information System (INIS)

    In the present work the accident in which either the outlet or the inlet coolant pipe connected to the bottom of the reactor tank in an open pool research reactor is completely ruptured has been analyzed. The 3-D transient computer code ThEAP-I developed at Democritus NRC has been utilized and applied to the 5 MW Greek Research Reactor (GRR-1). The results show that a partial melting of the reactor core is possible for the GRR-1, the amount of melting being roughly and conservatively estimated to be of the order of 20%. (author)

  19. REX 2000 core : a new material testing reactor project

    International Nuclear Information System (INIS)

    REX 2000 is a new research reactor project entirely dedicated to technological irradiations, which should be located on the CEA site of CADARACHE. It will be aimed at satisfying the future needs for the validation of new concepts of nuclear materials and fuels, and will take over and replace the present experimental reactors, which are 30 to 40 years old. The fundamental studies started by the CEA in 1993, on future irradiation needs expected in 2005, lead to the design of a reactor which will essentially meet the needs of PWRs, without forgetting the other fields such as FBRs, fusion... The current reactor project is based on a light water open pool concept, with a thermal power of 100 MW, in about 150 l, and characterized by an in-core-central hole. It reaches neutronic flux levels twice those of present French reactor fluxes. It allows many irradiations in the central loop under high fast neutron flux, in order to accelerate the aging of materials and analyze their behaviour. It also enables the achievement of power transient tests under high thermal neutron flux gradients. These performances are obtained with high forced flow rates and upward flow in the core, in order to preserve the operating flexibility of the reactor. This leads to the design of a specific assembly design. (author)

  20. 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)

  1. 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

  2. Quality of water in the Red River alluvial aquifer; Pool 1, Red River waterway area, Vick, Louisiana

    Science.gov (United States)

    Smoot, C.W.; Seanor, R.C.; Huff, G.F.

    1994-01-01

    Water-quality changes in the Red River alluvial aquifer within the area affected by pool 1 near Vick, Louisiana, were monitored during pre-construction (1974-78) and post-construction (1984-92) of Lock and Dam 1. Changes greater or less than background values have occurred in an area within 2 miles of Lock and Dam 1, and in one well located about 10 miles west of Lock and Dam 1. Comparison between the pre-construction and post-construction water-quality analyses indicated the total hardness as calcium carbonate and concentrations of dissolved chloride, iron, and manganese generally have decreased in the Red River alluvial aquifer south of the Red River and near Lock and Dam l. The maximum decrease of the median total hardness as calcium carbonate was from 730 to 330 mg/L (milligrams per liter), dissolved chloride from 77 to 46 mg/L, dissolved iron from 18 to 6.9 mg/L, and dissolved manganese from 1.4 to 0.56 mg/L. Analyses of water from wells west of Lock and Dam 1 indicated an increase of the median total hardness as calcium carbonate was from 200 to 260 mg/L, and dissolved iron concentration was from 0.33 to 1.4 mg/L. North of the river and 1 mile west of Lock and Dam l, the median concentration of dissolved chloride increased from 45 to 130 mg/L in water from one well, and median total hardness as calcium cabonate and concentrations of dissolved iron and manganese also increased. Because well Ct-74 is completed in a sand that is in contact with a saltwater sand of Tertiary age, this increase is probably a temporal increase due to upconing after lowering the water level in the alluvial aquifer by pumping of dewatering wells during construction of Lock and Dam 1.

  3. 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)

  4. Radiation chemistry in the nuclear power reactor environment: from laboratory study to practical application

    International Nuclear Information System (INIS)

    This paper discusses the work carried out at the Chalk River Nuclear Laboratories in underlying and applied radiation chemical research performed to optimise the processes occurring in the four aqueous systems in and around the core. The aqueous systems subject to radiolysis in CANDU reactors are Heat Transport System, Moderator, Liquid Zone Controls and End Shields.

  5. 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

  6. NRU turns 50! World's best research reactor still going strong after a half century

    International Nuclear Information System (INIS)

    This article gives a history of the NRU reactor at Chalk River Nuclear Labs (CRL), Atomic Energy of Canada Ltd. The NRU reactor design was started in 1949 when CRL was Atomic Energy Project of the National Research Council. It was and still is a national facility used by scientists across Canada. As well as providing a source of neutrons for research, it was a home for much of the research required to develop the CANDU reactor design for nuclear power stations.

  7. Characterization of radioactive contaminants and water treatment trials for the Taiwan Research Reactor's spent fuel pool

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Chun-Ping, E-mail: chunping@iner.gov.tw [Institute of Nuclear Energy Research, 1000, Wenhua Road, Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan, ROC (China); Lin, Tzung-Yi; Chiao, Ling-Huan; Chen, Hong-Bin [Institute of Nuclear Energy Research, 1000, Wenhua Road, Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan, ROC (China)

    2012-09-30

    Highlights: Black-Right-Pointing-Pointer Deal with a practical radioactive contamination in Taiwan Research Reactor spent fuel pool water. Black-Right-Pointing-Pointer Identify the properties of radioactive contaminants and performance test for water treatment materials. Black-Right-Pointing-Pointer The radioactive solids were primary attributed by ruptured spent fuels, spent resins, and metal debris. Black-Right-Pointing-Pointer The radioactive ions were major composed by uranium and fission products. Black-Right-Pointing-Pointer Diatomite-based ceramic depth filter can simultaneously removal radioactive solids and ions. - Abstract: There were approximately 926 m{sup 3} of water contaminated by fission products and actinides in the Taiwan Research Reactor's spent fuel pool (TRR SFP). The solid and ionic contaminants were thoroughly characterized using radiochemical analyses, scanning electron microscopy equipped with an energy dispersive spectrometer (SEM-EDS), and inductively coupled plasma optical emission spectrometry (ICP-OES) in this study. The sludge was made up of agglomerates contaminated by spent fuel particles. Suspended solids from spent ion-exchange resins interfered with the clarity of the water. In addition, the ionic radionuclides such as {sup 137}Cs, {sup 90}Sr, U, and {alpha}-emitters, present in the water were measured. Various filters and cation-exchange resins were employed for water treatment trials, and the results indicated that the solid and ionic contaminants could be effectively removed through the use of <0.9 {mu}m filters and cation exchange resins, respectively. Interestingly, the removal of U was obviously efficient by cation exchange resin, and the ceramic depth filter composed of diatomite exhibited the properties of both filtration and adsorption. It was found that the ceramic depth filter could adsorb {beta}-emitters, {alpha}-emitters, and uranium ions. The diatomite-based ceramic depth filter was able to simultaneously

  8. Virtual manipulation of topography to test potential pool-riffle maintenance mechanisms

    Science.gov (United States)

    Jackson, J. R.; Pasternack, G. B.; Wheaton, J. M.

    2015-01-01

    In this study, numerical experimentation with two-dimensional hydraulic modeling of pool-riffle river topography drawing on the testbed data from the classic Keller (1971) study was used to investigate the effect of synthetically manipulating topography on the occurrence and magnitude of velocity and Shields stress reversals in a pool-riffle sequence. Reversals in velocity and shear stress have been used to explore mechanisms of pool-riffle maintenance, while Shields stress (a combined measure of transport capacity and substrate erodibility) is emerging in importance. The original site topography was modeled alongside six altered ones to evaluate the sensitivity of hydraulic reversals to subtle morphology - five incrementally wider pools and a filled pool. The Caamaño (2009) criterion, a simplified geometric threshold for predicting velocity reversals, was applied to each terrain to evaluate its utility. The original pool-riffle topography was just over the threshold for a velocity reversal and well over the threshold for a strong Shields stress reversal. Overall, pool widening caused a predominantly local response, with change to pool hydraulics and no change in section-averaged velocity in the riffle beyond the initial widening of 10%. Filling in the pool significantly increased the magnitude of reversals, whereas expanding it eliminated the occurrence of a reversal in mean velocity, though the Shields stress reversal persisted because of strong differentiation in bed material texture. Using Shields stress as a reversal parameter enabled the quantification of pool modification effects on pool-riffle resiliency. The Caamaño (2009) criterion accurately predicted reversal occurrence for the altered terrains with exaggerated effects, but failed to predict the weak reversal for the original topography. Two-dimensional modeling coupled with previously accepted hydrologic, geomorphic, and engineering analyses is vital in project design and evaluation prior to

  9. 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

  10. 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.

  11. 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

  12. 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

  13. 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

  14. 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...... prediction, so we have investigated the use of velocity data to predict permeability. The compressional velocity fromwireline logs and core plugs of the chalk reservoir in the South Arne field, North Sea, has been used for this study. We compared various methods of permeability prediction from velocities....... The relationships between permeability and porosity from core data were first examined using Kozeny’s equation. The data were analyzed for any correlations to the specific surface of the grain, Sg, and to the hydraulic property defined as the flow zone indicator (FZI). These two methods use two...

  15. 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. 

  16. 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.

  17. Advanced Test Reactor National Scientific User Facility

    International Nuclear Information System (INIS)

    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.

  18. Stratigraphy of the unsaturated zone and uppermost part of the Snake River Plain Aquifer at the Idaho Chemical Processing Plant and Test Reactors Area, Idaho National Engineering Laboratory, Idaho

    International Nuclear Information System (INIS)

    A complex sequence of basalt flows and sedimentary interbeds underlies the Idaho Chemical Processing Plant and Test Reactors Area at the Idaho National Engineering Laboratory in eastern Idaho. Wells drilled to a depth of 700 feet penetrate a sequence of 23 basalt-flow groups and 15 to 20 sedimentary interbeds that range in age from 200,000 to 640,000 years. The 23 flow groups consist of about 40 separate basalt flows and flow units. Each flow group is made up of one to three petrographically similar basalt flows that erupted from related source areas during periods of less than 200 years. Sedimentary interbeds consist of fluvial, lacustrine, and eolian deposits of clay, silt, sand, and gravel that accumulated during periods of volcanic inactivity ranging from thousands to hundreds of thousands of years. Multiple flow groups and sedimentary interbeds of similar age and source form seven composite stratigraphic units with distinct upper and lower contacts. Composite units older than 350,000 years were tilted, folded, and fractured by differential subsidence and uplift. Basalt and sediment of this sequence are unsaturated to a depth that ranges from 430 to 480 feet below land surface. Basalt and sediment in the lower part of the sequence are saturated and make up the uppermost part of the Snake River Plain aquifer. Stratigraphic relations in the lowermost part of the aquifer below a depth of 700 feet are uncertain. 23 refs., 22 figs., 1 tab

  19. Calculated and measured behaviour of zircaloy fuel sheaths during in-reactor LOCA tests

    International Nuclear Information System (INIS)

    Six CANDU type fuel elements, containing UO2 fuel and sheathed in Zircaloy, have been subjected to transient conditions simulating a hypothetical large break loss of coolant accident (LOCA) in a CANDU reactor. The maximum transient sheath temperature was about 1273 K. Two single-element experiments were conducted at Chalk River Nuclear Laboratories in the X-2 loop of the NRX reactor. The other four elements experienced a single transient in the Power Burst Facility at Idaho National Engineering Laboratory. Comparisons are presented between data from these experiments and calculations by ELOCA, a computer code simulating fuel performance during transient conditions. The parameters evaluated included fuel sheath strain, internal element gas pressure, the mechanisms and timing of fuel element failure, fuel centreline temperature, sheath microstructure, and the thicknesses of zirconia and oxygen stabilized alpha-Zr layers on the sheaths. ELOCA calculations agreed well with the data. Some of the work described here was jointly funded by Atomic Energy of Canada Limited and Ontario Hydro, a Canadian utility, through the co-operative research and development program, CANDEV

  20. Development of a fuel failure monitoring method for a pool-type research reactor

    International Nuclear Information System (INIS)

    Studies on developing a sensitive monitoring method of possible release of fission products (FP) from a fuel element have been made for a pool-type research reactor. It consists of introducing gas bubbles into reactor coolant water to extract effectively the dissolved fission rare gases, 89Kr and 138Xe, produced somewhere in the core, and counting their respective daughter nuclides, 89Rb and 138Cs with high efficiency. The measurements were done by either method, (I) on a filter paper by sucking the bubbled gas and air covering water of the reactor tank, or (II) in the washing water of bubbled gas sampled into a bottle at the water surface. The followings are the summary of the results obtained. (1) DE increased as much as 30 times or more compared with no gas bubbling. (2) DE largely increased with increasing flow rate of introducing gas. (3) DE increased with increasing depth of the gas exit in the water. (4) DE at the same depth depended on the position of gas exit. It was larger for 'side' position than for 'center', due to the water convection in the tank. (5) DE largely depended on the condition of whether the primary cooling system was operated or not. (6) In Method II, DE depended on the time of standing, and it showed maximum at the theoretically predicted value. (7) The theoretical analysis for the effect of depth suggests that DE should be proportional to the value =(1+D /1033)2-(1+D /1033)5/3=, where D is the depth (in cm). The trend agreed at least partly with the observed data. (8) As a continuous mode experiment, we constructed an automatic fuel monitoring system for routine use by adopting Method II. It is composed of an intermittent sampling of the bubbling gas into bottle at the water surface, washing it with water after definite time of standing, and measuring the nuclides contained in the water. (J.P.N.)

  1. 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.

  2. 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.

  3. 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 ...

  4. 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)

  5. The RES Reactor. A test reactor for the French naval propulsion

    International Nuclear Information System (INIS)

    In the Cadarache nuclear research centre the French Atomic Energy Commission (CEA) operates, with the support of TECHNICATOME as nuclear operator, the experimental facilities which are necessary for the French naval propulsion program. Since the sixties these facilities have brought a large contribution to the development and to the technical support for the nuclear propulsion; they have been used also to train the French Navy operators. The last experimental reactor, the RNG, is now at the end of its life cycle after thirty years of a profitable operation. A replacement reactor is needed to sustain any evolution of the naval propulsion reactors as well as to guarantee a safe operation and a high level of availability of the existing onboard reactors. The aim of the RES program is namely to build such a test facility. Its construction program started in 2003. By the year 2009 the RES reactor will take over the mission of the RNG. We present hereafter: - A brief history of the French experimental reactors built in support to the naval propulsion, - The needs of the naval propulsion and the related objectives of the RES program, - The corresponding architecture and main characteristics of the RES facility, - The current status of the RES construction. The contents of the paper is as follows: 1. Introduction; 2. History of the French nuclear propulsion experimental reactors; 3. Needs of the naval propulsion and related objectives of the RES reactor; 4. RES architecture and main characteristics; 4.1. The pool module; 4.2. The reactor module; 4.3. The RES reactor, an innovative concept; 5. Realisation status; 6. Conclusion. To summarize, from the year 2009 the RES will be an efficient facility available for irradiation and qualification programs. Its large experimental capabilities will allow relevant fuel and core irradiations. This will give access to a real progress in the knowledge of fuel and core physics as well as in the related simulation tools. This reactor

  6. Dosimetric Implications of Atmospheric Dispersal of Tritium Near a Heavy-water Research Reactor Facility

    International Nuclear Information System (INIS)

    An estimate of the tritium dose to the public in the vicinity of the heavy water research reactor facility at AECL-Chalk River Laboratories, Ontario, Canada, has largely been accomplished from analyses on regularly-collected samples of air, precipitation, drinking water and foodstuffs (pasture, fruit, vegetables and milk) and environmental dose models. To increase the confidence with which public doses are calculated, tritium doses were estimated directly from the ratio of tritiated species in urine samples from members of the general public. Single cumulative 24 h urine samples from a few adults living in the vicinity of the heavy-water research reactor facility at Chalk River Laboratories, Canada were collected and analysed for tritiated water and organically bound tritium. The participants were from Ottawa (200 km east), Deep River (10 km west) and Chalk River Laboratories. Tritiated water concentrations in urine ranged from 6.5 Bq.l-1 for the Ottawa resident to 15.9 Bq.l-1 for the Deep River resident, and were comparable to the ambient levels of tritium-in-precipitation at their locations. The ultra-low levels of organically bound tritium in urine from these same individuals were measured by 3He-ingrowth mass spectrometry and were 0.06 Bq.l-1 (Ottawa) and 0.29 Bq.l-1 (Deep River). For Chalk River Laboratories workers, tritiated water concentrations in urine ranged from 32 Bq.l-1 to 9.2x104 Bq.l-1, depending on the ambient levels of tritium in their workplace. The organically bound tritium concentrations in urine from the same workers were between 0.08 Bq.l-1 and 350 Bq.l-1. With a model based on the ratio of tritiated water to organically bound tritium in urine, the estimated dose arising from organically bound tritium in the body for the Ottawa and Deep River residents was about 26% and 50%, respectively, of the body water tritium dose. The workers in a reactor building at Chalk River Laboratories has less than 10% dose contribution from organically bound

  7. NRU turns 50{exclamation_point} World's best research reactor still going strong after a half century

    Energy Technology Data Exchange (ETDEWEB)

    Boyd, F. [Canadian Nuclear Society, Toronto, Ontario (Canada)

    2007-12-15

    This article gives a history of the NRU reactor at Chalk River Nuclear Labs (CRL), Atomic Energy of Canada Ltd. The NRU reactor design was started in 1949 when CRL was Atomic Energy Project of the National Research Council. It was and still is a national facility used by scientists across Canada. As well as providing a source of neutrons for research, it was a home for much of the research required to develop the CANDU reactor design for nuclear power stations.

  8. Test Facility for SMART Reactor Flow Distribution

    International Nuclear Information System (INIS)

    A Reactor Flow Distribution Test Facilities for SMART, named SCOP (SMART Core Flow and Pressure Test Facility), were designed in order to simulate the distributions of (1) core flow and (2) reactor sectional flow resistance and flow rates. SCOP facility was designed based on the linear scaling law in order to preserve the flow characteristics of the prototype system, which are distributions of flow rate and pressure drop. The reduced scale was selected as a 1/5 of prototype length scale. The nominal flow condition was designed to be similar based on the velocity as that of the SMART reactor, which can minimize the flow distortion in the reduced scale of test facility by maintaining high Re number flow. Test facility includes fluid system, control/instrumentation system, data acquisition system, power system, which were designed to meet the requirement for each system. This report describes the details of the scaling and design features for the test facility

  9. 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)

  10. 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.

  11. 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

  12. Savannah River Site TEP-SET tests uncertainty report

    International Nuclear Information System (INIS)

    This document presents a measurement uncertainty analysis for the instruments used for the Phase I, II and III of the Savannah River One-Fourth Linear Scale, One-Sixth Sector, Tank/Muff/Pump (TMP) Separate Effects Tests (SET) Experiment Series. The Idaho National Engineering Laboratory conducted the tests for the Savannah River Site (SRS). The tests represented a range of hydraulic conditions and geometries that bound anticipated Large Break Loss of Coolant Accidents in the SRS reactors. Important hydraulic phenomena were identified from experiments. In addition, code calculations will be benchmarked from these experiments. The experimental system includes the following measurement groups: coolant density; absolute and differential pressures; turbine flowmeters (liquid phase); thermal flowmeters (gas phase); ultrasonic liquid level meters; temperatures; pump torque; pump speed; moderator tank liquid inventory via a load cells measurement; and relative humidity meters. This document also analyzes data acquisition system including the presampling filters as it relates to these measurements

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

    International Nuclear Information System (INIS)

    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

  14. Reliability tests for reactor internals replacement technology

    International Nuclear Information System (INIS)

    Structural damage due to aging degradation of LWR reactor internals has been reported in several nuclear plants. NUPEC has started a project to test the reliability of the technology for replacing reactor internals, which was directed at preventive maintenance before damage and repair after damage for the aging degradation. The project has been funded by the Ministry of International Trade and Industry (MITI) of Japan since 1995, and it follows the policy of a report that the MITI has formally issued in April 1996 summarizing the countermeasures to be considered for aging nuclear plants and equipment. This paper gives an outline of the whole test plans and the test results for the BWR reactor internals replacement methods; core shroud, ICM housing, and CRD Housing and stub tube. The test results have shown that the methods were reliable and the structural integrity was appropriate based on the evaluation. (author)

  15. 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.

  16. Startup testing of Romania dual-core test reactor

    International Nuclear Information System (INIS)

    Late in 1979 both the Annular Core Pulsed Reactor (ACPR) and the 14-MW steady-state reactor (SSR) were loaded to critical. The fuel loading in both was then carried to completion and low-power testing was conducted. Early in 1980 both reactors successfully underwent high-power testing. The ACPR was operated for several hours at 500 kW and underwent pulse tests culminating in pulses with reactivity insertions of $4.60, peak power levels of about 20,000 MW, energy releases of 100 MW-sec, and peak measured fuel temperatures of 830 deg. C. The SSR was operated in several modes, both with natural convection and forced cooling with one or more pumps. The reactor successfully completed a 120-hr full-power test. Subsequent fuel element inspections confirmed that the fuel has performed without fuel damage or distortion. (author)

  17. 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

  18. 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

  19. 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

  20. 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).

  1. 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.

  2. Reactor group constants and benchmark test

    International Nuclear Information System (INIS)

    The evaluated nuclear data files such as JENDL, ENDF/B-VI and JEF-2 are validated by analyzing critical mock-up experiments for various type reactors and assessing applicability for nuclear characteristics such as criticality, reaction rates, reactivities, etc. This is called Benchmark Testing. In the nuclear calculations, the diffusion and transport codes use the group constant library which is generated by processing the nuclear data files. In this paper, the calculation methods of the reactor group constants and benchmark test are described. Finally, a new group constants scheme is proposed. (author)

  3. 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

  4. Migration of activation products in discharges pool in T.H.O.R. nuclear research reactor site boundary

    International Nuclear Information System (INIS)

    Due to material degradation in tubing, the heat exchanger for the primary coolant of the 1 MW Tsing Hua Open-pool Reactor(THOR) has minor leakage to the secondary cooling system since 1970. In the past twenty years, trace amount of Co-58, 60, Cr-51, Cs-137, Mn-54, Sc-46, and Zn-65 have been leaked through the cooling system, accumulated and trapped eventually in the discharge pool right in front of the THOR facility. The distribution of these activation products in mud at different depths and various locations in the pool was measured with standard procedures of radioactive soil sampling and counting techniques. Concentration of activation products, with no more than 40 kBq/kg. dry at the hottest spot, was contour-mapped to reveal the migration of these trace level radioactive products in a period of 20 years

  5. Performance analyses of the pool-top radiation level reduction systems at the ETRR-2 research reactor

    International Nuclear Information System (INIS)

    An analysis of the performance of the hot water layer system, HWLS, and the interconnection system, IS, at the ETRR-2 research reactor is presented. The behavior of the HWLS during the formation period of the hot water layer has been studied and is presented together with the pool-top measured radiation level associated with that behavior. Two different designs of the HWLS have been experimentally evaluated. For these two different designs, the radiation level at the ETRR-2 pool-top were characterized and discussed. The effect of adding a water purification system to the HWLS is demonstrated as well as the effect of the IS on the pool-top radiation level considering different operational conditions of the HWLS. The study shows the importance of these systems from the radiological protection point of view. (orig.)

  6. 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.

  7. Materials R and D with neutron beams - how the NRU reactor serves Canada further as a unique resource for science and industry

    International Nuclear Information System (INIS)

    This presentation discusses the use of NRU reactor for materials research and development with neutron beams at the Canadian Neutron Beam Centre at the Chalk River Laboratories. The facility has 5 beams for research and development on hard materials, 1 beam for research and development on nano-film and 1 beam for research and development on nano-solution, still under development.

  8. Materials R and D with neutron beams - how the NRU reactor serves Canada further as a unique resource for science and industry

    Energy Technology Data Exchange (ETDEWEB)

    Root, J. [National Research Council Canada, Canadian Neutron Beam Centre, Ottawa, Ontario (Canada)

    2010-07-01

    This presentation discusses the use of NRU reactor for materials research and development with neutron beams at the Canadian Neutron Beam Centre at the Chalk River Laboratories. The facility has 5 beams for research and development on hard materials, 1 beam for research and development on nano-film and 1 beam for research and development on nano-solution, still under development.

  9. 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

  10. Simulation of the gamma dose rate in a loss of pool water accident of the second Egyptian research reactor ET-RR-2

    International Nuclear Information System (INIS)

    The second Egyptian research reactor ET-RR-2, is a pool type reactor. A sudden loss of pool water would leave the core region uncovered. The reactor core is surrounded by chimney chambers with water isolated from the pool water. This accident would lead to significant external doses. A model is developed and used to calculate the dose rates for key access-areas and traffic plans from indirect line of sight of the core which have a maximum dose rate. The model developed uses the discrete ordinate method as implemented in the code DOT3.5. (orig.)

  11. New pharmacokinetic methods. III. Two simple test for deep pool effect

    International Nuclear Information System (INIS)

    If a portion of administered drug is distributed into a deep peripheral compartment, the drug's actual elimination half-life during the terminal exponential phase of elimination may be longer than determined by a single dose study or a tracer dose study (deep pool effect). Two simple methods of testing for deep pool effect applicable to drugs with either linear or nonlinear pharmacokinetic properties are described. The methods are illustrated with stable isotope labeled (13C15N2) tracer dose studies of phenytoin. No significant (P less than .05) deep pool effect was detected

  12. 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.

  13. A review of experiments and results from the transient reactor test (TREAT) facility

    International Nuclear Information System (INIS)

    The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop

  14. 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

  15. Logic programming for operational analysis of the Savannah River reactors

    International Nuclear Information System (INIS)

    The Savannah River Plant (SRP) has installed an on-line reactor monitoring diagnostic computer system to direct operators to appropriate procedures for multiple, concurrent alarms. Off-line programs analyze the monitoring logic for consistency. Both applications can be described as logic programs. Emerging artificial intelligence (AI) computer technology promises operationally viable expert systems for diagnosis and control of reactor systems. The core of this technology is the use of computers for logical inference, in contrast to the more traditional number crunching and data processing functions. At present, the diagnostic logic represents surface knowledge of reactor operating procedures, while the analytical logic incorporates a small amount of real-world knowledge concerning superset/subset relationships among alarms. The challenge for the future is to create a comprehensive operations advisory system, with logic as engineering rules of thumb for cause/effect models of malfunction and corrective action, data reflecting the real-time status of all critical reactor components, and control working both bottom up from alarms to malfunctions, and top down from malfunction to corrective procedure

  16. 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)

  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. 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)

  19. 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).

  20. 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%

  1. Suppression pool testing at the SIET laboratory (4). Release of fission products into the environment under severe accident conditions

    International Nuclear Information System (INIS)

    Long term effects of radioactive cesium on the environments and the contaminated water are one of the key issues for restoration of the Fukushima Daiichi nuclear power plant (NPP) Accident. In order to evaluate the cesium sources and their behaviors, source terms under the severe accident at Fukushima Daiichi NPP were discussed from both viewpoints of short and long term fission product (FP) sources. The former was evaluated by analyzing radioactive species based on monitoring post data, which suggested that one of major FP sources was from wet-well venting for decreasing the primary containment vessel (PCV) pressure. The latter was evaluated by analyzing long term trends of the contaminated water in the reactor and turbine buildings, which suggested that FP concentrations in the contaminated water during the 2 years since the accident were determined by the short term FP sources, while their saturated concentrations, due to a balance between the release from the reactor and the clean-up, were determined by the long term FP sources. In order to determine the PCV water scrubbing effects on FP removal, two kinds of experiment were carried out. A mini scale scrubbing tests based on a 1 L of grass made pool with 0.02 kg/h of steam, 0.04 kg/h of carrier gas and cesium, iodine and hematite tracers and a large scale mock-up tests based on 1000L of transparent pool with 360 kg/h of steam and hematite tracer. As a result of the mini scale scrubbing tests, it was evaluated that the steam carry over rates of cesium during pool scrubbing around the boiling temperature was 50% and that during sub-cooled boiling it was about 30%, which was also confirmed by the mock-up experiments. The chemical forms of the long term cesium source in the reactor have not been determined yet. Survey of core debris and cesium remained in the reactor and the PCV is one of most importance issues to understand the FP source term behaviors during the severe accident conditions. (author)

  2. 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

  3. 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

  4. Dynamic simulation of a two-phase control absorber for neutron flux regulation in a nuclear reactor

    International Nuclear Information System (INIS)

    A dynamic simulation of the two-phase control absorber being proposed for future Canadian nuclear power reactors has been developed at Chalk River Nuclear Laboratories. The model, implemented on a hybrid computer, was developed to study absorber dynamics at different circuit operating conditions and with different circuit configurations. The simulation is modular, with as much correspondence as possible between individual modules and the physical entities. The dynamics of several of the modules are described by partial differential equations, with space and time as independent variables. These are solved via the Continuous Space/Discrete Time technique. The simulation has been validated with data from the Two-Phase Absorber Experimental (TOPAX) Rig installed at the ZED-2 test reactor. (author)

  5. In-reactor testing of ionic thermometers

    International Nuclear Information System (INIS)

    Ionic thermometers have been tested in a nuclear reactor with attention to the steepness of the ionic conductivity jump and the influence of a glass container on the accuracy of the temperature measurements. It was found that, at the neutron fluxes up to 1.5 x 1018 m-2 s-1 (thermal) and 3 x 1018 m-2 s-1 (fast) in a light water reactor, the change of conductivity jump slope is negligible or nil for an ionic thermometer filled by HgI2, i.e., at 256.0 +- 0.2 0C. The need to use boron-free glass was confirmed. The impact on the accuracy of the temperature point indication in a nuclear reactor core is discussed, as well as obvious inertness of the melting process mechanism to the intense irradiation field

  6. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Bakhtyar, S.; Iqbal, M.; Israr, M.; Pervez, S.; Salahuddin, A. [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)

    2004-07-01

    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)

  8. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Buck, M.; Buerger, M. [Univ Stuttgart, Inst Kernenerget and Energiesyst, D-70569 Stuttgart (Germany); Miassoedov, A.; Gaus-Liu, X.; Palagin, A. [IRSN Forschungszentrum Karlsruhe GmbH, D-76021 Karlsruhe, (Germany); Godin-Jacqmin, L. [CEA Cadarache, DEN STRI LMA, F-13115 St Paul Les Durance (France); Tran, C. T.; Ma, W. M. [KTH, AlbaNova Univ Ctr, S-10691 Stockholm (Sweden); Chudanov, V. [Nucl Safety Inst, Moscow 113191 (Russian Federation)

    2010-07-01

    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 (KNO{sub 3}-NaNO{sub 3}) 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

  10. 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

  11. Fission Product Release from Molten Pool: ceramic melt tests

    International Nuclear Information System (INIS)

    Experimental results are presented on the volatilisation of UO2±x, SrO, BaO, CeO2 from corium melts. Corium melts were generated by high frequency induction melting in a cold crucible. The surface temperature of the melts was in the range from 1753 to 3023 K. Some results of the tests are discussed and a comparison with published data is made. (author)

  12. Fission Product Release from Molten Pool: ceramic melt tests

    Energy Technology Data Exchange (ETDEWEB)

    Petrov, Yu.B.; Lopukh, D.B.; Petchenkov, A.Yu. [AO ' NP Sintez' , St. Petersburg (RU)] [and others

    1999-07-01

    Experimental results are presented on the volatilisation of UO{sub 2{+-}}{sub x}, SrO, BaO, CeO{sub 2} from corium melts. Corium melts were generated by high frequency induction melting in a cold crucible. The surface temperature of the melts was in the range from 1753 to 3023 K. Some results of the tests are discussed and a comparison with published data is made. (author)

  13. 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

  14. The importance of the regional species pool, ecological species traits and local habitat conditions for the colonization of restored river reaches by fish.

    Directory of Open Access Journals (Sweden)

    Stefan Stoll

    Full Text Available It is commonly assumed that the colonization of restored river reaches by fish depends on the regional species pools; however, quantifications of the relationship between the composition of the regional species pool and restoration outcome are lacking. We analyzed data from 18 German river restoration projects and adjacent river reaches constituting the regional species pools of the restored reaches. We found that the ability of statistical models to describe the fish assemblages established in the restored reaches was greater when these models were based on 'biotic' variables relating to the regional species pool and the ecological traits of species rather than on 'abiotic' variables relating to the hydromorphological habitat structure of the restored habitats and descriptors of the restoration projects. For species presence in restored reaches, 'biotic' variables explained 34% of variability, with the occurrence rate of a species in the regional species pool being the most important variable, while 'abiotic' variables explained only the negligible amount of 2% of variability. For fish density in restored reaches, about twice the amount of variability was explained by 'biotic' (38% compared to 'abiotic' (21% variables, with species density in the regional species pool being most important. These results indicate that the colonization of restored river reaches by fish is largely determined by the assemblages in the surrounding species pool. Knowledge of species presence and abundance in the regional species pool can be used to estimate the likelihood of fish species becoming established in restored reaches.

  15. Sipping tests for the irradiated fuel elements of the TR-2 research reactor

    International Nuclear Information System (INIS)

    Sipping tests have been performed for fuel elements of the TR-2 reactor at Cekmece Nuclear Research and Training Center (CNRTC), in order to find out which one failed in the core. A sipping assembly has been constructed and placed in the pool of the TR-2 reactor. The assembly identifies leaking fuel elements by collecting and measuring 137Cs that leak out from the defective fuel elements. 31 fuel elements in the reactor have been tested for the clad integrity. The measured 137Cs activity of the fuel element with an identification number S-104 is a 10247 Bq/(0.3 l). This value is approximately 234 times greater than the average of the other tested fuel elements in the reactor. (orig.)

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

    International Nuclear Information System (INIS)

    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

  17. 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.

  18. Fabrication, testing, and qualification of reactor graphites

    International Nuclear Information System (INIS)

    The work performed under the HBK project for development and testing of reactor graphites could have recourse to results and experience already gained in Great Britain, in the F.R.G., the USA, and the Netherlands. The specific problems to be tackled by the HBK project activities result from the particularly exacting requirements with regard to behaviour under irradiation that are to be met by the graphite reflector for the THTR follower plant. From a great number of candidate graphites, selected for testing and evaluation, the extensive irradiation experiments revealed a variety of graphites best suited to the various tasks in mind, as defined by the operational conditions. The tests examined radiation-induced changes of linear dimension, E-module, thermal expansion, and heat conductivity, as well as radiation-induced creep and corrosion in reactor graphites under specified normal and under accident conditions. The work performed also includes tests for defining design criteria for reactor graphite components. The goals have been achieved, but further work will be necessary, as new requirements are taking shape in the course of current THTR follower plant development. (orig.)

  19. Nuclear Safety Research Reactor (NSRR) as a facility for reactor safety research and its modification for the future test plan

    International Nuclear Information System (INIS)

    The NSRR is a modified TRIGA-ACPR (annular core pulse reactor), and attained the initial criticality in May, 1975. It was built for studying reactor fuel behavior under a reactivity-initiated accident condition. The reactor is installed in a pool of 3.6 m width, 4.5 m length and 9 m depth, and water above the reactor core serves as a radiation shield. The reactor core contains 149 driver fuel rods, 6 regulating rods, 2 safety rods and 3 transient rods. An arbitrary reactivity up to 4.67 $ can be set up almost instantaneously in the reactor core. The pulse power generation is terminated by the large negative reactivity induced by prompt temperature feedback without inserting the control rods. This is brought about by an excellent property of the driver fuel which contains 12 wt.% U-ZrH enriched to 20 wt.% U-235. As a unique feature, the NSRR is equipped with a big experimental cavity through the center of the reactor core. It has the diameter of 220 mm, and is called loading tube. It is branched into a vertical loading tube and an offset loading tube. The characteristics of the pulse operation in the NSRR, the outline of fuel irradiation experiment, the future test plan and the modification of the NSRR are described. (Kako, I.)

  20. Calculation and mapping of gamma radiation field in the pool of Apsara reactor

    International Nuclear Information System (INIS)

    Theoretical simulation of the radiation transport occurring in the Apsara core and bulk shield was carried out using two different radiation transport codes, MCNP and QADCG. The MCNP is a Monte Carlo based statistical method solving Boltzmann transport equation, where as the latter code QADCG is a point kernel based deterministic method with build-up factor correction. The aim of the simulation was to do a dose mapping and estimate the expected value of gamma dose rates at various locations where experimental measurements were conducted. Details regarding the simulation techniques employed by both the MCNP and QADCG software with reference to the Apsara core and shield geometry and source gamma energy distribution in the fuel plates are presented in this report. Different types of particle tallies requested in MCNP and QADCG are discussed. Details of variance reduction methods employed in reducing the statistical uncertainty of Monte Carlo simulation are also mentioned in the report. The statistical errors associated with Monte Carlo based simulation varied between 3% - 6% in most of the energy bins that contribute to the total fluence and hence to the dose rates. It was observed that the experimental values and the theoretically simulated values match each other closely following a similar trend except for certain experimental locations which had photon flux contributions from extraneous sources like the N-16 activity present in water, beam tubes and pool liner towards shielding corner. It is seen that the theoretical values are found to be larger than experimental values by factors ranging from 1.1 to 3 depending on the water shield thickness. This study served in validation of the experimental measurements conducted by GM counter based teletector and dipole based detectors. In addition, the comparison provided a confirmation of the accuracy of the radiation transport simulation techniques used for dose rate evaluation in case of complex source geometries and

  1. 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)

  2. 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.

  3. 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.

  4. Seismic appraisal test of control rod drive mechanism of China experiment fast reactor

    International Nuclear Information System (INIS)

    The structure of the control rod drive mechanism in pool type sodium-cooled fast reactor is the characterized by long, thin, and geometric nonlinearity, and the seismic load is multiple activation. The anti-seismic evaluation is always paid great attention by the countries developing the technology worldwide. This article introduces the seismic appraisal test of the control rod drive mechanism of China Experimental Fast Reactor (CEFR) performed on a seismic platform which is vertical shaft style and multiple activation. The result of the test shows the structural integrity and the function of the control rod drive mechanism could meet the design requirements of the earthquake intensity. (authors)

  5. Testing of the Micro-Reactor System

    Czech Academy of Sciences Publication Activity Database

    Krystyník, Pavel; Beneš, Ondřej; Klusoň, Petr; Šolcová, Olga

    Praha: Česká společnost průmyslové chemie, 2015, s. 30 /p104./. ISBN 978-80-86238-73-9. [mezinárodní chemicko-technologická konference (ICCT 2015) /3./. Mikulov (CZ), 13.04.2015-15.04.2015] R&D Projects: GA ČR GA15-14228S Institutional support: RVO:67985858 Keywords : micro-reactor technology * heat transfer * testing Subject RIV: CI - Industrial Chemistry, Chemical Engineering

  6. Testing of the Micro-Reactor System

    Czech Academy of Sciences Publication Activity Database

    Beneš, Ondřej; Hanková, Libuše; Klusoň, Petr; Šolcová, Olga

    Bratislava: Slovak Society of Chemical Engineering, 2015 - (Markoš, J.), s. 40 ISBN 978-80-89475-14-8. [International Conference of Slovak Society of Chemical Engineering /42./. Tatranské Matliare (SK), 25.05.2015-29.05.2015] R&D Projects: GA ČR GA15-14228S Institutional support: RVO:67985858 Keywords : micro-reactor technology * testing * partial oxidation Subject RIV: CI - Industrial Chemistry, Chemical Engineering

  7. 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

  8. Unusual occurrences in fast breeder test reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe sodium cooled mixed carbide fuelled reactor. Its main aim is to generate experience in the design, construction and operation of fast reactors including sodium systems and to serve as an irradiation facility for the development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct 85 with Mark I core (70% PuC - 30% UC). Steam generator was put in service in Jan 93 and power was raised to 10.5 MWt in Dec 93. Turbine generator was synchronised to the grid in Jul 97. The indigenously developed mixed carbide fuel has achieved a burnup of 44,000 MW-d/t max at a linear heat rating of 320 W/cm max without any fuel clad failure. The commissioning and operation of sodium systems and components have been smooth and performance of major components, viz., sodium pumps, intermediate heat exchangers and once through sodium heated steam generators (SG) have been excellent. There have been three minor incidents of Na/NaK leaks during the past 14 years, which are described in the paper. There have been no incident of a tube leak in SG. However, three incidents of water leaks from water / steam headers have been detailed. The plant has encountered some unusual occurrences, which were critically analysed and remedial measures, in terms of system and procedural modifications, incorporated to prevent recurrence. This paper describes unusual occurrences of fuel handling incident of May 1987, main boiler feed pump seizure in Apr 1992, reactivity transients in Nov 1994 and Apr 1995, and malfunctioning of the core cover plate mechanism in Jul 1995. These incidents have resulted in long plant shutdowns. During the course of investigation, various theoretical and experimental studies were carried out for better understanding of the phenomena and several inspection techniques and tools were developed resulting in enriching the technology of sodium cooled reactors. FBTR has 36 neutronic and process

  9. 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.

  10. 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,...

  11. 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,...

  12. 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,...

  13. 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,...

  14. 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,...

  15. 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,...

  16. 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,...

  17. 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,...

  18. 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,...

  19. 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,...

  20. 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,...

  1. 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,...

  2. 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,...

  3. 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,...

  4. 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,...

  5. 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,...

  6. 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,...

  7. 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,...

  8. 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,...

  9. 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,...

  10. 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,...

  11. 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,...

  12. 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,...

  13. 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,...

  14. 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

  15. Dissipation of the reactor heat at the Savannah River Plant

    International Nuclear Information System (INIS)

    The effluent cooling water from the heat exchangers of the Savannah River nuclear reactors is cooled by natural processes as it flows through the stream beds, canals, ponds, and swamps on the plant site. The Langhaar equation, which gives the rate of heat removal from the water surface as a function of the surface temperature, air temperature, relative humidity, and wind speed, is applied satisfactorily to calculate the cooling that occurs at all temperature levels and for all modes of water flow. The application of this equation requires an accounting of effects such as solar heating, shading, mixing, staging, stratification, underflow, rainfall, the imposed heat load, and the rate of change in heat content of the body of water

  16. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-02-15

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

  17. Performance tests for integral reactor nuclear fuel

    International Nuclear Information System (INIS)

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

  18. Developing the MAPLE materials test reactor concept

    International Nuclear Information System (INIS)

    MAPLE-MTR is a new multipurpose research facility being planned by AECL Research as a possible replacement for the 35-year-old NRU reactor. In developing the MAPLE-MTR concept, AECL is starting from the recent design and licensing experience with the MAPLE-X10 reactor. By starting from technology developed to support the MAPLE-X10 design and adapting it to produce a concept that satisfies the requirements of fuel channel materials testing and fuel irradiation programs, AECL expects to minimize the need for major advances in nuclear technology (e.g., fuel, heat transfer). Formulation of the MAPLE-MTR concept is at an early stage. This report describes the irradiation requirements of the research areas, how these needs are translated into design criteria for the project and elements of the preliminary design concept

  19. 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.

  20. 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

  1. 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)

  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

    2011-01-01

    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

  3. 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

  4. Neutron spectrometry and dosimetry study at two research nuclear reactors using bonner sphere spectrometer (BSS), rotational spectrometer (ROSPEC) and cylindrical nested neutron spectrometer (NNS)

    International Nuclear Information System (INIS)

    Neutron spectrometry and subsequent dosimetry measurements were undertaken at the McMaster Nuclear Reactor (MNR) and AECL Chalk River National Research Universal (NRU) Reactor. The instruments used were a Bonner sphere spectrometer (BSS), a cylindrical nested neutron spectrometer (NNS) and a commercially available rotational proton recoil spectrometer. The purposes of these measurements were to: (1) compare the results obtained by three different neutron measuring instruments and (2) quantify neutron fields of interest. The results showed vastly different neutron spectral shapes for the two different reactors. This is not surprising, considering the type of the reactors and the locations where the measurements were performed. MNR is a heavily shielded light water moderated reactor, while NRU is a heavy water moderated reactor. The measurements at MNR were taken at the base of the reactor pool, where a large amount of water and concrete shielding is present, while measurements at NRU were taken at the top of the reactor (TOR) plate, where there is only heavy water and steel between the reactor core and the measuring instrument. As a result, a large component of the thermal neutron fluence was measured at MNR, while a negligible amount of thermal neutrons was measured at NRU. The neutron ambient dose rates at NRU TOR were measured to be between 0.03 and 0.06 mSv h-1, while at MNR, these values were between 0.07 and 2.8 mSv h-1 inside the beam port and -1 between two operating beam ports. The conservative uncertainty of these values is 15 %. The conservative uncertainty of the measured integral neutron fluence is 5 %. It was also found that BSS over-responded slightly due to a non-calibrated response matrix. (authors)

  5. 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

  6. 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.

  7. 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

  8. 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 power 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.)

  9. Simulation of a steam bubble transport in the primary system of the pool type lead cooled fast reactors

    International Nuclear Information System (INIS)

    Pool-type design makes Lead cooled Fast Reactor (LFR) economically competitive with other advanced reactor designs considered under the Generation IV framework. However, close proximity of steam generator to the core increases the risks associated with Steam Generator Tube Rupture/Leakage (SGTR/SGTL) such as voiding of the core and resulting reactivity insertion and/or local damage (burnout) of fuel rod cladding. Analysis of consequences of SGTL provided in present paper suggests that small bubbles of steam can be dragged by the turbulent coolant flow into the core region. Trajectories of the bubbles are determined by location of the leak, bubbles size and turbulent flow field of lead coolant. The influence of epistemic uncertainty in drag coefficient on prediction of the fraction of bubbles that can reach the core and accumulate in the primary coolant system is discussed in the paper. (author)

  10. The life-extension and upgrade program of the Tsing Hua Open-pool Reactor (THOR) and its research prospectives

    International Nuclear Information System (INIS)

    The Tsing Hua Open-Pool Reactor (THOR) has been operated for thirty years. It is the regulations of the ROCAEC that any reactor shall be decommissioned after forty-year operation since the first fuel loading. Therefore, for extending the lifetime of THOR, it is necessary to have a life-extension program to be approved by the ROCAEC and also completed by the year of 1997. At the same time, for proceeding new research purposes, it is planed to upgrade the thermal power of THOR from 1 Wth up to 3 Wth and hopefully to reach the maximum thermal neutron flux of 5x1013 n/cm2.s and the fast flux close to that order. New research directions involve (a) boron-captured neutron cancer therapy (BNCT) (b) small-angle neutron scattering (SANS). (author)

  11. 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.)

  12. Calculations of partial LOCA in a swimming-pool-reactor with MTR-elements and planned mock-up experiment

    International Nuclear Information System (INIS)

    A partial uncovering of the MTR fuel plates of the swimming pool reactor SAPHIR located at the Swiss Federal Reactor Research Institute (E.I.R.) could be caused by a loss of coolant accident due to a beam tube break. The transient temperature excursions of the fuel plates during the LOCA have been predicted with computer simulations. Because a reliable prediction of the flow regime and hence the heat transfer in the uncovered part of the plate is not possible with current knowledge, a parametric study employing different heat transfer models is presented in this paper. The results show, that the heat transfer model in the uncovered part of the fuel plate has an important influence on the predicted temperatures. A mock-up experimental facility, which will supply data for the heat transfer occuring in the uncovered part, will also be described at the end of the paper. (author)

  13. Off reactor testings. Technological engineering applicative research

    International Nuclear Information System (INIS)

    By the end of year 2000 over 400 nuclear electro-power units were operating world wide, summing up a 350,000 MW total capacity, with a total production of 2,300 TWh, representing 16% of the world's electricity production. Other 36 units, totalizing 28,000 MW, were in construction, while a manifest orientation towards nuclear power development was observed in principal Asian countries like China, India, Japan and Korea. In the same world's trend one find also Romania, the Cernavoda NPP Unit 1 generating electrical energy into the national system beginning with 2 December 1996. Recently, the commercial contract was completed for finishing the Cernavoda NPP Unit 2 and launching it into operation by the end of year 2004. An important role in developing the activity of research and technological engineering, as technical support for manufacturing the CANDU type nuclear fuel and supplying with equipment the Cernavoda units, was played by the Division 7 TAR of the INR Pitesti. Qualification testings were conducted for: - off-reactor CANDU type nuclear fuel; - FARE tools, pressure regulators, explosion proof panels; channel shutting, as well as functional testing for spare pushing facility as a first step in the frame of the qualification tests for the charging/discharging machine (MID) 4 and 5 endings. Testing facilities are described, as well as high pressure hot/cool loops, measuring chains, all of them fulfilling the requirements of quality assurance. The nuclear fuel off-reactor tests were carried out to determine: strength; endurance; impact, pressure fall and wear resistance. For Cernavoda NPP equipment testings were carried out for: the explosion proof panels, pressure regulators, behaviour to vibration and wear of the steam generation tubings, effects of vibration upon different electronic component, channel shutting (for Cernavoda Unit 2), MID operating at 300 and 500 cycles. A number of R and D programs were conducted in the frame of division 7 TAR of INR

  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. 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

    ... 10 Energy 3 2010-01-01 2010-01-01 false Uniform Test Method for Measuring the Energy Consumption... Subpart B of Part 430—Uniform Test Method for Measuring the Energy Consumption of Pool Heaters 1. Test... 2.9 of ANSI Z21.56-1994. The measurement of energy consumption for oil-fired pool heaters in Btu...

  16. 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

  17. Safety aspects of the cleaning and conditioning of radioactive sludge from spent fuel storage pool on 'RA' Research reactor in the Vinca Institute

    International Nuclear Information System (INIS)

    Spent fuel elements from nuclear reactors in the Vinca Institute have been temporary stored in water filled storage pool. Due to the fact that the water in the spent fuel elements storage pool have not been purified for a long time, all metallic components submerged in the water have been hardly corroded and significant amount of the sludge has been settled on the bottom of the pool. As a first step in improving spent fuel elements storage conditions and slowing down corrosion in the storage spent fuel elements pool we have decided to remove the sludge from the bottom of the pool. Although not high, but slightly radioactive, this sludge had to be treated as radioactive waste material. Some safety aspects and radiation protection measures in the process of the spent fuel storage pool cleaning are presented in this paper

  18. Sloshing and fluid-structure interaction in a 400-MWe pool-type advanced fast reactor

    International Nuclear Information System (INIS)

    This paper describes the seismic analysis of a 400-MWe advanced fast reactor under 0.3 g SSE ground excitation. Two types of analyses are performed - the sloshing analysis and the fluid-structure interaction analysis. In the sloshing analysis, the sloshing frequency and wave patterns are calculated. The maximum wave height and the sloshing forces exerted on the submerged components and the primary tank are evaluated. In the fluid-structure interaction analysis, the maximum horizontal acceleration for the reactor core and the relative displacement between the reactor core and UIS are examined. The fluid-coupling phenomena between various components are investigated. Seismic stresses at critical areas are examined. The results obtained from this study are very useful to the design of the advanced reactors. Meanwhile, the computer code FLUSTR-ANL has proved to be a useful analytical tool for assessing the complicated seismic fluid-structure interactions and sloshing in the fast reactor systems. 10 refs., 25 figs

  19. 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.

  20. Failure Status Design of the Air Operated Valves and Solenoid Valves of Hot Water Layer System in the Open-pool Type Research Reactor

    International Nuclear Information System (INIS)

    Since the HWLS has ion exchangers, ionized radionuclides in the hot water layer are also purified. Thus the dose at the pool top should be maintained as low as reasonably achievable when the reactor is in normal operation. The HWLS consists of pumps, ion exchangers, heaters, flow meter orifices, all the necessary inter-connecting pipes, valves, and instruments, located in the HWLS equipment room as shown in Fig. 1. Each component, such as the pump, ion exchanger, strainer, and heater has 100% capacity to ensure that failure of one component does not result in the functional failure of the whole system. The suction line is split to the inlet of each pump to take the pool water to the ion exchangers. The design flow rate passes through the ion exchanger to remove the radioactive ions and impurities, and then go to the heater. The coolant is heated up to the desired temperature and flows back to the upper part of the reactor pool. Demineralized water is also supplied to the reactor pool by the HWLS when the pool water level drops to low level by an evaporation loss in order to maintain the normal pool water level. Operability of the HWLS will be maintained when all valves are fully opened. And, opening status of the valve in the demi-water make-up line has no impact the reactor operation. There are results of the failure status design of the air operated valves and solenoid valves

  1. Work plan for testing silicone impression material and fixture on pool cell capsule

    International Nuclear Information System (INIS)

    The purpose of this work plan is to provide a safe procedure to test a cesium capsule impression fixture at Waste Encapsulation and Storage Facility (WESF). The impression will be taken with silicone dental impression material pressed down upon the capsule using the impression fixture. This test will evaluate the performance of the fixture and impression material under high radiation and temperature conditions on a capsule in a WESF pool cell

  2. The parallex project: CANDU MOX fuel testing with weapons-derived plutonium

    International Nuclear Information System (INIS)

    The Parallex Project consists of a parallel experiment in which weapons-derived plutonium (WPu) from the United States and from the Russian Federation will be tested as mixed-oxide (MOX) CANDU fuel in the National Research Universal (NRU) reactor at the Chalk River Laboratories in Canada. Plutonium derived from excess weapons will be fabricated into CANDU MOX fuel at the A.A. Bochvar Institute in Moscow and at the Los Alamos National Laboratory in the United States. The MOX fuel will be transported to CRL, where it will be characterized, assembled into fuel bundles and then irradiated in the NRU reactor. Following irradiation, the fuel will be examined in hot cells to assess its irradiation performance. This paper describes the scope, rationale and current status of the Parallex Project. (author)

  3. SRS reactor stack plume marking tests

    International Nuclear Information System (INIS)

    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

  4. Liquid sloshing in gravity driven water pool of Advanced Heavy Water Reactor - pool liquid under design seismic load and slosh control studies

    International Nuclear Information System (INIS)

    Sloshing phenomenon is well understood in regular cylindrical and rectangular liquid tanks subjected to earthquake. However, seismic behaviour of water in complex geometry such as a sectored annular tank, e.g., Gravity Driven Water Pool (GDWP) which is located in Advanced Heavy Water Reactor (AHWR) need to be investigated in detail in the view of safety significance. Initially, for validation of Computational Fluid Dynamics (CFD) procedure, square and four sectored square tanks are taken. Slosh height and liquid pressure are calculated over time through theoretical and experimental procedures. Results from theoretical and experimental approaches are compared with CFD results and found to be in agreement. The present work has two main objectives. The first one is to investigate the sloshing behaviour in an un-baffled and baffled three dimensional single sector of GDWP of AHWR under sinusoidal excitation. Other one is to study the sloshing in GDWP water using simulated seismic load along the three orthogonal directions. This simulated seismic load is generated from design basis floor response spectrum data (FRS) of AHWR building. For this, the annular tank is modelled along with water and numerical simulation is carried out. The sinusoidal and earthquake excitations are applied as acceleration force along with gravity. For the earthquake case, acceleration-time history is generated compatible to the design FRS of AHWR building. The free surface is captured by Volume of Fluid (VOF) technique and the fluid domain is solved by finite volume method while the structural domain is solved by finite element approach. Un-baffled and baffled tank configurations are compared to show the reduction in wave height under excitation. The interaction between the fluid and pool wall deformation is simulated using a partitioned fluid-structure coupling. In the earthquake case, a user subroutine function is developed to convert FRS in to time history of acceleration in three directions

  5. 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.

  6. 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)

  7. 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.

  8. 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

  9. 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

  10. 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.

  11. 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

  12. Self Compacting Concrete with Chalk Filler

    DEFF Research Database (Denmark)

    Sørensen, Eigil V.

    2007-01-01

    Utilisation of Danish chalk filler has been investigated as a means to produce self compacting concrete (SCC) at lower strength levels for service in non aggressive environments. Stable SCC mixtures were prepared at chalk filler contents up to 60% by volume of binder to yield compressive strengths...

  13. 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.

  14. 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.

  15. 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...

  16. 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...

  17. 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...

  18. 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...

  19. 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...

  20. 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...