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Sample records for chalk river pool test reactor

  1. 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

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

    Science.gov (United States)

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

    2016-01-01

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

  3. 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

  4. 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.

  5. Assessing sediment toxicity from navigational pools of the Upper Mississippi River using a 28-day Hyalella azteca test

    Science.gov (United States)

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

    1998-01-01

    To assess the extent of sediment contamination in the Upper Mississippi River (UMR) system after the flood of 1993, sediment samples were collected from 24 of the 26 navigational pools in the river and from one site in the Saint Croix River in the summer of 1994. Whole-sediment tests were conducted with the amphipod Hyalella azteca for 28 days measuring the effects on survival, growth, and sexual maturation. Amphipod survival was significantly reduced in only one sediment (13B) relative to the control and reference sediments. Body length of amphipods was significantly reduced relative to the control and reference sediments in only one sample (26C). Sexual maturation was not significantly reduced in any treatment when compared to the control and reference sediments. No significant correlations were observed between survival, growth, and maturation to either the physical or chemical characteristics of the sediment samples from the river. When highly reliable effect range medians (ERMs) were used to evaluate sediment chemistry, 47 of 49 (96%) of the samples were correctly classified as nontoxic. These results indicate that sediment samples from the Upper Mississippi River are relatively uncontaminated compared to other areas of known contamination in the United States.

  6. 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

  7. 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

  8. 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.

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

    Science.gov (United States)

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

    2005-01-01

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

  10. 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.

  11. Convective cooling in a pool-type research reactor

    Science.gov (United States)

    Sipaun, Susan; Usman, Shoaib

    2016-01-01

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

  12. 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.

  13. Hillerslev outcrop chalk

    Energy Technology Data Exchange (ETDEWEB)

    Lykke, M.M.

    2003-08-01

    Fractures are a great benefit to production of oil, since the matrix permeability in the oil bearing chalk reservoirs in the North Sea is low. Many of the oil fields would be marginally economic to produce without natural or induced fractures to enhance the effective permeability of the reservoirs. However, when oil is produced by use of waterflooding, an important issue is whether water fingering (fracture flow) will occur. Water fingering is due to faster flow of water in the fractures than in the matrix during waterflooding. Capillary suction of water (spontaneous or forced) must exist for waterflooding to be economic. If the matrix sucks water from the fractures, waterflooding can be a very efficient mechanism. If not, the waterflooding may fail, since the water will travel directly from the injector to the producer through the fractures, i.e. the result would be recycled water. In my Ph.D., two-phase fracture flow is investigated. The investigation is based on waterflooding tests on fractured outcrop Hillerslev chalk specimens. It is chosen to use Hillerslev outcrop chalk due to that this chalk is highly fractured and that it can be regarded as a close analogue to the oil producing Tor formation of the Valhall field located in the North Sea. To investigate fracture flow, it is important to obtain knowledge of the fractures in the chalk, i.e. it is necessary to perform a fracture study of the chalk. A field trip was made to the Hillerslev outcrop chalk quarry located in the northern part of Jutland. Here, a (global) fracture description was carried out and twelve chalk block samples were recovered at a chosen location in the Hillerslev quarry. For comparison of earlier work performed in the Hillerslev chalk quarry, this report contains a summary of the fracture description and sampling carried out during EFP-98, EFP-96 and earlier work. Measured values of porosity, permeability and capillary pressure curves of Hillerslev outcrop chalk are included to obtain

  14. Vernal Pool Study 2005 Wallkill River National Wildlife Refuge

    Data.gov (United States)

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

  15. Decommissioning of the pool reactor Thetis in Ghent, Belgium

    Energy Technology Data Exchange (ETDEWEB)

    Cortenbosch, Geert; Mommaert, Chantal [Bel V, Brussels (Belgium); Tierens, Hubert; Monsieurs, Myriam; Meierlaen, Isabelle; Strijckmans, Karel [Ghent Univ. (Belgium)

    2016-11-15

    The Thetis research pool reactor (with a nominal power of 150 kW) of the Ghent University was operational from 1967 till December 2003. The first phase of the decommissioning of the reactor, the removal of the spent fuel from the site, took place in 2010. The cumulative dose received was only 404 man . μSv. During the second phase, the transition period between the removal of the spent fuel in 2010 and the start of the decommissioning phase in March 2013, 3-monthly internal inspections and inspections by Bel V, were performed. The third and final decommissioning phase started on March 18, 2013. The total dose received between March 2013 and August 2013 was 1561 man . μSv. The declassification from a Class I installation to a Class II installation was possible by the end of 2015. The activated concrete in the reactor pool will remain under regulatory control until the activation levels are lower than the limits for free release.

  16. 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.

  17. 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.

  18. Savannah River Site reactor safety assessment. Draft

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-02-28

    This report gives the results of a Savannah River Site (SRS) Production Reactor risk assessment. 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 timely information to the US 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 Site programs in Heavy Water Reactor safety.

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

    Science.gov (United States)

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

    2017-02-01

    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.

  20. Reactor Simulator Testing

    Science.gov (United States)

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

    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. Keywords: fission, space power, nuclear, liquid metal, NaK.

  1. 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 2010...

  2. 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 2010...

  3. 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 2010...

  4. 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...

  5. 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...

  6. 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...

  7. 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...

  8. 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...

  9. Preliminary Requirement of Hot Pool Free Surface Level from PGSFR Reactor Head

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Gyeonghoi; Joo, Hyeongkook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The sensitivity study on structural integrity evaluations are carried out to make a decision of a hot pool free surface location from the reactor head for a preliminary designed reactor enclosure system. To do this, the thermal stress evaluations for a reactor vessel are carried out for a steady state normal operating condition with detailed heat transfer analyses through the reactor enclosure system. From these results, the preliminary design requirement of a hot pool free surface location from the reactor head is established to be 2.0m. From the sensitivity studies on the structural integrity evaluations for a steady state condition, the preliminary distance from the hot pool free surface to the reactor head is determined to be 2.0m same as a conceptual design. More detailed structural analyses for a reactor enclosure system will be carried out as a PGSFR structural design goes forward in detail.

  10. 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.

  11. Detectability prediction for a thermoacoustic sensor in the breazeale nuclear reactor pool

    Energy Technology Data Exchange (ETDEWEB)

    Smith, James [Idaho National Laboratory, Idaho Falls, ID (United States); Hrisko, Joshua [Idaho National Laboratory, Idaho Falls, ID (United States); Garrett, Steven [Idaho National Laboratory, Idaho Falls, ID (United States)

    2016-03-01

    Laboratory experiments have suggested that thermoacoustic engines can be in- corporated within nuclear fuel rods. Such engines would radiate sounds that could be used to measure and acoustically-telemeter information about the op- eration of the nuclear reactor (e.g., coolant temperature or uxes of neutrons or other energetic particles) or the physical condition of the nuclear fuel itself (e.g., changes in temperature, evolved gases) that are encoded as the frequency and/or amplitude of the radiated sound [IEEE Measurement and Instrumen- tation 16(3), 18-25 (2013)]. For such acoustic information to be detectable, it is important to characterize the vibroacoustical environments within reactors. Measurements will be presented of the background noise spectra (with and with- out coolant pumps) and reverberation times within the 70,000 gallon pool that cools and shields the fuel in the 1 MW research reactor on Penn State's campus using two hydrophones, a piezoelectric projector, and an accelerometer. Sev- eral signal-processing techniques will be demonstrated to enhance the measured results. Background vibrational measurement were also taken at the 250 MW Advanced Test Reactor, located at the Idaho National Laboratory, using ac- celerometers mounted outside the reactor's pressure vessel and on plumbing will also be presented. The detectability predictions made in the thesis were validated in September 2015 using a nuclear ssion-heated thermoacoustic sensor that was placed in the core of the Breazeale Nuclear Reactor on Penn State's campus. Some features of the thermoacoustic device used in that experiment will also be revealed. [Work supported by the U.S. Department of Energy.

  12. 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.

  13. FASTER Test Reactor Preconceptual Design Report

    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. J. [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, S. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-03-31

    The FASTER test 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

  15. 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 are......: 1) Pure carbonate chalk with mudstone texture from Stevns Klint (Denmark), 2) Relatively pure chalk sediments with varying content of microfossils from the Ontong Java Plateau (Western Pacific), 3) Clay-rich chalk and mixed sediments from the Caribbean. The tested samples were characterised...... of microfossils and fine-grained silica and clay. Samples with relatively pure chalk mud supported texture compact along a common stress - matrix porosity trend. Microfossils thus have a passive role, apparently because they are supported by the chalk mud. Samples with fine-grained silica and clay can be modelled...

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

    Science.gov (United States)

    Robertson, J.B.

    1974-01-01

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

  17. FASTER test reactor preconceptual design report summary

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-29

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

  18. Radiological performance of hot water layer system in open pool type reactor

    Directory of Open Access Journals (Sweden)

    Amr Abdelhady

    2013-06-01

    Full Text Available The paper presents the calculated dose rate carried out by using MicroShield code to show the importance of hot water layer system (HWL in 22 MW open pool type reactor from the radiation protection safety point of view. The paper presents the dose rate profiles over the pool surface in normal and abnormal operations of HWL system. The results show that, in case of losing the hot water layer effect, the radiation dose rate profiles over the pool surface will increase from values lower than the worker permissible dose limits to values very higher than the permissible dose limits.

  19. The noncondensable gas effects on loss-of-coolant accident steam condensation loads in boiling water reactor pressure suppression pool

    Energy Technology Data Exchange (ETDEWEB)

    Kukita, Y.; Namatame, K.; Shiba, M.; Takeshita, I.

    1983-11-01

    The noncondensable gas effects on the loss-ofcoolant-accident-induced steam condensation loads in the boiling water reactor pressure suppression pool have been investigated with regard to experimental data obtained from a large-scale multivent test program. Previous studies have noted that the presence of the noncondensable gas (air), which initially fills the containment drywell space, stabilizes the direct-contact condensation in the pressure suppression pool and hampers onset of the chugging phenomenon, which induces most significant steam condensation load onto the pool boundary. This was found to be true for the tests with relatively small-break diameters, where the maximum steam mass fluxes in the vent pipe were lower than the upper threshold value for the onset of chugging. However, in the tests with the maximum vent steam mass fluxes moderately higher than the chugging upper threshold value, early depletion of the noncondensable gas tended to result in significant stabilization of steam condensation accompanied by an excursion of temperature of pool water surrounding the vent pipe outlets, which led to a delayed onset of chugging. Due to this combined influence of the noncondensable gas and nonuniform pool temperature, and due to dependence of magnitude of chugging load on the vent steam mass flux, the peak magnitude of the steam condensation load appearing in a blowdown can be very sensitive to the initial and break conditions.

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

    Energy Technology Data Exchange (ETDEWEB)

    Rod, S R

    1991-08-01

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

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Bum; Hahn Do Hee

    2003-01-01

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

  2. 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.

  3. 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

  4. 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.

  5. Porosity variation in chalk

    DEFF Research Database (Denmark)

    Lind, Ida; Grøn, Peter

    1996-01-01

    Vertical porosity variations in chalk are generally assumed to result from either a vaguely defined combination of primary sedimentary and diagenetic processes or solely to diagenetic processes. In this study, image analysis of backscatter electron images of polished samples and geochemical...... microprobe mapping were applied to measure the porosity variation in a limited number of chalk samples. Microscope data indicate that in all cases the chalk has been subjected to diagenetic processes, but our data suggest that the variations in porosity originate in primary sedimentary differences....

  6. 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.

  7. 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)

  8. 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.

  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. Environmental Impact Study of the Northern Section of the Upper Mississippi River. Pool 9.

    Science.gov (United States)

    1973-11-01

    basedl on !u o, 1973 survo- vrio 145 served on the river per hour of oav use is al o g reate. t in tht uper two- thirds of Pool 9. The numbers of people...Upper Mississippi River Comprehensive Basin Study. 1970. Appendix J. page 90. U.S. Public Health Service U.S. Department of Health, Education , and

  11. Full-Length High-Temperature Severe Fuel Damage Test No. 5: Final safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lanning, D.D.; Lombardo, N.J.; Panisko, F.E.

    1993-09-01

    This report presents the final safety analysis for the preparation, conduct, and post-test discharge operation for the Full-Length High Temperature Experiment-5 (FLHT-5) to be conducted in the L-24 position of the National Research Universal (NRU) Reactor at Chalk River Nuclear Laboratories (CRNL), Ontario, Canada. The test is sponsored by an international group organized by the US Nuclear Regulatory Commission. The test is designed and conducted by staff from Pacific Northwest Laboratory with CRNL staff support. The test will study the consequences of loss-of-coolant and the progression of severe fuel damage.

  12. Chalk as a reservoir

    DEFF Research Database (Denmark)

    Fabricius, Ida Lykke

    reduces porosity at the same time as it increases specific surface and thus cause permeability to be low. In the Central North Sea the silica is quartzitic. Silica rich chalk intervals are typically found in the Ekofisk and Hod formations. In addition to silica, Upper Cretaceous and Palæogene chalks...... stabilizes chemically by recrystallization. This process requires energy and is promoted by temperature. This recrystallization in principle does not influence porosity, but only specific surface, which decreases during recrystallization, causing permeability to increase. The central North Sea is a warm...... have hardly any stylolites and can have porosity above 40% or even 50% and thus also have relatively high permeability. Such intervals have the problem though, that increasing effective stress caused by hydrocarbon production results in mechanical compaction and overall subsidence. Most other chalk...

  13. Two black carbon pools transported by the Changjiang and Huanghe Rivers in China

    Science.gov (United States)

    Wang, Xuchen; Xu, Caili; Druffel, Ellen M.; Xue, Yuejun; Qi, Yuanzhi

    2016-12-01

    Major rivers play important roles in transporting large amounts of terrestrial organic matter from land to the ocean each year, and the organic matter carried by rivers contains a significant fraction of black carbon (BC). A recent study estimated that 0.027 Gt of BC is transported in the dissolved phase by rivers each year, which accounts for 10% of the global flux of dissolved organic carbon. The relative sources of this large amount of riverine dissolved black carbon (DBC) from biomass burning (young, modern 14C) and fossil fuel (old, 14C free) combustion are not known. We present radiocarbon measurements of BC in both dissolved and particulate phases transported by the Changjiang and Huanghe Rivers, the two largest rivers in China, during 2015. We show that two, distinct BC pools (young and old) were carried by the rivers. The DBC pool was much younger than the particulate BC (PBC) pool. Mass balance calculations indicate that most (78-85%) of the DBC in the Changjiang and Huanghe Rivers was derived from biomass burning, and only 15-22% was from fossil fuel combustion. In contrast, PBC from biomass burning and fossil fuel combustion were approximately equal in these two rivers. Export of PBC and DBC by the rivers are decoupled, and fluxes of PBC were 4.1 and 6.7 times higher than DBC in the Changjiang and Huanghe Rivers, respectively. The 14C age differences of the two BC pools suggest that BC derived from biomass burning and fossil fuel combustion are mobilized in different phases and on different time scales in these rivers.

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

    Energy Technology Data Exchange (ETDEWEB)

    Pivet, Sylvestre [CEA, Centre de Cadarache, F-13108 Saint Paul lez Durance (France); Minguet, Jean-Luc [AREVA-Technicatome, BP17, 91192 Gif-sur-Yvette (France)

    2006-07-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

  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. 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...

  18. 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...

  19. 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...

  20. 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...

  1. 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...

  2. 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...

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

    Data.gov (United States)

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

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

  5. 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...

  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 8

    Data.gov (United States)

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

  8. 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...

  9. 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...

  10. 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...

  11. 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...

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

    Data.gov (United States)

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

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

  14. 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...

  15. 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...

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

    Data.gov (United States)

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

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

  18. 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...

  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 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...

  1. 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...

  2. 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 track...

  3. 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 track...

  4. 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 track...

  5. 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 track...

  6. Review of fuel assembly and pool thermal hydraulics for fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Roelofs, Ferry, E-mail: roelofs@nrg.eu; Gopala, Vinay R.; Jayaraju, Santhosh; Shams, Afaque; Komen, Ed

    2013-12-15

    Highlights: • Literature review of fuel assembly and pool thermal hydraulics for fast reactors. • Experiments and state-of-the-art simulations. • For wire wrapped fuel assemblies RANS for complete fuel assembly is state-of-the-art, LES serves reference. • For pool thermal hydraulics, typically 5 to 20 million computational volumes are used in RANS simulations. • Gas entrainment analyses are extremely demanding as in addition they request multiphase modelling. -- Abstract: Liquid metal cooled reactors are envisaged to play an important role in the future of nuclear energy production because of their possible efficient use of uranium and the possibility to reduce the volume and lifetime of nuclear waste. Thermal-hydraulics is recognized as a key scientific subject in the development of such reactors. Two important challenges for the design of liquid metal fast reactors (LMFRs) are fuel assembly and pool thermal hydraulics. The heart of every nuclear reactor is the core, where the nuclear chain reaction takes place. Heat is produced in the nuclear fuel and transported to the coolant. LMFR core designs consist of many fuel assemblies which in turn consist of a large number of fuel rods. Wire wraps are commonly envisaged as spacer design in LMFR fuel assemblies. For the design and safety analyses of such reactors, simulations of the heat transport within the core are essential. The flow exiting the core is made up of the outlets of many different fuel assemblies. The liquid metal in these assemblies may be heated up to different temperatures. This leads to temperature fluctuations on various above core structures. As these temperature fluctuations may lead to thermal fatigue damage of the structures, an accurate characterization of the liquid metal flow field in the above core region is very important. This paper will provide an overview of state-of-the-art evaluations of fuel assembly and pool thermal hydraulics for LMFRs. It will show the tight interaction

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

    Science.gov (United States)

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

    1998-01-01

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

  8. 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.

  9. 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

  10. Geomorphic Mapping Pool 7 - Upper Mississippi River Basin

    Science.gov (United States)

    1987-10-05

    in the conversion of raw organic material to humus , the mixing of organic and inorganic material, the creation of channels, and the vertical...Mississippi River, Quaternary Research 20: 165-176. GALLAGHER, JAMES P., ROLAND RODELL, and KATHERINE STEVENSON , 1982, The 1980-1982 LaCrosse Area

  11. SPARC fast reactor design : Design of two passively safe metal-fuelled sodium-cooled pool-type small modular fast reactors with Autonomous Reactivity Control

    OpenAIRE

    Lindström, Tobias

    2015-01-01

    In this master thesis a small modular sodium-cooled metal-fuelled pool-type fast reactor design, called SPARC - Safe and Passive with Autonomous Reactivity control, has been designed. The long term reactivity changes in the SPARC are managed by implementation of the the Autonomous Reactivity Control (ARC) system, which is the novelty of the design. The overall design is mainly based on the Integral Fast Reactor project (IFR), which experimentally demonstrated the passive safety characteristic...

  12. 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.

  13. 33 CFR 207.320 - Mississippi River, Twin City Locks and Dam, St. Paul and Minneapolis, Minn.; pool level.

    Science.gov (United States)

    2010-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ruedig, E. [Colorado State University (United States); Higley, K. [Oregon State University (United States)

    2014-07-01

    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

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

    Science.gov (United States)

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

    2006-01-01

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

  16. 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.

  17. 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.

  18. Evaluation of light penetration on Navigation Pools 8 and 13 of the Upper Mississippi River

    Science.gov (United States)

    Giblin, Shawn; Hoff, Kraig; Fischer, Jim; Dukerschein, Terry

    2010-01-01

    The availability of light can have a dramatic affect on macrophyte and phytoplankton abundance in virtually all aquatic ecosystems. The Long Term Resource Monitoring Program and other monitoring programs often measure factors that affect light extinction (nonvolatile suspended solids, volatile suspended solids, and chlorophyll) and correlates of light extinction (turbidity and Secchi depth), but rarely do they directly measure light extinction. Data on light extinction, Secchi depth, transparency tube, turbidity, total suspended solids, and volatile suspended solids were collected during summer 2003 on Pools 8 and 13 of the Upper Mississippi River. Regressions were developed to predict light extinction based upon Secchi depth, transparency tube, turbidity, and total suspended solids. Transparency tube, Secchi depth, and turbidity all showed strong relations with light extinction and can effectively predict light extinction. Total suspended solids did not show as strong a relation to light extinction. Volatile suspended solids had a greater affect on light extinction than nonvolatile suspended solids. The data were compared to recommended criteria established for light extinction, Secchi depth, total suspended solids, and turbidity by the Upper Mississippi River Conservation Committee to sustain submersed aquatic vegetation in the Upper Mississippi River. During the study period, the average condition in Pool 8 met or exceeded all of the criteria whereas the average condition in Pool 13 failed to meet any of the criteria. This report provides river managers with an effective tool to predict light extinction based upon readily available data.

  19. To What Degree Thermal Cycles Affect Chalk Strength

    DEFF Research Database (Denmark)

    Livada, Tijana; Nermoen, Anders; Korsnes, Reidar Inger;

    Chalk reservoirs could potentially undergo destabilization as the result of repeated cold water injection into a hot reservoir during water flooding. Preliminary results of an ongoing study are presented in this paper, which compare the impact of temperature cycling on mechanical behavior on dry...... triaxial cell experiments. For dry rock, no significant effects of temperature cycling was found on average tensile strength, however the range of the tensile failure stress is doubled for the samples exposed to 50 temperature cycles, as opposed to those to none. For water saturated cores, the temperature...... and water saturated chalk. Sixty disks of dry Kansas chalk exposed to different number of temperature cycles were tested for tensile strength using a Brazilian test. Changes in elastic properties as function of number of temperature cycles of the same chalk, but now saturated in water, were studied using...

  20. 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)

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

    Science.gov (United States)

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

    2006-01-01

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

  2. Environmental Impact Study of the Northern Section of the Upper Mississippi River. Pool 7.

    Science.gov (United States)

    1973-11-01

    are for similar community situations in Western Wisconsin (Curtis, 1971). Algae Members of all classes and virtually all common orders of algae are...checklist of algae collected from Navigation Pool No. 7, Upper Mississippi River, 1970-1971. CIILOROPIIYTA Class Chlorophyceae Order Volvocales Family...Marssoniella elegans Miroysti sp.uc Mecrcsteia spau. Order Hormogonales Family Oscil1la Loriaceae Oscillatoria sp. Spirulina laxa Family Nostocaceae

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

    Science.gov (United States)

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

    2015-11-01

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

  4. 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

  5. Post shut-down decay heat removal from nuclear reactor core by natural convection loops in sodium pool

    Energy Technology Data Exchange (ETDEWEB)

    Rajamani, A. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Sundararajan, T., E-mail: tsundar@iitm.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Prasad, B.V.S.S.S. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Parthasarathy, U.; Velusamy, K. [Nuclear Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

    2016-05-15

    Highlights: • Transient simulations are performed for a worst case scenario of station black-out. • Inter-wrapper flow between various sub-assemblies reduces peak core temperature. • Various natural convection paths limits fuel clad temperatures below critical level. - Abstract: The 500 MWe Indian pool type Prototype Fast Breeder Reactor (PFBR) has a passive core cooling system, known as the Safety Grade Decay Heat Removal System (SGDHRS) which aids to remove decay heat after shut down phase. Immediately after reactor shut down the fission products in the core continue to generate heat due to beta decay which exponentially decreases with time. In the event of a complete station blackout, the coolant pump system may not be available and the safety grade decay heat removal system transports the decay heat from the core and dissipates it safely to the atmosphere. Apart from SGDHRS, various natural convection loops in the sodium pool carry the heat away from the core and deposit it temporarily in the sodium pool. The buoyancy driven flow through the small inter-wrapper gaps (known as inter-wrapper flow) between fuel subassemblies plays an important role in carrying the decay heat from the sub-assemblies to the hot sodium pool, immediately after reactor shut down. This paper presents the transient prediction of flow and temperature evolution in the reactor subassemblies and the sodium pool, coupled with the safety grade decay heat removal system. It is shown that with a properly sized decay heat exchanger based on liquid sodium and air chimney stacks, the post shutdown decay heat can be safely dissipated to atmospheric air passively.

  6. 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.

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

    Science.gov (United States)

    Huang, Chun-Ping; Lin, Tzung-Yi; Chiao, Ling-Huan; Chen, Hong-Bin

    2012-09-30

    There were approximately 926 m(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 (137)Cs, (90)Sr, 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 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.

  8. 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 the...

  9. Optimal item pool design for computerized adaptive tests with polytomous items using GPCM

    Directory of Open Access Journals (Sweden)

    Xuechun Zhou

    2014-09-01

    Full Text Available Computerized adaptive testing (CAT is a testing procedure with advantages in improving measurement precision and increasing test efficiency. An item pool with optimal characteristics is the foundation for a CAT program to achieve those desirable psychometric features. This study proposed a method to design an optimal item pool for tests with polytomous items using the generalized partial credit model (G-PCM. It extended a method for approximating optimality with polytomous items being described succinctly for the purpose of pool design. Optimal item pools were generated using CAT simulations with and without practical constraints of content balancing and item exposure control. The performances of the item pools were evaluated against an operational item pool. The results indicated that the item pools designed with stratification based on discrimination parameters performed well with an efficient use of the less discriminative items within the target accuracy levels. The implications for developing item pools are also discussed.

  10. C Reactor overbore test facility review

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, P.A.; Nilson, R.

    1964-04-24

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Bousbia-salah, Anis [Dipartimento di Ingegneria Meccanica, Nucleari e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2, 56126 Pisa (Italy)]. E-mail: b.salah@ing.unipi.it; Meftah, Brahim [Division Reacteur - Centre de Recherche Nucleaire Draria (CRND), BP 43 Sebala DRARIA - Algiers (Algeria); Hamidouche, Tewfik [Laboratoire des Analyses de Surete, Centre de Recherche Nucleaire d' Alger (CRNA), 02 Boulevard Frantz Fanon, B.P. 399, 16000 Algiers (Algeria)]. E-mail: thamidouche@comena-dz.org; Si-Ahmed, El Khider [Laboratoire des Ecoulements Polyhpasiques, Universite des Sciences et de la Technologie d' Alger, Algiers (Algeria)

    2006-03-15

    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. Under such conditions, a core overheat takes place, and the thermal energy 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 3D 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 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.

  12. 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.

  13. 1989 and 2010/11 Aquatic Areas Data: Mississippi River Navigation Pools 4, 8, 13, 26, Open River Section 2, and the Illinois River’s La Grange Pool

    Science.gov (United States)

    Ruhser, Janis; DeJager, Nathan R.; Rogala, James T.

    2016-01-01

    The U.S. Geological Survey's Upper Midwest Environmental Sciences Center (UMESC) has created high-resolution land cover/use data sets for the Upper Mississippi River System (UMRS). Aerial images of Pools 1-13 Upper Mississippi River System and Pools, Alton-Marseilles, Illinois River were collected in color infrared (CIR) in August of 2010 at 8”/pixel and 16”/pixel respectively using a mapping-grade Applanix DSS 439 digital aerial camera. In August 2011, CIR aerial images of Pools 14-Open River South, Upper Mississippi River and Pools Dresden-Lockport, Illinois River were collected at 16”/pixel with the same camera. The CIR aerial images were interpreted and automated using a 31-class LTRMP vegetation classification. These data have been used to create a variety of products, one of which is a data set used to classify aquatic areas. The 2010/11 aquatic areas data sets were created by first generalizing the available land cover/use data into a land/water data set, then reinterpreting the areas classified as water to determine the type of aquatic area. Area coverage for this data set is the Upper Mississippi River between Minneapolis, MN and Cairo, IL, and the Illinois River from its confluence with the Mississippi to Joliet, IL.

  14. 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.

  15. 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.

  16. Savannah River Reactor Operation: Indices of risk for emergency planning

    Energy Technology Data Exchange (ETDEWEB)

    O' Kula, K.R.; East, J.M.

    1990-10-01

    Periodically it is necessary to re-examine the implications of new source terms for neighboring offsite populations as Probabilistic Risk Assessment (PRA) and Severe Accident studies mature, and lead to a better understanding of the progression of hypothetical core melt accidents in the Savannah River Site (SRS) reactors. In this application multiple-system failure, low-frequency events, and consequently higher radiological source terms than from normal operation or design basis accidents (DBAs) are considered. Measures of consequence such as constant dose vs distance, boundary doses, and health effects to close-in populations are usually examined in this context. A set of source terms developed for the Safety Information Document (SID) for support of the Reactor Operation Environmental Impact Statement (EIS) forms the basis for the revised risk evaluation discussed herein. The intent of this review is not to completely substantiate the sufficiency of the current Emergency Planning Zone (EPZ). However, the two principal measures (200-rem red-bone marrow dose vs distance and 300-rem thyroid dose vs distance) for setting an EPZ are considered. Additional dose-at-distance calculations and consideration of DBA doses would be needed to complete a re-evaluation of the current EPZ. These subject areas are not addressed in the current document. Also, this report evaluates the sensitivity of individual risk estimates to the extent of offsite evacuation assumed from a K reactor severe accident and compares these risks to the Draft DOE Safety Guidelines. 14 refs., 8 figs., 4 tabs.

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

    Science.gov (United States)

    He, Wei; Reckase, Mark D.

    2014-01-01

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

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

    Science.gov (United States)

    Atanackovic, J; Matysiak, W; Hakmana Witharana, S S; Aslam, I; Dubeau, J; Waker, A J

    2013-01-01

    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 <0.2 mSv h⁻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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

  20. 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.

  1. 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.

  2. 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.

  3. 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).

  4. Dielectric Heaters for Testing Spacecraft Nuclear Reactors

    Science.gov (United States)

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

    2006-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-01

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

  6. 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.

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

    Science.gov (United States)

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

    2016-02-01

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

  8. Environmental Impact Study of the Northern Section of the Upper Mississippi River, St. Croix River Pool.

    Science.gov (United States)

    1973-11-01

    CYPERACEAE Carex norma is Sedge Carex sartwelii Sartwell’s sedge Carex stipata Awl-fruited sedge Carex nbellata Sedge Carex vuipinoidea Fox sedge ...the time of sampling had cattails and sedges ; nearby, also were watercress and duckweed. Trees on the bluff slopes include ri.ver maple, cottonwood, ash...islandica Icelandic yellow cress IRorippa obtusa Obtuse yellow cress IUnidentified sp. P CUCURBITACEAE ISicyos angulatus Bur-cucumber P l CYPERACEAE Carex

  9. 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.

  10. Chalk and Cheese

    Institute of Scientific and Technical Information of China (English)

    FRANCISCO; LITTLE

    2006-01-01

    Inever thought I would ever be attending a banquet on the balcony of a restaurant that looked out across the Yalujiang River, straight into the Democratic People's Republic of Korea (North Korea). There we were, a delegation from

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-01-04

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

  12. The sedimentology of redeposited chalk

    DEFF Research Database (Denmark)

    Anderskouv, Kresten; Surlyk, Finn; Gale, Andy

    Redeposited facies in the Upper Cretaceous Chalk Group constitute major hydrocarbon reservoirs in the North Sea Central Graben. Existing facies models are largely based on publications from the early 1980's dealing with core material from the Norwegian sector. However, the recognition, interpreta...

  13. Analysis of the Pentagon’s Press Pool Tests

    Science.gov (United States)

    1987-04-14

    the beginning of the Associated Press in 1849. Member organizations of a pool also have rules to abide by. As Victor Rosewater said: The original...parties to the arrangement. IVictor Rosewater , History of Cooperative News-Gathering in the United States (New York: D. Appleton and Company, 1930...Company, Inc., 1984. Reston, James. The Artillery of the Press. New York: harper & Row, 1967. Rosewater , Victor. History of Cooperative News

  14. 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. 

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

    CERN Document Server

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

    2002-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-01-01

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

  17. 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.

  18. 33 CFR 207.170 - Federal Dam, Oklawaha River, Moss Bluff, Fla.; pool level.

    Science.gov (United States)

    2010-07-01

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

  19. 33 CFR 207.60 - Federal Dam, Hudson River, Troy, N.Y.; pool level.

    Science.gov (United States)

    2010-07-01

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

  20. Analysis of production reactor response during a postulated Loss-of-River Water event using CONTAIN/SR

    Energy Technology Data Exchange (ETDEWEB)

    O' Kula, K.R.; Wooten, L.A. (Westinghouse Savannah River Co., Aiken, SC (United States)); Jenkins, T.B. (Concord Associates, Inc., Knoxville, TN (United States))

    1992-06-01

    This report discusses the CONTAIN/SR computer code, developed at the Savannah River Technology Center and Sandia National Laboratories for Probabilistic Safety Assessment (PSA) applications, which is used to analyze K Reactor plant conditions following a design basis earthquake to assist post-accident recovery planning. The postulated event, a Loss-of-River Water (LORW) accident, requires analysis of the K Reactor confinement system assuming seismic event-caused loss of forced air flow through Radiologically Controlled (RCAs) and other building areas, including adjoining personnel and auxiliary equipment zones. The CONTAIN/SR code calculations predict the expected environment in the K Reactor building with a seismically-qualified flow path for natural circulation, under design basis conditions specifying a 50 gal/min leak of tritiated heavy water. Despite loss of active fan flow, preferential air flow patterns are calculated to flow from clean'' areas towards the RCAs. Ventilation characteristics of the building reduce tritiated water vapor concentrations to habitable levels, assuming plastic suits and clean breathing air supplies are available. Unprotected dose rates to recovery workers in the heat exchanger zone of the building will range from 120 mrem/hour to 780 mrem/hour, depending on evaporation conditions near spilled heavy-water pools. It is concluded habitability issues for recovery are not driven by temperature concerns in reactor building zones. However, the results indicate radiological suits with cool air supplies will assure adequate conditions for operators and recovery teams, and mitigate tritium uptake hazards from splashing or other direct contact mechanisms.

  1. Analysis of production reactor response during a postulated Loss-of-River Water event using CONTAIN/SR

    Energy Technology Data Exchange (ETDEWEB)

    O`Kula, K.R.; Wooten, L.A. [Westinghouse Savannah River Co., Aiken, SC (United States); Jenkins, T.B. [Concord Associates, Inc., Knoxville, TN (United States)

    1992-06-01

    This report discusses the CONTAIN/SR computer code, developed at the Savannah River Technology Center and Sandia National Laboratories for Probabilistic Safety Assessment (PSA) applications, which is used to analyze K Reactor plant conditions following a design basis earthquake to assist post-accident recovery planning. The postulated event, a Loss-of-River Water (LORW) accident, requires analysis of the K Reactor confinement system assuming seismic event-caused loss of forced air flow through Radiologically Controlled (RCAs) and other building areas, including adjoining personnel and auxiliary equipment zones. The CONTAIN/SR code calculations predict the expected environment in the K Reactor building with a seismically-qualified flow path for natural circulation, under design basis conditions specifying a 50 gal/min leak of tritiated heavy water. Despite loss of active fan flow, preferential air flow patterns are calculated to flow from ``clean`` areas towards the RCAs. Ventilation characteristics of the building reduce tritiated water vapor concentrations to habitable levels, assuming plastic suits and clean breathing air supplies are available. Unprotected dose rates to recovery workers in the heat exchanger zone of the building will range from 120 mrem/hour to 780 mrem/hour, depending on evaporation conditions near spilled heavy-water pools. It is concluded habitability issues for recovery are not driven by temperature concerns in reactor building zones. However, the results indicate radiological suits with cool air supplies will assure adequate conditions for operators and recovery teams, and mitigate tritium uptake hazards from splashing or other direct contact mechanisms.

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

    Directory of Open Access Journals (Sweden)

    Kwon-Yeong Lee

    2015-01-01

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

  3. Sewage effluent clean-up reduces phosphorus but not phytoplankton in lowland chalk stream (River Kennet, UK) impacted by water mixing from adjacent canal.

    Science.gov (United States)

    Neal, Colin; Martin, Ellie; Neal, Margaret; Hallett, John; Wickham, Heather D; Harman, Sarah A; Armstrong, Linda K; Bowes, Mike J; Wade, Andrew J; Keay, David

    2010-10-15

    Information is provided on phosphorus in the River Kennet and the adjacent Kennet and Avon Canal in southern England to assess their interactions and the changes following phosphorus reductions in sewage treatment work (STW) effluent inputs. A step reduction in soluble reactive phosphorus (SRP) concentration within the effluent (5 to 13 fold) was observed from several STWs discharging to the river in the mid-2000s. This translated to over halving of SRP concentrations within the lower Kennet. Lower Kennet SRP concentrations change from being highest under base-flow to highest under storm-flow conditions. This represented a major shift from direct effluent inputs to a within-catchment source dominated system characteristic of the upper part to the catchment. Average SRP concentrations in the lower Kennet reduced over time towards the target for good water quality. Critically, there was no corresponding reduction in chlorophyll-a concentration, the waters remaining eutrophic when set against standards for lakes. Following the up gradient input of the main water and SRP source (Wilton Water), SRP concentrations in the canal reduced down gradient to below detection limits at times near its junction with the Kennet downstream. However, chlorophyll concentrations in the canal were in an order of magnitude higher than in the river. This probably resulted from long water residence times and higher temperatures promoting progressive algal and suspended sediment generations that consumed SRP. The canal acted as a point source for sediment, algae and total phosphorus to the river especially during the summer months when boat traffic disturbed the canal's bottom sediments and the locks were being regularly opened. The short-term dynamics of this transfer was complex. For the canal and the supply source at Wilton Water, conditions remained hypertrophic when set against standards for lakes even when SRP concentrations were extremely low.

  4. Chalk Line Mill, Anniston, AL

    Science.gov (United States)

    The Chalk Line Mill property was the site of a textile mill which operated from 1887 until 1994. Demolition activities in 2004 removed most of the structures on-site, but also left large, unsightly piles of debris scattered across this 14-acre property. The City applied for and received a $200,000 Brownfields cleanup grant in 2007 to address contamination on the property and the Appalachian Regional Commission provided an additional $150,000 in funding.

  5. Neutron activation analysis at the Livermore pool-type reactor for the environmental research program. [Identification of trace element contaminants

    Energy Technology Data Exchange (ETDEWEB)

    Ragaini, R.C.; Heft, R.E.; Garvis, D.

    1976-07-02

    Instrumental neutron activation analysis is a technique of trace analysis using measurements of radioactivity induced in the sample by exposure to a source of neutrons. The induced activity is measured by the emitted gamma radiation. Each gamma emitter can then be identified by the energy of the photopeaks produced as the nuclide decays and by the half-life of the neutron-induced activity. A complex computer program GAMANAL has been used to accomplish the major tasks of nuclide identification and quantification. The nuclide data output from GAMANAL is processed by a second computer code NADAC, which develops elemental abundance data from disintegration rates observed. The methods are those employed at the Livermore Pool-Type Reactor in support of the environmental research trace element analysis program. Among the procedures described and discussed are sample preparation, irradiation, analysis, and application of the technique.

  6. Risk Pooling, Commitment and Information: An experimental test of two fundamental assumptions

    OpenAIRE

    Abigail Barr

    2003-01-01

    This paper presents rigorous and direct tests of two assumptions relating to limited commitment and asymmetric information that current underpin current models of risk pooling. A specially designed economic experiment involving 678 subjects across 23 Zimbabwean villages is used to solve the problems of observability and quantification that have frustrated previous attempts to conduct such tests. I find that more extrinsic commitment is associated with more risk pooling, but that more informat...

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

    Science.gov (United States)

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

    2009-01-01

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

  8. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

    This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.

  9. The RES reactor, a test reactor for naval propulsion; Le reacteur d'essais RES, reacteur d'essais de la propulsion navale

    Energy Technology Data Exchange (ETDEWEB)

    Pivet, S. [CEA Bruyeres-le-Chatel, 91 (France); Minguet, J.L. [AREVA-Technicatome, 13 - Aix en Provence (France)

    2005-07-01

    The RES, the new test reactor for naval propulsion, will replace the RNG that nears the end of its operating life after 30 years in service. The main asset of a land-based installation is to provide an in-core instrumented reactor while the on-board system must stay as simple as possible for robustness reasons. The objective of the RES is fivefold: 1) to foresee and help solving problems likely to happen on on-board reactor, 2) to validate nuclear fuels and reactor systems for naval propulsion, 3) to validate reactor system and equipment for the Barracuda submarine program, 4) to upgrade the on-ground facility located at Cadarache, and 5) to provide the Cea with a new capacity for the storing of spent fuels from naval propulsion systems and from Cea research reactors. The RES facility is made of 2 parts: one that houses the reactor and the other that is dedicated to the handling on spent fuels, their examination through a gamma spectrometry bench and their storing in a pool. The RES facility is scheduled to open in 2009. (A.C.)

  10. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryskamp, J.M. [ed.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG&G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  11. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryskamp, J.M. (ed.); Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-09-01

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

  13. Chalk-microfluidic: flooding microsystems with reactive fluids

    Science.gov (United States)

    Neuville, Amélie; Dysthe, Dag Kristian; Li, Lei; Hiorth, Aksel

    2014-05-01

    Experiments on core scale and field tests that have been carried out the last decade have clearly shown that water chemistry affects the final oil recovery. However, there is generally no consensus in the scientific community of why additional oil is released. Part of the reason for this is that there are very few in-situ observations of how the water chemistry affects fluid distributions on the pore scale, and/or the pore surface characteristics. In this work, as a first step, our aim is to focus on in-situ observations of single phase flow and interactions at the pore scale. In order to work at this small scale, we first investigate how to control the flow location. We propose to use the same principle as "paper-microfluidic": some areas of the chalk are chemically treated so that no fluid flows inside while other areas let the fluids flow in the chalk pores. Since chalk and paper obviously has different mechanical behavior, we need to adapt this technique. Custom-made microsystems with chalk and calcite will be presented. We will then show experiments with reacting fluids in these microsystems. These experiments are observed using wide field fluorescence microscopy and white light vertical/phase shift interferometric microscopy.

  14. Vibration test on KMRR reactor structure and primary cooling system piping

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-10-01

    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author).

  15. 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.

  16. Poroelasticity of high porosity chalk under depletion

    DEFF Research Database (Denmark)

    Andreassen, Katrine Alling; Fabricius, Ida Lykke

    2013-01-01

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

  17. Self Compacting Concrete with Chalk Filler

    DEFF Research Database (Denmark)

    Sørensen, Eigil V.

    2007-01-01

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

  18. High-Temperature Gas-Cooled Test Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

    2016-04-01

    A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

  19. Entrained Flow Reactor Test of Potassium Capture by Kaolin

    DEFF Research Database (Denmark)

    Wang, Guoliang; Jensen, Peter Arendt; Wu, Hao

    2015-01-01

    In the present study a method to simulate the reaction between gaseous KCl and kaolin at suspension fired condition was developed using a pilot-scale entrained flow reactor (EFR). Kaolin was injected into the EFR for primary test of this method. By adding kaolin, KCl can effectively be captured......-bed reactor. The method using the EFR developed in this study will be applied for further systematic investigation of different additives....

  20. Investigation of Thermal Hydraulics of a Nuclear Reactor Moderator

    Science.gov (United States)

    Sarchami, Araz

    A three-dimensional numerical modeling of the thermo hydraulics of Canadian Deuterium Uranium (CANDU) nuclear reactor is conducted. The moderator tank is a Pressurized heavy water reactor which uses heavy water as moderator in a cylindrical tank. The main use of the tank is to bring the fast neutrons to the thermal neutron energy levels. The moderator tank compromises of several bundled tubes containing nuclear rods immersed inside the heavy water. It is important to keep the water temperature in the moderator at sub-cooled conditions, to prevent potential failure due to overheating of the tubes. Because of difficulties in measuring flow characteristics and temperature conditions inside a real reactor moderator, tests are conducted using a scaled moderator in moderator test facility (MTF) by Chalk River Laboratories of Atomic Energy of Canada Limited (CRL, AECL). MTF tests are conducted using heating elements to heat tube surfaces. This is different than the real reactor where nuclear radiation is the source of heating which results in a volumetric heating of the heavy water. The data recorded inside the MTF tank have shown levels of fluctuations in the moderator temperatures and requires in depth investigation of causes and effects. The purpose of the current investigation is to determine the causes for, and the nature of the moderator temperature fluctuations using three-dimensional simulation of MTF with both (surface heating and volumetric heating) modes. In addition, three dimensional simulation of full scale actual moderator tank with volumetric heating is conducted to investigate the effects of scaling on the temperature distribution. The numerical simulations are performed on a 24-processor cluster using parallel version of the FLUENT 12. During the transient simulation, 55 points of interest inside the tank are monitored for their temperature and velocity fluctuations with time.

  1. 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

  2. Environmental Impact Study of the Northern Section of the Upper Mississippi River. Pool 8.

    Science.gov (United States)

    1973-11-01

    Potamogeton crispus most common. 00/ ’.1~ )~iU% ~N -. %.1 ’ ~ N POOL 8 TRANSECT B Taxon Collection No. SUBMERGENT, AQUATIC PLANTS Elodea canadensis Michx...8217I U _POOL 8 TRANSECT B Taxon Collection No. SUBMERGENT AQUATIC PLANTS Ceratophyllum demersum L. 7609 Elodea canadensis Michx. 7606 Potamogeton...and P. pectinatus most abundant. POOL 8 TRANSECT B Taxon Collection No. SUBMERGENT, AQUATIC PLANTS Ceratophyllum demersum L. 7531 ,7537 Elodea

  3. Characterization of spent fuel elements stored at IEA-R1 research reactor based on visual inspections and sipping tests

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

    Aluminum spent nuclear fuels are susceptible to corrosion attack, or mechanical damage from improper handling, while in pool reactor storage. Storage practices have been modified to reduce the potential for damage, based on recommendations presented at second WS on Spent Fuel Characterization, promoted by IAEA. In this work, we present the inspection program proposed to the IEA-R1 stored spent fuel elements, in order to provide information on the physical condition during the interim storage time under wet condition at the reactor pool. The inspection program is based on non-destructive tests results (visual inspection and sipping tests) already periodically performed to exam the IEA-R1 stored spent fuel and fuel elements from the core reactor. To record the available information and examination results it was elaborated a document in the format of a catalogue containing the proposed inspection program for the IEA-R1 stored spent fuel, the description of the visual inspection and sipping tests systems, a compilation of information and images result from the tests performed for all stored standard spent fuel element and, in annexes, copies of the reference documents. That document constitutes an important step of the effective implementation of the referred IEA-R1 spent fuel inspection program and can be used to address regulatory and operational needs for the demonstration, for example, of safe storage throughout the pool storage period. (author)

  4. UMRR Topobathy Data (Tier 3) of Mississippi Pools 25, Open River North, Open River South and Illinois River Starved Rock Reach

    Science.gov (United States)

    Stone, Jayme; Rogala, James T.; Sattler, Stephanie; Hanson, Jenny L.

    2016-01-01

    These data are associated with the U.S. Army Corps of Engineers’ Upper Mississippi River Restoration (UMRR) Program project to generate high resolution elevation datasets for the Upper Mississippi River floodplain, between Minneapolis, MN, and Cairo, IL and the Illinois River. The Tier 1 data sets were designed to make the lidar data publicly available as soon as possible. The Tier 2 data have undergone quality assurance testing, water-masking, contour-smoothing, and point-reclassification where necessary, and have been approved for release by the USGS. The Tier 3 data are products of bathymetry merged with terrestrial lidar. UMESC is the lead office for the final processing of these data, which includes additional quality control and serving the data on the UMRR Long Term Resource Monitoring (LTRM) element’s website.

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

    Science.gov (United States)

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

    2010-01-01

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

  6. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris{o} National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately.

  7. Reactor coolant pump testing using motor current signatures analysis

    Energy Technology Data Exchange (ETDEWEB)

    Burstein, N.; Bellamy, J.

    1996-12-01

    This paper describes reactor coolant pump motor testing carried out at Florida Power Corporation`s Crystal River plant using Framatome Technologies` new EMPATH (Electric Motor Performance Analysis and Trending Hardware) system. EMPATH{trademark} uses an improved form of Motor Current Signature Analysis (MCSA), technology, originally developed at Oak Ridge National Laboratories, for detecting deterioration in the rotors of AC induction motors. Motor Current Signature Analysis (MCSA) is a monitoring tool for motor driven equipment that provides a non-intrusive means for detecting the presence of mechanical and electrical abnormalities in the motor and the driven equipment. The base technology was developed at the Oak Ridge National Laboratory as a means for determining the affects of aging and service wear specifically on motor-operated valves used in nuclear power plant safety systems, but it is applicable to a broad range of electric machinery. MCSA is based on the recognition that an electric motor (ac or dc) driving a mechanical load acts as an efficient and permanently available transducer by sensing mechanical load variations, large and small, long-term and rapid, and converting them into variations in the induced current generated in the motor windings. The motor current variations, resulting from changes in load caused by gears, pulleys, friction, bearings, and other conditions that may change over the life of the motor, are carried by the electrical cables powering the motor and are extracted at any convenient location along the motor lead. These variations modulate the 60 Hz carrier frequency and appear as sidebands in the spectral plot.

  8. 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,...

  9. 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,...

  10. 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,...

  11. 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,...

  12. 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,...

  13. 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,...

  14. 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,...

  15. 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,...

  16. Final safeguards analysis, high temperature lattice test reactor. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Hanthorn, H.E.; Brown, W.W.; Clark, R.G.; Heineman, R.E.; Humes, R.M.

    1966-01-01

    The PMACS `reactor-normal` signal signifies that important process variables do not exceed their set points, that various interlocks are properly set, that functional tests of the computer operation are satisfactory, and that the reactor flux level and period derived from two additional, independent, and dissimilar channels are within set limits. This safety circuit combines the features of redundancy, dissimilar components, and frequent testing which are required for best reliability. The experimental equipment auxiliary to the reactor includes two oscillator mechanisms, one to move the test cell or the adjoining cell into and out of position, the other to move small specimens in the test cell or adjoining cells. They have cooling chambers for the removal of specimens from the test cell without the necessity of cooling the reactor. A neutron chopper and time-of-flight spectrometer are provided; the neutron detectors, at the end of a 25-meter flight tube, are in an adjoining small building. Test cores may be assembled on a core dolly have a load capacity of 14,000 lb. Two wire traverse mechanisms are provided for measurements of flux distribution.

  17. Restart of K-Reactor, Savannah River Site: Safety evaluation report

    Energy Technology Data Exchange (ETDEWEB)

    1991-04-01

    This Safety Evaluation Report (SER) focuses on those issues required to support the restart of the K-Reactor at the Savannah River Plant. This SER provides the safety criteria for restart and documents the results of the staff reviews of the DOE and operating contractor activities to meet these criteria. To develop the restart criteria for the issues discussed in this SER, the Savannah River Restart Office and Savannah River Special Projects Office staffs relied, when possible, on commercial industry codes and standards and on NRC requirements and guidelines for the commercial nuclear industry. However, because of the age and uniqueness of the Savannah River reactors, criteria for the commercial plants were not always applicable. In these cases, alternate criteria were developed. The restart criteria applicable to each of the issues are identified in the safety evaluations for each issue. The restart criteria identified in this report are intended to apply only to restart of the Savannah River reactors. Following the development of the acceptance criteria, the DOE staff and their support contractors evaluated the results of the DOE and operating contractor (WSRC) activities to meet these criteria. The results of those evaluations are documented in this report. Deviations or failures to meet the requirements are either justified in the report or carried as open or confirmatory items to be completed and evaluated in supplements to this report before restart. 62 refs., 1 fig.

  18. The technology development for surveillance test of reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Sun Phil; Park, Day Young; Choi, Kwen Jai

    1997-12-01

    Benchmark test was performed in accordance with the requirement of US NRC Reg. Guide DG-1053 for Kori unit-1 in order to determine best-estimated fast neutron fluence irradiated into reactor vessel. Since the uncertainty of radiation analysis comes from the calculation error due to neutron cross-section data, reactor core geometrical dimension, core source, mesh density, angular expansion and convergence criteria, evaluation of calculational uncertainty due to analytical method was performed in accordance with the regulatory guide and the proof was performed for entire analysis by comparing the measurement value obtained by neutron dosimetry located in surveillance capsule. Best-estimated neutron fluence in reactor vessel was calculated by bias factor, neutron flux measurement value/calculational value, from reanalysis result from previous 1st through 4th surveillance testing and finally fluence prediction was performed for the end of reactor life and the entire period of plant life extension. Pressurized thermal shock analysis was performed in accordance with 10 CFR 50.61 using the result of neutron fluence analysis in order to predict the life of reactor vessel material and the criteria of safe operation for Kori unit 1 was reestablished. (author). 55 refs., 55 figs.

  19. Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

    Science.gov (United States)

    Bragg-Sitton, Shannon M.

    2005-01-01

    The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .tests, the point kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

  20. Optimal stratification of item pools in α-stratified computerized adaptive testing

    NARCIS (Netherlands)

    Chang, Hua-Hua; Linden, van der Wim J.

    2003-01-01

    A method based on 0-1 linear programming (LP) is presented to stratify an item pool optimally for use in α-stratified adaptive testing. Because the 0-1 LP model belongs to the subclass of models with a network flow structure, efficient solutions are possible. The method is applied to a previous item

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

    Science.gov (United States)

    He, Wei

    2010-01-01

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

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

    Directory of Open Access Journals (Sweden)

    E. Samaritani

    2011-06-01

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

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

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

  3. 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

  4. Irradiation Testing Vehicles for Fast Reactors from Open Test Assemblies to Closed Loops

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-12-15

    A review of irradiation testing vehicle approaches and designs that have been incorporated into past Sodium-Cooled Fast Reactors (SFRs) or envisioned for incorporation has been carried out. The objective is to understand the essential features of the approaches and designs so that they can inform test vehicle designs for a future U.S. Fast Test Reactor. Fast test reactor designs examined include EBR-II, FFTF, JOYO, BOR-60, PHÉNIX, JHR, and MBIR. Previous designers exhibited great ingenuity in overcoming design and operational challenges especially when the original reactor plant’s mission changed to an irradiation testing mission as in the EBRII reactor plant. The various irradiation testing vehicles can be categorized as: Uninstrumented open assemblies that fit into core locations; Instrumented open test assemblies that fit into special core locations; Self-contained closed loops; and External closed loops. A special emphasis is devoted to closed loops as they are regarded as a very desirable feature of a future U.S. Fast Test Reactor. Closed loops are an important technology for irradiation of fuels and materials in separate controlled environments. The impact of closed loops on the design of fast reactors is also discussed in this report.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-26

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

  6. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jennifer Lyons; Wade R. Marcum; Mark D. DeHart; Sean R. Morrell

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  7. Effects of pool formation and flash flooding on relative abundance of young-of-year flannelmouth suckers in the Paria River, Arizona

    Science.gov (United States)

    Thieme, Michele L.; McIvor, C.C.; Brouder, Mark J.; Hoffnagle, T.L.

    2001-01-01

    Flannelmouth sucker, Catostomus latipinnis, a fish endemic to the Colorado River basin in the western United States, appears to experience poor recruitment to adult size in the Colorado River, downstream of Glen Canyon Dam. Lack or impermanence of rearing areas for young-of-year (YOY) fish is hypothesized to be the problem. Knowing the importance of tributary mouths as rearing areas in other river systems, we studied use of the mouth of the Paria River, a tributary of the Colorado River, by YOY flannelmouth suckers, and the availability of rearing area in the mouth at different flow levels in the Colorado River in 1996 and 1997. We also examined the relationship between flash floods in the Paria River and catch-per-unit-effort (CPUE) of YOY in the Paria River between 1991 and 1996. Maximum mean daily discharge in the Paria River was inversely correlated with CPUE of YOY flannelmouth suckers (Spearman Rho = -0.9856, p = 0.0003) during their critical rearing period (15 March-30 June). Thus, it appears that YOY flannelmouth suckers rear longer in the Paria River in years when flash flooding is minimal. Recruitment of YOY flannelmouth suckers at the Paria River may also be improved by enhancing pool formation during spring and summer rearing seasons. YOY flannelmouth sucker was captured in a pool created by high Colorado River flows (??? 336 m3/s) that inundated the mouth of the Paria River during spring and summer, 1996. In 1997, high flows (about 550-750 m3/s) in the Colorado River during winter and spring initially inundated the Paria River and formed a pool in the mouth. However, these high flows eventually caused 0.5-1.0 m of suspended sediment from the incoming Paria River to deposit in the mouth. Thus, despite higher flows than 1996, the slackwater area formed only occasionally in 1997. Differences in pool formation between 1996 and 1997 demonstrate that pool formation cannot be inferred solely from Colorado River flows. ?? 2001 John Wiley & Sons, Ltd.

  8. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1988-05-01

    The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm/sup 3/ was by then in routine use, illustrated how far work has progressed.

  9. Standard Guide for Benchmark Testing of Light Water Reactor Calculations

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with ...

  10. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  11. Partial Nucleate Pool Boiling at Low Heat Flux: Preliminary Ground Test for SOBER-SJ10

    Science.gov (United States)

    Wu, Ke; Li, Zhen-Dong; Zhao, Jian-Fu; Li, Hui-Xiong; Li, Kai

    2016-05-01

    Focusing on partial nucleate pool boiling at low heat flux, SOBER-SJ10, one of 27 experiments of the program SJ-10, has been proposed to study local convection and heat transfer around an isolated growing vapor bubble during nucleate pool boiling on a well characterized flat surface in microgravity. An integrated micro heater has been developed. By using a local pulse overheating method in the experimental mode of single bubble boiling, a bubble nucleus can be excited with accurate spatial and temporal positioning on the top-side of a quartz glass substrate with a thickness of 2 mm and an effective heating area of 4.5 mm in diameter, and then grows under an approximate constant heat input provided by the main heater on the back-side of the substrate. Ten thin film micro-RTDs are used for local temperature measurements on the heating surface underneath the growing bubble. Normal pool boiling experiments can also be carried out with step-by-step increase of heating voltage. A series of ground test of the flight module of SOBER-SJ10 have been conducted. Good agreement of the measured data of single phase natural convection with the common-used empirical correlation warrants reasonable confidence in the data. It is found that the values of the incipience superheat of pool boiling at different subcooling are consistent with each others, verifying that the influence of subcooling on boiling incipience can be neglected. Pool boiling curves are also obtained, which shows great influence of subcooling on heat transfer of partial nucleate pool boiling, particularly in lower heat flux.

  12. Spatial and temporal dynamics of suspended particle characteristics and composition in Navigation Pool 19 of the Upper Mississippi River

    Science.gov (United States)

    Milde, Amanda S.; Richardson, William B.; Strauss, Eric A.; Larson, James H.; Vallazza, Jon; Knights, Brent C.

    2017-01-01

    Suspended particles are an essential component of large rivers influencing channel geomorphology, biogeochemical cycling of nutrients, and food web resources. The Upper Mississippi River is a large floodplain river that exhibits pronounced spatiotemporal variation in environmental conditions and biota, providing an ideal environment for investigating dynamics of suspended particles in large river ecosystems. Here we investigated two questions: (i) How do suspended particle characteristics (e.g. size and morphology) vary temporally and spatially? and (ii) What environmental variables have the strongest association with particle characteristics? Water sampling was conducted in June, August, and September of 2013 and 2014 in Navigation Pool 19 of the Upper Mississippi River. A FlowCAM® (Flow Cytometer and Microscope) particle imaging system was used to enumerate and measure particles 53–300 μm in diameter for size and shape characteristics (e.g. volume, elongation, and symmetry). Suspended particle characteristics varied considerably over space and time and were strongly associated with discharge and concentrations of nitrate + nitrite (NO3−) and soluble reactive phosphorus. Particle characteristics in backwaters were distinct from those in other habitats for most of the study period, likely due to reduced hydrologic connectivity and higher biotic production in backwaters. During low discharge, phytoplankton and zooplankton made up relatively greater proportions of the observed particles. Concurrently during low discharge, concentrations of chlorophyll, volatile suspended solids, and total phosphorus were higher. Our results suggest that there are complex interactions among space, time, discharge, and other environmental variables (e.g. water nutrients), which drive suspended particle dynamics in large rivers.

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

    Directory of Open Access Journals (Sweden)

    Nitin eUdpa

    2011-11-01

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

  14. Testing piezoelectric sensors in a nuclear reactor environment

    Science.gov (United States)

    Reinhardt, Brian T.; Suprock, Andy; Tittmann, Bernhard

    2017-02-01

    Several Department of Energy Office of Nuclear Energy (DOE-NE) programs, such as the Fuel Cycle Research and Development (FCRD), Advanced Reactor Concepts (ARC), Light Water Reactor Sustainability, and Next Generation Nuclear Power Plants (NGNP), are investigating new fuels, materials, and inspection paradigms for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials during irradiation. In DOE-NE's FCRD program, ultrasonic based technology was identified as a key approach that should be pursued to obtain the high-fidelity, high-accuracy data required to characterize the behavior and performance of new candidate fuels and structural materials during irradiation testing. The radiation, high temperatures, and pressure can limit the available tools and characterization methods. In this work piezoelectric transducers capable of making these measurements are developed. Specifically, three piezoelectric sensors (Bismuth Titanate, Aluminum Nitride, and Zinc Oxide) are tested in the Massachusetts Institute of Technology Research reactor to a fast neutron fluence of 8.65×1020 nf/cm2. It is demonstrated that Bismuth Titanate is capable of transduction up to 5 × 1020 nf/cm2, Zinc Oxide is capable of transduction up to at least 6.27 × 1020 nf/cm2, and Aluminum Nitride is capable of transduction up to at least 8.65 × 1020 nf/cm2.

  15. Cultural Resources Literature Search and Records Review - Upper Mississippi River Basin. Volume 7. Pool 6.

    Science.gov (United States)

    1983-01-01

    topographic- ILocation of Collections: Wagner, Ken , (Trempealeau, WI) and Gallup Clark (L aCrosse, WI). L A 52 A I POOL 6 53 Map # 17 (14 4)State...Wi A ’S’ POOL 6 75 Map * 28(145) State Codification # TR-0057 County Trempealeau(WI) Site Name: Wilber Legal Description: Township & Range T18N R9W...Civil Township Trempeale Section Location: SW , SEI , SE% sec 17 Present Owner: Wilber , Elmer Address: Wilber Road , Trempealeau , Wi Recorded By: (name

  16. Enhanced in-pile instrumentation at the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T.; Chase, B. M.; Palmer, J.; Condie, K. G.; Davis, K. L. [Idaho National Laboratory, MS 3840, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

    2011-07-01

    Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper reports results from this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and realtime flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted. (authors)

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

    DEFF Research Database (Denmark)

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

    2011-01-01

    Changes in chalk due to EOR by injecting supercritical CO2 (CO2-EOR) can ideally be predicted by applying geophysical methods designed from laboratory-determined petrophysical and rock mechanics properties. A series of petrophysical and rock mechanics tests were performed on Ekofisk Formation...... and Tor Formation chalk of the South Arne field to reveal the changes in petrophysical and rock mechanics properties of chalk due to the injection of CO2 at supercritical state. An increase in porosity and decrease in specific surface was observed due to injection of supercritical CO2. This indicates...... that a reaction between CO2 enriched water and particles takes place which smoothens the particle surface. Accordingly, partial increase in permeability was also noticed. An effect is also observed from the decrease in pore-space stiffness, calculated from sonic velocity. No significant effect on wettability...

  18. Benthic macrofauna variations and community structure in Cenomanian cyclic chalk-marl from Southerham Grey Pit, SE England

    DEFF Research Database (Denmark)

    Lauridsen, Bodil Wesenberg; Gale, A. S.; Surlyk, Finn

    2009-01-01

    Cenomanian chalk-marl couplets from England represent the 20 ka Milankovitch precession cycle. Fossil communities from both chalk and marl are identified to test if the orbital fluctuations and the associated changes in substrate lithology and climate exerted any control on the benthic macrofauna...... adapted to both facies and thus to the fine grain size of the substrate rather than to lithology. The systematic difference in diversity between chalk and marl samples was possibly caused by long-term climatic and oceanographic changes and thus could represent a biological response to Milankovitch...

  19. Conceptual design of the Clinch River Breeder Reactor spent-fuel shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    Pope, R B; Diggs, J M [eds.

    1982-04-01

    Details of a baseline conceptual design of a spent fuel shipping cask for the Clinch River Breeder Reactor (CRBR) are presented including an assessment of shielding, structural, thermal, fabrication and cask/plant interfacing problems. A basis for continued cask development and for new technological development is established. Alternates to the baseline design are briefly presented. Estimates of development schedules, cask utilization and cost schedules, and of personnel dose commitments during CRBR in-plant handling of the cask are also presented.

  20. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 5a

    Data.gov (United States)

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

  1. 2002 Land Cover/Use Data for Pool 8 - Upper Mississippi River

    Data.gov (United States)

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

  2. 1989-91 Aquatic Habitats - Upper Mississippi River System - Pool 07

    Data.gov (United States)

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

  3. Bathymetric data for the Upper Mississippi and Illinois Rivers -- Pool 10

    Data.gov (United States)

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

  4. Bathymetric data for the Upper Mississippi and Illinois Rivers -- Pool 08

    Data.gov (United States)

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

  5. 2000 Aerial Photo Mosaics - Upper Mississippi River System -- Pool 8 Color Infrared

    Data.gov (United States)

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

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

    Data.gov (United States)

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

  7. 2010 UMRS Color Infrared Aerial Photo Mosaic - Illinois River, Marseilles Pool

    Data.gov (United States)

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

  8. 2010/11 Aquatic Areas - 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  9. 2011 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 25

    Data.gov (United States)

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

  10. 2011 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 18

    Data.gov (United States)

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

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

    Data.gov (United States)

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

  12. 1989-91 Aquatic Habitats - Upper Mississippi River System - Brandon Pool

    Data.gov (United States)

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

  13. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  14. 1989 Land Cover/Use Data for the Upper Mississippi River System--Brandon Pool

    Data.gov (United States)

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

  15. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  16. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  17. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  18. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  19. 1989 Land Cover/Use Data for the Upper Mississippi River System--Marseillies Pool

    Data.gov (United States)

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

  20. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  1. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  2. 1989 Land Cover/Use Data for the Upper Mississippi River System--Lockport Pool

    Data.gov (United States)

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

  3. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  4. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  5. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  6. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  7. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  8. 1989 Land Cover/Use Data for the Upper Mississippi River System--Peoria Pool

    Data.gov (United States)

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

  9. 1989 Land Cover/Use Data for the Upper Mississippi River System--Pool 5a

    Data.gov (United States)

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

  10. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  11. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  12. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  13. 1989 Land Cover/Use Data for the Upper Mississippi River System--Dresden Pool

    Data.gov (United States)

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

  14. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  15. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  16. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  17. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  18. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  19. 1989 Land Cover/Use Data for the Upper Mississippi River System--LaGrange Pool

    Data.gov (United States)

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

  20. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  1. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  2. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  3. 1989 Land Cover/Use Data for the Upper Mississippi River System--Starved Rock Pool

    Data.gov (United States)

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

  4. 1989 Land Cover/Use Data for the Upper Mississippi River System--Alton Pool

    Data.gov (United States)

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

  5. 1989 Land Cover/Use Data for the 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  6. 1989-91 Aquatic Habitats - 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  7. 2010 UMRS Color Infrared Aerial Photo Mosaic - Illinois River, Alton Pool

    Data.gov (United States)

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

  8. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 5

    Data.gov (United States)

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

  9. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 6

    Data.gov (United States)

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

  10. 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...

  11. 2011 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 15

    Data.gov (United States)

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

  12. 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...

  13. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 3

    Data.gov (United States)

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

  14. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 13 South

    Data.gov (United States)

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

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

    Data.gov (United States)

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

  16. 2011 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 20

    Data.gov (United States)

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

  17. 2011 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 19

    Data.gov (United States)

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

  18. 2011 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 24

    Data.gov (United States)

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

  19. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 11

    Data.gov (United States)

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

  20. 2011 UMRS Color Infrared Aerial Photo Mosaic - Illinois River, Lockport Pool

    Data.gov (United States)

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

  1. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 7

    Data.gov (United States)

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

  2. 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...

  3. 2011 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 21

    Data.gov (United States)

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

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

    Data.gov (United States)

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

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

    Data.gov (United States)

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

  6. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 4 North

    Data.gov (United States)

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

  7. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 2

    Data.gov (United States)

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

  8. 2011 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 22

    Data.gov (United States)

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

  9. 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...

  10. 2010 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 12

    Data.gov (United States)

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

  11. 2011 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 17

    Data.gov (United States)

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

  12. 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...

  13. 2011 UMRS Color Infrared Aerial Photo Mosaic - Illinois River, Dresden Pool

    Data.gov (United States)

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

  14. 2011 UMRS Color Infrared Aerial Photo Mosaic - Mississippi River, Pool 26

    Data.gov (United States)

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

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

    Data.gov (United States)

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

  16. 2010 UMRS Color Infrared Aerial Photo Mosaic - Illinois River, Starved Rock Pool

    Data.gov (United States)

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

  17. 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...

  18. 2010 UMRS Color Infrared Aerial Photo Mosaic - Illinois River, Peoria Pool

    Data.gov (United States)

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

  19. 2010/11 Aquatic Areas - Upper Mississippi River System - Pool 08

    Data.gov (United States)

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

  20. 1989-91 Aquatic Habitats - 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  1. 1989-91 Aquatic Habitats - 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) has created high-resolution land cover/use data sets for the Upper Mississippi River...

  2. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baek, W. P.; Song, C. H.; Kim, Y. S. and others

    2005-02-15

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

  3. Steady-state and loss-of-pumping accident analyses of the Savannah River new production reactor representative design

    Energy Technology Data Exchange (ETDEWEB)

    Pryor, R.J.; Maloney, K.J.

    1990-10-01

    This document contains the steady-state and loss-of-pumping accident analysis of the representative design for the Savannah River heavy water new production reactor. A description of the reactor system and computer input model, the results of the steady-state analysis, and the results of four loss-of-pumping accident calculations are presented. 5 refs., 37 figs., 4 tabs.

  4. Core melt progression and consequence analysis methodology development in support of the Savannah River Reactor PSA

    Energy Technology Data Exchange (ETDEWEB)

    O' Kula, K.R.; Sharp, D.A. (Westinghouse Savannah River Co., Aiken, SC (United States)); Amos, C.N.; Wagner, K.C.; Bradley, D.R. (Science Applications International Corp., Albuquerque, NM (United States))

    1992-01-01

    A three-level Probabilistic Safety Assessment (PSA) of production reactor operation has been underway since 1985 at the US Department of Energy's Savannah River Site (SRS). The goals of this analysis are to: Analyze existing margins of safety provided by the heavy-water reactor (HWR) design challenged by postulated severe accidents; Compare measures of risk to the general public and onsite workers to guideline values, as well as to those posed by commercial reactor operation; and Develop the methodology and database necessary to prioritize improvements to engineering safety systems and components, operator training, and engineering projects that contribute significantly to improving plant safety. PSA technical staff from the Westinghouse Savannah River Company (WSRC) and Science Applications International Corporation (SAIC) have performed the assessment despite two obstacles: A variable baseline plant configuration and power level; and a lack of technically applicable code methodology to model the SRS reactor conditions. This paper discusses the detailed effort necessary to modify the requisite codes before accident analysis insights for the risk assessment were obtained.

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

    Science.gov (United States)

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

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

  6. Advanced Gas Reactor (AGR)-5/6/7 Fuel Irradiation Experiments in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; David A. Petti; S. Blaine Grover

    2014-04-01

    The United States Department of Energy’s Very High Temperature Reactor (VHTR) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which each consist of at least five separate capsules, are being irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gases also have on-line fission product monitoring the effluent from each capsule to track performance of the fuel during irradiation. The first two experiments (designated AGR-1 and AGR-2), have been completed. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. The design of the fuel qualification experiment, designated AGR-5/6/7, is well underway and incorporates lessons learned from the three previous experiments. Various design issues will be discussed with particular details related to selection of thermometry.

  7. The Jules Horowitz reactor (JHR), a European material testing reactor (MTR), with extended experimental capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A.; Bergamaschi, Y.; Bouilloux, Y.; Bravo, X.; Guigon, B.; Rommens, M.; Tremodeux, P. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, 13 - Saint-Paul-lez-Durance (France)]|[CEA Saclay Dir. de l' Energie Nucleaire DEN, 91 - Gif sur Yvette (France)

    2003-07-01

    The Jules Horowitz Reactor (JHR) is the European MTR (Material Testing Reactor) designed to provide, after 2010, the necessary knowledge for keeping the existing power plants in operation and to design innovative reactors types with new objectives such as: minimizing the radioactive waste production, taking into account additional safety requirements, preventing risks of nuclear proliferation... To achieve such an ambitious objective. The JHR is designed with a high flexibility in order to satisfy the current demand from European industry, research and to be able to accommodate future requirements. The JHR will offer a wide range of performances and services in gathering, in a single site at Cadarache, all the necessary functionalities and facilities for an effective production of results: e.g. fuel fabrication laboratories, preparation of the instrumented devices, interpretation of the experiments, modelling. The JHR must rely on a top level scientific environment based on experts teams from CEA and EC and local universities. With a thermal flux of 7,4.10{sup 14} ncm{sup -2} s{sup -1} and a fast flux of 6,4.10{sup 14} ncm{sup -2}s{sup -1}, it is possible to carry out irradiation experiments on materials and fuels whatever the reactor type considered. It will also be possible to carry out locally, fast neutron irradiation to achieve damage effect up to 25 dpa/year. (dpa = displacement per atom.) The study of the fuels behavior under accidental conditions, from analytical experiments, on a limited amount of irradiated fuel, is a major objective of the project. These oriented safety tests are possible by taking into account specific requirements in the design of the facility such as the tightness level of the containment building, the addition of an alpha hot cell and a laboratory for on line fission products measurement. (authors)

  8. Management of New Production Reactor waste streams at Savannah River

    Energy Technology Data Exchange (ETDEWEB)

    McDonell, W.R.; Newman, J.L.

    1992-12-31

    To ensure the adequacy of available facilities, the disposition of the several waste types generated in support of a heavy-water NPR operation at the Savannah River Site were projected through waste- treatment and disposal facilities after the year 2000. Volumes of high-level, low-level radioactive, TRU, hazardous, mixed and non-radioactive waste were predicted for early assessments of environmental impacts and to provide a baseline for future waste-minimization initiatives. Life-cycle unit costs for disposal of the waste, adjusted to reflect waste management capabilities in the NPR operating time frame, were developed to evaluate the economic effectiveness of waste-minimization activities in the NPR program.

  9. Management of New Production Reactor waste streams at Savannah River

    Energy Technology Data Exchange (ETDEWEB)

    McDonell, W.R.; Newman, J.L.

    1992-01-01

    To ensure the adequacy of available facilities, the disposition of the several waste types generated in support of a heavy-water NPR operation at the Savannah River Site were projected through waste- treatment and disposal facilities after the year 2000. Volumes of high-level, low-level radioactive, TRU, hazardous, mixed and non-radioactive waste were predicted for early assessments of environmental impacts and to provide a baseline for future waste-minimization initiatives. Life-cycle unit costs for disposal of the waste, adjusted to reflect waste management capabilities in the NPR operating time frame, were developed to evaluate the economic effectiveness of waste-minimization activities in the NPR program.

  10. New results from pulse tests in the CABRI reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schmitz, F.; Papin, J.; Haessler, M. [Institut de Proterction et de Surete Nucleaire, Saint Paul Lez Durance (France)] [and others

    1996-03-01

    At the 21st and 22nd WRSM (1,2), the motivation and objectives of the French program on the behaviour of high burnup PWR fuel under RIA conditions in the CABRI test reactor has been presented. The major results of the three first tests of the test matrix were presented and in particular REP-Na1, which failed at an unexpected low level of fuel enthalpy, was exposed to the community of nuclear safety research. At this time, no final understanding was reached for the origin of the failure. This objective is reached now. Two further tests, REP-Na4 and 5, have been performed in 1995, they demonstrated a satisfactory and safe behaviour by resisting to the early phase of severe loading during the RIA pulse test. Further examination work and analytical testing is in progress and the next tests with MOX fuel are being prepared.

  11. Adsorption of hydrocarbons in chalk reservoirs

    Energy Technology Data Exchange (ETDEWEB)

    Madsen, L.

    1996-12-31

    The present work is a study on the wettability of hydrocarbon bearing chalk reservoirs. Wettability is a major factor that influences flow, location and distribution of oil and water in the reservoir. The wettability of the hydrocarbon reservoirs depends on how and to what extent the organic compounds are adsorbed onto the surfaces of calcite, quartz and clay. Organic compounds such as carboxylic acids are found in formation waters from various hydrocarbon reservoirs and in crude oils. In the present investigation the wetting behaviour of chalk is studied by the adsorption of the carboxylic acids onto synthetic calcite, kaolinite, quartz, {alpha}-alumina, and chalk dispersed in an aqueous phase and an organic phase. In the aqueous phase the results clearly demonstrate the differences between the adsorption behaviour of benzoic acid and hexanoic acid onto the surfaces of oxide minerals and carbonates. With NaCl concentration of 0.1 M and with pH {approx_equal} 6 the maximum adsorption of benzoic acid decreases in the order: quartz, {alpha}-alumina, kaolinite. For synthetic calcite and chalk no detectable adsorption was obtaind. In the organic phase the order is reversed. The maximum adsorption of benzoic acid onto the different surfaces decreases in the order: synthetic calcite, chalk, kaolinite and quartz. Also a marked difference in adsorption behaviour between probes with different functional groups onto synthetic calcite from organic phase is observed. The maximum adsorption decreases in the order: benzoic acid, benzyl alcohol and benzylamine. (au) 54 refs.

  12. Environmental Impact Study of the Northern Section of the Upper Mississippi River. Pool 3.

    Science.gov (United States)

    1973-11-01

    Willow) Cyperaceas Typhaceae I( Sedges ) Eleocharis opp. Typha latifolia jScirpusi spp. (Cattail) Lemunaceae I(Duckweed) Lemna minor Wolf fti p2iinctzta AIM...acted as a substrate for the anchoring of riverine successional growth such as Willows, Grasses, Sedges , Poplars, and other biota which would be found on...than grasses and sedges and a few spike-rushes. The bank is actually a hand-constructed stone wall about fifteen feet high, with the Pool Three

  13. Final Physics Report for the Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wolfe, I. B. [Savannah River Site (SRS), Aiken, SC (United States)

    1956-06-25

    This report is a summary of the physics design work performed on the Engineering Test Reactor. The ETR presents computational difficulties not found in other reactors because of the large number of experimental holes in the core. The physics of the ETR depends strongly upon the contents of the in-core experimental facilities. In order to properly evaluate the reactor; taking into account the experiments in the core, multi-region, two-dimensional calculations are required. These calculations require .the use of a large computer such as the Remington Rand Univac and are complex and expensive enough to warrant a five-stage program: 1. In the early stages of design, only preliminary two-dimensional calculations were performed .in order to obtain a rough idea of the general behavior of the reactor and its critical mass with tentative experiments in place. 2. A large amount of work was carried out in which the reactor was approximated as one with a uniform homogeneous core. With this model, detailed studies were carried out to investigate the feasibility and to obtain general design data on such points as the design and properties of the gray and black-control rods, the design of the beryllium reflector, gamma and neutron heating, the use of burnable poisons, etc. In performing these calculations, use was made of the IBM 650 PROD code obtained from KAPL. 3. With stages 1 and 2 carried out, two-dimensional calculations of the core at start-up conditions were performed on the Univac computer. 4. Detailed two-dimensional calculations of the properties of the ETR with a proposed first set of experiments in place were carried out. 5. A series of nuclear tests were performed at the reactivity measurements facility at the MTR site in order to confirm the validity of the analytical techniques in physics analysis. In performing the two-dimensional Univac calculations, the MUG code developed by KAPL and the Cuthill code developed at the David Taylor Model Basin were utilized. In

  14. Advanced Test Reactor National Scientific User Facility Progress

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Todd R. Allen; James I. Cole; Jeff B. Benson; Mary Catherine Thelen

    2012-10-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is one of the world’s premier test reactors for studying the effects of intense neutron radiation on reactor materials and fuels. The ATR began operation in 1967, and has operated continuously since then, averaging approximately 250 operating days per year. The combination of high flux, large test volumes, and multiple experiment configuration options provide unique testing opportunities for nuclear fuels and material researchers. The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected highly-enriched uranium fueled, reactor with a maximum operating power of 250 MWth. The ATR peak thermal flux can reach 1.0 x1015 n/cm2-sec, and the core configuration creates five main reactor power lobes (regions) that can be operated at different powers during the same operating cycle. In addition to these nine flux traps there are 68 irradiation positions in the reactor core reflector tank. The test positions range from 0.5” to 5.0” in diameter and are all 48” in length, the active length of the fuel. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. Goals of the ATR NSUF are to define the cutting edge of nuclear technology research in high temperature and radiation environments, contribute to improved industry performance of current and future light water reactors, and stimulate cooperative research between user groups conducting basic and applied research. The ATR NSUF has developed partnerships with other universities and national laboratories to enable ATR NSUF researchers to perform research at these other facilities, when the research objectives

  15. Conceptual design study of a scyllac fusion test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Thomassen, K.I. (comp.)

    1975-07-01

    The report describes a conceptual design study of a fusion test reactor based on the Scyllac toroidal theta-pinch approach to fusion. It is not the first attempt to describe the physics and technology required for demonstrating scientific feasibility of the approach, but it is the most complete design in the sense that the physics necessary to achieve the device goals is extrapolated from experimentally tested MHD theories of toroidal systems,and it uses technological systems whose engineering performance has been carefully calculated to ensure that they meet the machine requirements.

  16. Archaeology, Geomorphology and Historic Surveys in Pools 13-14, Upper Mississippi River. Volume 1

    Science.gov (United States)

    1989-09-01

    River and west of the Mississippi) was part of the Spanish province of Luisiana, officially a dependency of Cuba . Spanish colonial policy during this...governor at 3t. Louis, including land grants in what Is now Iowa. The Spanish even tried to promote emigraCion from the United States to Louis;,ana, w!th

  17. Electroosmotic dewatering of chalk sludge, iron hydroxide sludge, wet fly ash and biomass sludge

    DEFF Research Database (Denmark)

    Hansen, H.K.; Christensen, Iben Vernegren; Ottosen, Lisbeth M.;

    2003-01-01

    Electroosmotic dewatering has been tested in laboratory cells on four different porous materials: chalk sludge, iron hydroxide sludge, wet fly ash and biomass sludge from enzyme production. In all cases it was possible to remove water when passing electric DC current through the material. Casagra......Electroosmotic dewatering has been tested in laboratory cells on four different porous materials: chalk sludge, iron hydroxide sludge, wet fly ash and biomass sludge from enzyme production. In all cases it was possible to remove water when passing electric DC current through the material....... Casagrande's coefficients were determined for the four materials at different water contents. The experiments in this work showed that chalk could be dewatered from 40% to 79% DM (dry matter), fly ash from 75 to 82% DM, iron hydroxide sludge from 2.7 to 19% DM and biomass from 3 to 33% DM by electroosmosis...

  18. An Experimental Test of Buffer Utility as a Technique for Managing Pool-Breeding Amphibians.

    Science.gov (United States)

    Powell, Jessica S Veysey; Babbitt, Kimberly J

    2015-01-01

    Vegetated buffers are used extensively to manage wetland-dependent wildlife. Despite widespread application, buffer utility has not been experimentally validated for most species. To address this gap, we conducted a six-year, landscape-scale experiment, testing how buffers of different widths affect the demographic structure of two amphibian species at 11 ephemeral pools in a working forest of the northeastern U.S. We randomly assigned each pool to one of three treatments (i.e., reference, 100m buffer, 30m buffer) and clearcut to create buffers. We captured all spotted salamanders and wood frogs breeding in each pool and examined the impacts of treatment and hydroperiod on breeding-population abundance, sex ratio, and recapture rate. The negative effects of clearcutting tended to increase as forest-buffer width decreased and be strongest for salamanders and when other stressors were present (e.g., at short-hydroperiod pools). Recapture rates were reduced in the 30m, but not 100m, treatment. Throughout the experiment for frogs, and during the first year post-cut for salamanders, the predicted mean proportion of recaptured adults in the 30m treatment was only 62% and 40%, respectively, of that in the reference treatment. Frog sex ratio and abundance did not differ across treatments, but salamander sex ratios were increasingly male-biased in both cut treatments. By the final year, there were on average, only about 40% and 65% as many females predicted in the 100m and 30m treatments, respectively, compared to the first year. Breeding salamanders at short-hydroperiod pools were about 10% as abundant in the 100m versus reference treatment. Our study demonstrates that buffers partially mitigate the impacts of habitat disturbance on wetland-dependent amphibians, but buffer width and hydroperiod critically mediate that process. We provide the first experimental evidence showing that 30-m-wide buffers may be insufficient for maintaining resilient breeding populations of pool

  19. An Experimental Test of Buffer Utility as a Technique for Managing Pool-Breeding Amphibians.

    Directory of Open Access Journals (Sweden)

    Jessica S Veysey Powell

    Full Text Available Vegetated buffers are used extensively to manage wetland-dependent wildlife. Despite widespread application, buffer utility has not been experimentally validated for most species. To address this gap, we conducted a six-year, landscape-scale experiment, testing how buffers of different widths affect the demographic structure of two amphibian species at 11 ephemeral pools in a working forest of the northeastern U.S. We randomly assigned each pool to one of three treatments (i.e., reference, 100m buffer, 30m buffer and clearcut to create buffers. We captured all spotted salamanders and wood frogs breeding in each pool and examined the impacts of treatment and hydroperiod on breeding-population abundance, sex ratio, and recapture rate. The negative effects of clearcutting tended to increase as forest-buffer width decreased and be strongest for salamanders and when other stressors were present (e.g., at short-hydroperiod pools. Recapture rates were reduced in the 30m, but not 100m, treatment. Throughout the experiment for frogs, and during the first year post-cut for salamanders, the predicted mean proportion of recaptured adults in the 30m treatment was only 62% and 40%, respectively, of that in the reference treatment. Frog sex ratio and abundance did not differ across treatments, but salamander sex ratios were increasingly male-biased in both cut treatments. By the final year, there were on average, only about 40% and 65% as many females predicted in the 100m and 30m treatments, respectively, compared to the first year. Breeding salamanders at short-hydroperiod pools were about 10% as abundant in the 100m versus reference treatment. Our study demonstrates that buffers partially mitigate the impacts of habitat disturbance on wetland-dependent amphibians, but buffer width and hydroperiod critically mediate that process. We provide the first experimental evidence showing that 30-m-wide buffers may be insufficient for maintaining resilient breeding

  20. Adsorption Properties of Chalk Reservoir Materials

    DEFF Research Database (Denmark)

    Okhrimenko, Denis

    Understanding adsorption energetics and wetting properties of calcium carbonate surfaces is essential for developing remediation strategies for aquifers, improving oil recovery, minimising risk in CO2 storage and optimising industrial processes. This PhD was focussed on comparing the vapour....../gas adsorption properties of synthetic calcium carbonate phases (calcite, vaterite and aragonite) with chalk, which is composed of biogenic calcite (>98%). In combination with data from nanotechniques, the results demonstrate the complexity of chalk behavior and the role of nanoscale clay particles. The results...

  1. Elastic behaviour of North Sea chalk

    DEFF Research Database (Denmark)

    Gommesen, Lars; Fabricius, Ida Lykke; Mukerji, T.

    2007-01-01

    We present two different elastic models for, respectively, cemented and uncemented North Sea chalk well-log data. We find that low Biot coefficients correlate with anomalously low cementation factors from resistivity measurements at low porosity and we interpret this as an indication of cementation...... to logging data than the Gassmann prediction for the far, virgin zone. We thus conclude that the Gassmann approach predicts hydrocarbons accurately in chalk in the sonic-frequency domain, but the fluid effects as recorded by the acoustic tool are significantly affected by invasion of mud filtrate...

  2. Reactor Testing and Qualification: Prioritized High-level Criticality Testing Needs

    Energy Technology Data Exchange (ETDEWEB)

    S. Bragg-Sitton; J. Bess; J. Werner; G. Harms; S. Bailey

    2011-09-01

    Researchers at the Idaho National Laboratory (INL) were tasked with reviewing possible criticality testing needs to support development of the fission surface power system reactor design. Reactor physics testing can provide significant information to aid in development of technologies associated with small, fast spectrum reactors that could be applied for non-terrestrial power systems, leading to eventual system qualification. Several studies have been conducted in recent years to assess the data and analyses required to design and build a space fission power system with high confidence that the system will perform as designed [Marcille, 2004a, 2004b; Weaver, 2007; Parry et al., 2008]. This report will provide a summary of previous critical tests and physics measurements that are potentially applicable to the current reactor design (both those that have been benchmarked and those not yet benchmarked), summarize recent studies of potential nuclear testing needs for space reactor development and their applicability to the current baseline fission surface power (FSP) system design, and provide an overview of a suite of tests (separate effects, sub-critical or critical) that could fill in the information database to improve the accuracy of physics modeling efforts as the FSP design is refined. Some recommendations for tasks that could be completed in the near term are also included. Specific recommendations on critical test configurations will be reserved until after the sensitivity analyses being conducted by Los Alamos National Laboratory (LANL) are completed (due August 2011).

  3. Local stability tests in Dresden 2 boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    March-Leuba, J.; Fry, D.N.; Buchanan, M.E.; McNew, C.O.

    1984-04-01

    This report presents the results of a local stability test performed at Dresden Unit 2 in May 1983 to determine the effect of a new fuel element design on local channel stability. This test was performed because the diameter of the new fuel rods increases the heat transfer coefficient, making the reactor more responsive and, thus, more susceptible to instabilities. After four of the new fuel elements with a 9 x 9 array of fuel rods were loaded into Dresden 2, the test was performed by inserting an adjacent control rod all the way in and then withdrawing it to its original position at maximum speed. At the moment of the test, reactor conditions were 52.7% power and 38.9% flow. Both the new 9 x 9 fuel elements and the standard 8 x 8 ones proved to be locally stable when operating at minimum pump speed at the beginning of cycle in Dresden 2, and no significant difference was found between the behavior of the two fuel types. Finally, Dresden 2 showed a high degree of stability during control rod and normal noise type perturbations.

  4. Archaeological Investigations, Navigation Pool II, Upper Mississippi River Basin. Volume 1. Narrative

    Science.gov (United States)

    1985-03-01

    unable to identify the precise depth of the pre-settlement ( pie -1850) surface. As an example, at the Turkey River Public Use Area, post-settlement...r-constructions of Late Paleoindian lifeways have been ,1v,-loprd for this region and Quimby’s (1960) Aqua- Plano tradition and 74ason’s (1963) Late...explicit frame of reference, and Quimby has characterized them as dominated by the "Aqua- Plano " tradition (1960: 34- 42). For reasons propounded

  5. Present status and future perspectives of research and test reactor in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kaneko, Yoshihiko [Atomic Energy Research Laboratory, Musashi Institute of Technology, Kawasaki, Kanagawa (Japan); Kaieda, Keisuke [Department of Research Reactor, Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    2000-10-01

    Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfill a major role in the study of nuclear energy and fundamental research. At present four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR) are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has recently reached first criticality and now in the power up test. In 1966, the Kyoto University built the Kyoto University Reactor (KUR) and started its operation for joint use program of the Japanese universities. This paper introduces these reactors and describes their present operational status and also efforts for aging management. The recent tendency of utilization and future perspectives is also reported. (author)

  6. High temperature indentation tests on fusion reactor candidate materials

    Energy Technology Data Exchange (ETDEWEB)

    Montanari, R. [Dipartimento di Ingegneria Meccanica, Universita di Roma-Tor Vergata, Via del Politecnico 1, I-00133 Rome (Italy)]. E-mail: roberto.montanari@uniroma2.it; Filacchioni, G. [ENEA CR Casaccia, Via Anguillarese 301, I-00060 S.M. di Galeria, Rome (Italy); Iacovone, B. [Dipartimento di Ingegneria Meccanica, Universita di Roma-Tor Vergata, Via del Politecnico 1, I-00133 Rome (Italy); Plini, P. [Dipartimento di Ingegneria Meccanica, Universita di Roma-Tor Vergata, Via del Politecnico 1, I-00133 Rome (Italy); Riccardi, B. [Associazione EURATOM-ENEA sulla Fusione, P.O. Box 65, I-00044 Frascati, Rome (Italy)

    2007-08-01

    Flat-top cylinder indenter for mechanical characterization (FIMEC) is an indentation technique employing cylindrical punches with diameters ranging from 0.5 to 2 mm. The test gives pressure-penetration curves from which the yield stress can be determined. The FIMEC apparatus was developed to test materials in the temperature range from -180 to +200 {sup o}C. Recently, the heating system of FIMEC apparatus has been modified to operate up to 500 {sup o}C. So, in addition to providing yield stress over a more extended temperature range, it is possible to perform stress-relaxation tests at temperatures of great interest for several nuclear fusion reactor (NFR) alloys. Data on MANET-II, F82H mod., Eurofer-97, EM-10, AISI 316 L, Ti6Al4V and CuCrZr are presented and compared with those obtained by mechanical tests with standard methods.

  7. Proposal of novel method of continuous monitoring of possible fuel failure of a pool-type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sasaki, K. [Rikkyo University, Nishi-Ikebukuro, Toshima-ku, Tokyo (Japan). College of Science; Hayashi, S.A.; Matsura, T. [Rikkyo University, Nagasaka, Yokosuka (Japan). Institute for Atomic Energy

    1997-10-01

    During the course of studies on fuel failure detection, we have found that the bubbling of a gas such as nitrogen into a reactor coolant water effectively purges the dissolved fission rare gases ({sup 89}Kr, T{sub 1/2}=3.15 min, and {sup 138}Xe, T{sub 1/2}=14.08 min) and that the respective daughter nuclides ({sup 89}Rb, T{sub 1/2}=15.15 min and {sup 138}Cs, T{sub 1/2}=33.41 min) are detected in the washing water of the collected gas mixture. The detected activity depends on the time of standing between sampling and washing of the gas, and the dependence agreed well with the theoretical prediction from the consecutive radioactive decay for both pairs ({sup 89}Kr-{sup 89}Rb, and {sup 138}Xe-{sup 138}Cs). Based on these findings, we have recently constructed a semi-continuous fuel monitoring system, which consists of an automatic and intermittent gas sampler (1 litre bottles) and a bottle conveying unit. After standing for a definite time, bottled gas is shaken with a small amount of water, and the activity of the water is measured. This system operates satisfactorily, but the whole system involves several sophisticated steps so that is rather costly. Quite recently we have got an idea of a simpler, more economical, fully automated continuous system. The system consists in principle only of a large cylinder with packing materials just as in a fractional distiller. On the top of the cylinder there are an inlet of washing water and an outlet of the gas, and at the bottom there are an inlet of the collected gas from the coolant and an outlet of the washing water. The whole system can be operated fully automatically and continuously, with continuous feeding of bubbling gas into the reactor coolant. This has not yet been experimentally tested at present, and in this presentation, information about the setup parameters such as the flow rate of the bubbling gas, the volume of the cylinder and vacant space, the flow rate of the washing water, etc. are reported

  8. Geologic setting of the New Production Reactor within the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Price, V. [Westinghouse Savannah River Co., Aiken, SC (United States); Fallaw, W.C. [Furman Univ., Greenville, SC (United States). Dept. of Geology; McKinney, J.B. [Exploration Resources, Inc., Athens, GA (United States)

    1991-12-31

    The geology and hydrology of the reference New Production Reactor (NPR) site at Savannah River Site (SRS) have been summarized using the available information from the NPR site and areas adjacent to the site, particularly the away from reactor spent fuel storage site (AFR site). Lithologic and geophysical logs from wells drilled near the NPR site do not indicate any faults in the upper several hundred feet of the Coastal Plain sediments. However, the Pen Branch Fault is located about 1 mile south of the site and extends into the upper 100 ft of the Coastal Plain sequence. Subsurface voids, resulting from the dissolution of calcareous portions of the sediments, may be present within 200 ft of the surface at the NPR site. The water table is located within 30 to 70 ft of the surface. The NPR site is located on a groundwater divide, and groundwater flow for the shallowest hydraulic zones is predominantly toward local streams. Groundwater flow in deeper Tertiary sediments is north to Upper Three Runs Creek or west to the Savannah River Swamp. Groundwater flow in the Cretaceous sediments is west to the Savannah River.

  9. Completion summary for borehole USGS 136 near the Advanced Test Reactor Complex, Idaho National Laboratory, Idaho

    Science.gov (United States)

    Twining, Brian V.; Bartholomay, Roy C.; Hodges, Mary K.V.

    2012-01-01

    In 2011, the U.S. Geological Survey, in cooperation with the U.S. Department of Energy, cored and completed borehole USGS 136 for stratigraphic framework analyses and long-term groundwater monitoring of the eastern Snake River Plain aquifer at the Idaho National Laboratory. The borehole was initially cored to a depth of 1,048 feet (ft) below land surface (BLS) to collect core, open-borehole water samples, and geophysical data. After these data were collected, borehole USGS 136 was cemented and backfilled between 560 and 1,048 ft BLS. The final construction of borehole USGS 136 required that the borehole be reamed to allow for installation of 6-inch (in.) diameter carbon-steel casing and 5-in. diameter stainless-steel screen; the screened monitoring interval was completed between 500 and 551 ft BLS. A dedicated pump and water-level access line were placed to allow for aquifer testing, for collecting periodic water samples, and for measuring water levels. Geophysical and borehole video logs were collected after coring and after the completion of the monitor well. Geophysical logs were examined in conjunction with the borehole core to describe borehole lithology and to identify primary flow paths for groundwater, which occur in intervals of fractured and vesicular basalt. A single-well aquifer test was used to define hydraulic characteristics for borehole USGS 136 in the eastern Snake River Plain aquifer. Specific-capacity, transmissivity, and hydraulic conductivity from the aquifer test were at least 975 gallons per minute per foot, 1.4 × 105 feet squared per day (ft2/d), and 254 feet per day, respectively. The amount of measureable drawdown during the aquifer test was about 0.02 ft. The transmissivity for borehole USGS 136 was in the range of values determined from previous aquifer tests conducted in other wells near the Advanced Test Reactor Complex: 9.5 × 103 to 1.9 × 105 ft2/d. Water samples were analyzed for cations, anions, metals, nutrients, total organic

  10. Action Memorandum for the Engineering Test Reactor under the Idaho Cleanup Project

    Energy Technology Data Exchange (ETDEWEB)

    A. B. Culp

    2007-01-26

    This Action Memorandum documents the selected alternative for decommissioning of the Engineering Test Reactor at the Idaho National Laboratory under the Idaho Cleanup Project. Since the missions of the Engineering Test Reactor Complex have been completed, an engineering evaluation/cost analysis that evaluated alternatives to accomplish the decommissioning of the Engineering Test Reactor Complex was prepared adn released for public comment. The scope of this Action Memorandum is to encompass the final end state of the Complex and disposal of the Engineering Test Reactor vessol. The selected removal action includes removing and disposing of the vessel at the Idaho CERCLA Disposal Facility and demolishing the reactor building to ground surface.

  11. Safety Analyses at the Idaho National Engineering and Environmental Laboratory Test Reactor Area - Past to Present

    Energy Technology Data Exchange (ETDEWEB)

    Ambrosek, Richard Garry; Ingram, Frederick William

    1999-11-01

    Test reactors are unique in that the core configuration may change with each operating interval. The process of safety analyses for test reactors at the Idaho National Engineering and Environmental Test Reactor Area has evolved as the computing capabilities, software, and regulatory requirements have changed. The evaluations for experiments and the reactor have moved from measurements in a set configuration and then application to other configurations with a relatively large error to modeling in three-dimensions and explicit analyses for each experiment and operating interval. This evolution is briefly discussed for the Test Reactor Area.

  12. Alpha Particle Physics Experiments in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Budny, R.V.; Darrow, D.S.; Medley, S.S.; Nazikian, R.; Zweben, S.J.; et al.

    1998-12-14

    Alpha particle physics experiments were done on the Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium (DT) run from 1993-1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single-particle confinement model in magnetohydrodynamic (MHD) quiescent discharges. Also, the observed alpha particle interactions with sawteeth, toroidal Alfvén eigenmodes (TAE), and ion cyclotron resonant frequency (ICRF) waves were roughly consistent with theoretical modeling. This paper reviews what was learned and identifies what remains to be understood.

  13. EPR/PTFE dosimetry for test reactor environments

    Energy Technology Data Exchange (ETDEWEB)

    Vehar, D.W.; Griffin, P.J.; Quirk, T.J. [Sandia National Laboratories, Albuquerque, NM 87185-1146 (United States)

    2011-07-01

    The use of Electron Paramagnetic Resonance (EPR) spectroscopy with materials such as alanine is well established as a technique for measurement of ionizing radiation absorbed dose in photon and electron fields such as Co-60, high-energy bremsstrahlung and electron-beam fields [1]. In fact, EPR/Alanine dosimetry has become a routine transfer standard for national standards bodies such as NIST and NPL. In 1992 the Radiation Metrology Laboratory (RML) at Sandia National Laboratories implemented EPR/Alanine capabilities for use in routine and calibration activities at its Co-60 and pulsed-power facilities. At that time it also investigated the usefulness of the system for measurement of absorbed dose in the mixed neutron/photon environments of reactors such as the Sandia Pulsed Reactor and the Annular Core Research Reactor used for hardness testing of electronics. The RML concluded that the neutron response of alanine was a sufficiently high fraction of the overall dosimeter response that the resulting uncertainties in the photon dose would be unacceptably large for silicon-device testing. However, it also suggested that non-hydrogenous materials such as polytetrafluoroethylene (PTFE) would exhibit smaller neutron response and might be useful in mixed environments. Preliminary research with PTFE in photon environments indicated considerable promise, but further development was not pursued at that time. Because of renewed interest in absorbed dose measurements that could better define the individual contributions of photon and neutron components to the overall dose delivered to a test object, the RML has re-initiated the development of an EPR/PTFE dosimetry system. This effort consists of three stages: 1) Identification of PTFE materials that may be suitable for dosimetry applications. It was speculated that the inconsistency of EPR signatures in the earlier samples may have been due to variability in PTFE manufacturing processes. 2) Characterization of dosimetry in

  14. Population dynamics of three gammarid species (Crustacea, Amphipoda) in a French chalk stream. Part II. Standing crop

    NARCIS (Netherlands)

    Goedemakers, Annemarie

    1981-01-01

    The standing crop of Gammarus pulex pulex (Linnaeus, 1758), G. fossarum Koch in Panzer, 1836 and Echinogammarus berilloni (Catta, 1878) has been studied in a small French chalk stream, the Slack. A brief description of all amphipod species encountered in this river is given, with a key to different

  15. LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, William Jonathan [Idaho National Laboratory; Barrett, Kristine Eloise [Idaho National Laboratory; Chichester, Heather Jean MacLean [Idaho National Laboratory

    2015-09-01

    The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirements for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.

  16. Core-concrete interactions with overlying water pools. The WETCOR-1 test

    Energy Technology Data Exchange (ETDEWEB)

    Blose, R.E. [Ktech Corp., Albuquerque, NM (United States); Powers, D.A.; Copus, E.R.; Brockmann, J.E.; Simpson, R.B.; Lucero, D.A. [Sandia National Labs., Albuquerque, NM (United States)

    1993-11-01

    The WETCOR-1 test of simultaneous interactions of a high-temperature melt with water and a limestone/common-sand concrete is described. The test used a 34.1-kg melt of 76.8 w/o Al{sub 2}O{sub 3}, 16.9 w/o CaO, and 4.0 w/o SiO{sub 2} heated by induction using tungsten susceptors. Once quasi-steady attack on concrete by the melt was established, an attempt was made to quench the melt at 1850 K with 295 K water flowing at 57 liters per minute. Net power into the melt at the time of water addition was 0.61 {plus_minus} 0.19 W/cm{sup 3}. The test configuration used in the WETCOR-1 test was designed to delay melt freezing to the walls of the test fixture. This was done to test hypotheses concerning the inherent stability of crust formation when high-temperature melts are exposed to water. No instability in crust formation was observed. The flux of heat through the crust to the water pool maintained over the melt in the test was found to be 0.52 {plus_minus} 0.13 MW/m{sup 2}. Solidified crusts were found to attenuate aerosol emissions during the melt concrete interactions by factors of 1.3 to 3.5. The combination of a solidified crust and a 30-cm deep subcooled water pool was found to attenuate aerosol emissions by factors of 3 to 15.

  17. Strength and Biot's coefficient for high-porosity oil- or water-saturated chalk

    DEFF Research Database (Denmark)

    Andreassen, Katrine Alling

    . The Biot coefficient states the degree of cementation or how the pore pressure contributes to the strain resulting from an external load for a porous material. It is here calculated from dynamic measurements and correlated with the strength of outcrop chalk characterized by the onset of pore collapse...... during hydrostatic loading. The hypothesis is that the Biot coefficient and the theory of poroelasticity may cover the fluid effect by including the increased fluid bulk modulus from oil to water. A high number of test results for both oil- and water-saturated high-porosity outcrop chalk show correlation...

  18. Nickel adsorption on chalk and calcite

    DEFF Research Database (Denmark)

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

    2014-01-01

    and chalk and pH ranging from 7.7 to 8.8, explored the influence of initial concentration and the amount and type of sorbent on Ni uptake. Adsorption increases with increased surface area and pH. A surface complexation model describes the data well. Stability constants for the Ni surface complex are log KNi...

  19. Late Maastrichtian chalk mounds, Stevns Klint, Denmark

    DEFF Research Database (Denmark)

    Anderskouv, Kresten; Damholt, Tove; Surlyk, Finn

    Upper Maastrichtian chalk exposed in the Sigerslev quarry, Stevns Klint, Denmark shows wavy and mound-like bedding geometries outlined by bands of black flint nodules. Bedding geometries are highly variable, but four morphological elements are recognized: Southward migrating mounds, eastward...

  20. Late Maastrichtian chalk mounds, Stevns Klint, Denmark

    DEFF Research Database (Denmark)

    Anderskouv, Kresten; Damholt, Tove; Surlyk, Finn

    2007-01-01

    Upper Maastrichtian chalk exposed at the Sigerslev quarry, Stevns Klint, Denmark is characterized by wavy and mound-like bedding geometries outlined by bands of black flint nodules. Four morphological elements are recognized, although bedding geometries are highly variable: southward migrating...

  1. Slope failure of chalk channel margins

    DEFF Research Database (Denmark)

    Gale, A.; Anderskouv, Kresten; Surlyk, Finn

    2015-01-01

    The importance of mass transport and bottom currents is now widely recognized in the Upper Cretaceous Chalk Group of Northern Europe. The detailed dynamics and interaction of the two phenomena are difficult to study as most evidence is based on seismic data and drill core. Here, field observation...

  2. Sodium reflux pool-boiler solar receiver on-sun test results

    Energy Technology Data Exchange (ETDEWEB)

    Andraka, C E; Moreno, J B; Diver, R B; Moss, T A [Oak Ridge National Lab., TN (United States)

    1992-06-01

    The efficient operation of a Stirling engine requires the application of a high heat flux to the relatively small area occupied by the heater head tubes. Previous attempts to couple solar energy to Stirling engines generally involved directly illuminating the heater head tubes with concentrated sunlight. In this study, operation of a 75-kW{sub t} sodium reflux pool-boiler solar receiver has been demonstrated and its performance characterized on Sandia's nominal 75-kW{sub t} parabolic-dish concentrator, using a cold-water gas-gap calorimeter to simulate Stirling engine operation. The pool boiler (and more generally liquid-metal reflux receivers) supplies heat to the engine in the form of latent heat released from condensation of the metal vapor on the heater head tubes. The advantages of the pool boiler include uniform tube temperature, leading to longer life and higher temperature available to the engine, and decoupling of the design of the solar absorber from the engine heater head. The two-phase system allows high input thermal flux, reducing the receiver size and losses, therefore improving system efficiency. The receiver thermal efficiency was about 90% when operated at full power and 800{degree}C. Stable sodium boiling was promoted by the addition of 35 equally spaced artificial cavities in the wetted absorber surface. High incipient boiling superheats following cloud transients were suppressed passively by the addition of small amounts of xenon gas to the receiver volume. Stable boiling without excessive incipient boiling superheats was observed under all operating conditions. The receiver developed a leak during performance evaluation, terminating the testing after accumulating about 50 hours on sun. The receiver design is reported here along with test results including transient operations, steady-state performance evaluation, operation at various temperatures, infrared thermography, x-ray studies of the boiling behavior, and a postmortem analysis.

  3. Meso-scale modeling of irradiated concrete in test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Giorla, A. [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Vaitová, M. [Czech Technical University, Thakurova 7, 166 29 Praha 6 (Czech Republic); Le Pape, Y., E-mail: lepapeym@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Štemberk, P. [Czech Technical University, Thakurova 7, 166 29 Praha 6 (Czech Republic)

    2015-12-15

    Highlights: • A meso-scale finite element model for irradiated concrete is developed. • Neutron radiation-induced volumetric expansion is a predominant degradation mode. • Confrontation with expansion and damage obtained from experiments is successful. • Effects of paste shrinkage, creep and ductility are discussed. - Abstract: A numerical model accounting for the effects of neutron irradiation on concrete at the mesoscale is detailed in this paper. Irradiation experiments in test reactor (Elleuch et al., 1972), i.e., in accelerated conditions, are simulated. Concrete is considered as a two-phase material made of elastic inclusions (aggregate) subjected to thermal and irradiation-induced swelling and embedded in a cementitious matrix subjected to shrinkage and thermal expansion. The role of the hardened cement paste in the post-peak regime (brittle-ductile transition with decreasing loading rate), and creep effects are investigated. Radiation-induced volumetric expansion (RIVE) of the aggregate cause the development and propagation of damage around the aggregate which further develops in bridging cracks across the hardened cement paste between the individual aggregate particles. The development of damage is aggravated when shrinkage occurs simultaneously with RIVE during the irradiation experiment. The post-irradiation expansion derived from the simulation is well correlated with the experimental data and, the obtained damage levels are fully consistent with previous estimations based on a micromechanical interpretation of the experimental post-irradiation elastic properties (Le Pape et al., 2015). The proposed modeling opens new perspectives for the interpretation of test reactor experiments in regards to the actual operation of light water reactors.

  4. ASME N510 test results for Savannah River Site AACS filter compartments

    Energy Technology Data Exchange (ETDEWEB)

    Paul, J.D.; Punch, T.M. [Westinghouse Savannah River Company, Aiken, SC (United States)

    1995-02-01

    The K-Reactor at the Savannah River Site recently implemented design improvements for the Airborne Activity Confinement System (AACS) by procuring, installing, and testing new Air Cleaning Units, or filter compartments, to ASME AG-11, N509, and N510 requirements. Specifically, these new units provide documentable seismic resistance to a Design Basis Accident earthquake, provide 2 inch adsorber beds with 0.25 second residence time, and meet all AG-1, N509, and N510 requirements for testability and maintainability. This paper presents the results of the Site acceptance testing and discusses an issue associated with sample manifold qualification testing.

  5. Statistical model specification and power: recommendations on the use of test-qualified pooling in analysis of experimental data.

    Science.gov (United States)

    Colegrave, Nick; Ruxton, Graeme D

    2017-03-29

    A common approach to the analysis of experimental data across much of the biological sciences is test-qualified pooling. Here non-significant terms are dropped from a statistical model, effectively pooling the variation associated with each removed term with the error term used to test hypotheses (or estimate effect sizes). This pooling is only carried out if statistical testing on the basis of applying that data to a previous more complicated model provides motivation for this model simplification; hence the pooling is test-qualified. In pooling, the researcher increases the degrees of freedom of the error term with the aim of increasing statistical power to test their hypotheses of interest. Despite this approach being widely adopted and explicitly recommended by some of the most widely cited statistical textbooks aimed at biologists, here we argue that (except in highly specialized circumstances that we can identify) the hoped-for improvement in statistical power will be small or non-existent, and there is likely to be much reduced reliability of the statistical procedures through deviation of type I error rates from nominal levels. We thus call for greatly reduced use of test-qualified pooling across experimental biology, more careful justification of any use that continues, and a different philosophy for initial selection of statistical models in the light of this change in procedure.

  6. Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

    2009-06-15

    Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

  7. Emission and transmission tomography systems to be developed for the future needs of Jules Horowitz material testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kotiluoto, Petri [VTT Technical Research Centre of Finland, P.O.Box 1000, FI-02044 VTT (Finland)], E-mail: petri.kotiluoto@vtt.fi; Wasastjerna, Frej; Kekki, Tommi [VTT Technical Research Centre of Finland, P.O.Box 1000, FI-02044 VTT (Finland); Sipilae, Heikki; Banzuzi, Kukka [Oxford Instruments Analytical Oy, Nihtisillankuja 5, P.O.Box 85, FI-02631 Espoo (Finland); Kinnunen, Petri; Heikinheimo, Liisa [VTT Technical Research Centre of Finland, P.O.Box 1000, FI-02044 VTT (Finland)

    2009-08-01

    The new 100 MW Jules Horowitz material testing reactor will be built in Cadarache, France. It will support, for instance, research on new types of innovative nuclear fuel. As a Finnish in-kind contribution, 3D emission and transmission tomography equipment will be delivered for both the reactor and the active component storage pool. The image reconstruction of activities inside the used nuclear fuel will be based on gamma spectrometry measurements. A new type of underwater digital X-ray linear detector array is under development for transmission imaging, based on GaAs and direct conversion of X-rays into an electrical signal. A shared collimator will be used for both emission and transmission measurements. Some preliminary design has been performed. For the current design, the expected gamma spectrometric response of a typical high-purity germanium detector has been simulated with MCNP for minimum and maximum source activities (specified by CEA) to be measured in future.

  8. Scaled Facility Design Approach for Pool-Type Lead-Bismuth Eutectic Cooled Small Modular Reactor Utilizing Natural Circulation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sangrok; Shin, Yong-Hoon; Lee, Jueun; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

    2015-10-15

    In low carbon era, nuclear energy is the most prominent energy source of electricity. For steady ecofriendly nuclear energy supply, Generation IV reactors which are future nuclear reactor require safety, sustainability, economics and non-proliferation as four criteria. Lead cooled fast reactor (LFR) is one of these reactor type and Generation IV international forum (GIF) adapted three reference LFR systems which are a small and movable systems with long life without refueling, intermediate size and huge electricity generation system for power grid. NUTRECK (Nuclear Transmutation Energy Center of Korea) has been designed reactor called URANUS (Ubiquitous, Rugged, Accident-forgiving, Non-proliferating, and Ultra-lasting Sustainer) which is small modular reactor and using lead-bismuth eutectic coolant. To prove natural circulation capability of URANUS and analyze design based accidents, scaling mock-up experiment facility will be constructed. In this paper, simple specifications of URANUS will be presented. Then based on this feature, scaling law and scaled facility design results are presented. To validate safety feature and thermodynamics characteristic of URANUS, scaled mockup facility of URANUS is designed based on the scaling law. This mockup adapts two area scale factors, core and lower parts of mock-up are scaled for 3D flow experiment. Upper parts are scaled different size to reduce electricity power and LBE tonnage. This hybrid scaling method could distort some thermal-hydraulic parameters, however, key parameters for experiment will be matched for up-scaling. Detailed design of mock-up will be determined through iteration for design optimization.

  9. VISTA : thermal-hydraulic integral test facility for SMART reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, K. Y.; Park, H. S.; Cho, S.; Park, C. K.; Lee, S. J.; Song, C. H.; Chung, M. K. [KAERI, Taejon (Korea, Republic of)

    2003-07-01

    Preliminary performance tests were carried out using the thermal-hydraulic integral test facility, VISTA (Experimental Verification by Integral Simulation of Transients and Accidents), which has been constructed to simulate the SMART-P. The VISTA facility is an integral test facility including the primary and secondary systems as well as safety-related Passive Residual Heat Removal (PRHR) systems. Its scaled ratio with respect to the SMART-P is 1/1 in height and 1/96 in volume and heater power. Several steady states and power changing tests have been carried out to verify the overall thermal hydraulic primary and secondary characteristics in the range of 10% to 100% power operation. As for the preliminary results, the steady state conditions were found to coincide with the expected design values of the SMART-P. But the major thermal hydraulic parameters are greatly affected by the initial water level and the nitrogen pressure in the reactor's upper annular cavity. The power step/ramp changing tests are successfully carried out and the system responses are observed. The primary natural circulation operation is achieved, but advanced control logics need to be developed to reach the natural circulation mode without pressure excursion. In the PRHR transient tests, the natural circulation flow rate through the PRHR system was found to be about 10 percent in the early phases of PRHR operation.

  10. Effective-stress-law behavior of Austin chalk rocks for deformation and fracture conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Warpinski, N.R.; Teufel, L.W.

    1994-08-01

    Austin chalk core has been tested to determine the effective law for deformation of the matrix material and the stress-sensitive conductivity of the natural fractures. For deformation behavior, two samples provided data on the variations of the poroelastic parameter, {alpha}, for Austin chalk, giving values around 0.4. The effective-stress-law behavior of a Saratoga limestone sample was also measured for the purpose of obtaining a comparison with a somewhat more porous carbonate rock. {alpha} for this rock was found to be near 0.9. The low {alpha} for the Austin chalk suggests that stresses in the reservoir, or around the wellbore, will not change much with changes in pore pressure, as the contribution of the fluid pressure is small. Three natural fractures from the Austin chalk were tested, but two of the fractures were very tight and probably do not contribute much to production. The third sample was highly conductive and showed some stress sensitivity with a factor of three reduction in conductivity over a net stress increase of 3000 psi. Natural fractures also showed a propensity for permanent damage when net stressed exceeded about 3000 psi. This damage was irreversible and significantly affected conductivity. {alpha} was difficult to determine and most tests were inconclusive, although the results from one sample suggested that {alpha} was near unity.

  11. ZEEP: Canada's first nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Green, R.E.; Okazaki, A. [retired, Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2015-09-15

    In 1905 Albert Einstein published his historic paper on special relativity, which contained the equation E=mc 2. The significance of this mass-energy relationship became evident with the discovery of nuclear fission in 1939, when it was realized that large amounts of energy would be released in a fission chain reaction. Canadian scientists were involved in this field from the beginning and their efforts resulted in the startup in September 1945 of the ZEEP reactor at Chalk River, the first reactor to go critical outside the USA. In this paper we recall some of the events that led to the construction of ZEEP, and describe the role it played in the development of the Canadian nuclear energy program. (author)

  12. Development and verification test of integral reactor major components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. I.; Kim, Y. W.; Kim, J. H. and others

    1999-03-01

    The conceptual designs for SG, MCP, CEDM to be installed in the integral reactor SMART were developed. Three-dimensional CAD models for the major components were developed to visualize the design concepts. Once-through helical steam generator was conceptually designed for SMART. Canned motor pump was adopted in the conceptual design of MCP. Linear pulse motor type and ballscrew type CEDM, which have fine control capabilities were studied for adoption in SMART. In parallel with the structural design, the electro-magnetic design was performed for the sizing motors and electro-magnet. Prototypes for the CEDM and MCP sub-assemblies were developed and tested to verify the performance. The impeller design procedure and the computer program to analyze the dynamic characteristics of MCP rotor shaft were developed. The design concepts of SG, MCP, CEDM were also invetigated for the fabricability.

  13. Natural radioactive materials at the Arco Reactor Test Site

    Energy Technology Data Exchange (ETDEWEB)

    Singlevich, W; Healy, J W; Paas, H J; Carey, Z E

    1951-05-28

    At the request of the Division of Biology and Medicine of the AEC, the Biophysics Section of the Radiological Sciences Department at Hanford undertook the task of conducting a background survey for naturally occurring radioactive materials in the environs of the Arco Reactor Test Site in Central Idaho. This survey was part of an overall study which included meteorological measurements by the the Air Weather Service, Geological Studies by the USGS, and an ecological survey of plants and animals by members of the Idaho State College at Pocatello. In general, the measurements at Arco followed the pattern established for environmental monitoring at the Hanford Site with some additional measurements made for natural isotopes not normally of concern at Hanford. A number of analysis included materials such as plutonium and I-131 which were carried out for the purpose of establishing analytical backgrounds for the procedures used. 20 refs., 13 figs., 11 tabs.

  14. Recent results on the RIA test in IGR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V.; Yegorova, L. [Nuclear Safety Institute, Moscow (Russian Federation)

    1997-01-01

    At the 23d WRSM meeting the data base characterizing results of VVER high burnup fuel rods tests under reactivity-initiated accident (RIA) conditions was presented. Comparison of PWR and VVER failure thresholds was given also. Additional analysis of the obtained results was being carried out during 1996. The results of analysis show that the two different failure mechanisms were observed for PWR and VVER fuel rods. Some factors which can be as the possible reasons of these differences are presented. First of them is the state of preirradiated cladding. Published test data for PWR high burnup fuel rods demonstrated that the PWR high burnup fuel rods failed at the RIA test are characterized by very high level of oxidation and hydriding for the claddings. Corresponding researches were performed at Institute of Atomic Reactors (RLAR, Dimitrovgrad, Russia) for large set of VVER high burnup fuel rods. Results of these investigations show that preirradiated commercial Zr-1%Nb claddings practically keep their initial levels of oxidation and H{sub 2} concentration. Consequently the VVER preirradiated cladding must keep the high level of mechanical properties. The second reason leading to differences between failure mechanisms for two types of high burnup fuel rods can be the test conditions. Now such kind of analysis have been performed by two methods.

  15. Chalk: composition, diagenesis and physical properties

    DEFF Research Database (Denmark)

    Fabricius, Ida Lykke

    2007-01-01

    Chalk is a sedimentary rock of unusually high homogeneity on the scale where physical properties are measured, but the properties fall in wide ranges. Chalk may thus be seen as the ideal starting point for a physical understanding of rocks in general. Properties as porosity, permeability, capillary...... entry pressure, and elastic moduli are consequences of primary sediment composition and of subsequent diagenetic history as caused by microbial action, burial stress, temperature, and pore pressure. Porosity is a main determining factor for other properties. For a given porosity, the specific surface...... of the sediment controls permeability and capillary entry pressure. As diagenesis progresses, the specific surface is less and less due to the calcite component and more and more due to the fine-grained silicates, as a reflection of the coarsening and cementation of the calcite crystals. The elastic moduli, which...

  16. Commissioning of the STAR test section for experimental simulation of loss of coolant accident using the EC-208 instrumented fuel assembly of the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maprelian, Eduardo; Torres, Walmir M.; Prado, Adelk C.; Umbehaun, Pedro E.; Franca, Renato L.; Santos, Samuel C.; Macedo, Luiz A.; Sabundjian, Gaiane, E-mail: emaprel@ipen.br, E-mail: wmtorres@ipen.br, E-mail: acprado@ipen.br, E-mail: umbehaun@ipen.br, E-mail: rlfranca@ipen.br, E-mail: samuelcs@ipen.br, E-mail: lamacedo@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SO (Brazil)

    2015-07-01

    The three basic safety functions of Research Reactors (RR) are the safe shutdown of the reactor, the proper cooling of the decay heat of the fuel elements and the confinement of radioactive materials. Compared to Nuclear Power Reactors, RR power release is small, yet its three safety functions must be met to ensure the integrity of the reactor. During a loss of coolant accident (LOCA) in pool type RR, partial or complete loss of pool water may occur, with consequent partial or complete uncovering of the fuel assemblies. In such an accident, the decay heat removal safety function must not be compromised. The Test Section for Experimental Simulation of Loss of Coolant Accident (STAR) is in commissioning phase. This test section will provide experimental data on partial and total uncovering of the EC-208 instrumented fuel assembly (IFA) irradiated in the IEA-R1. Experimental results will be useful in validation of computer codes for RR safety analysis, particularly on heat removal efficiency aspects (safety function) in accident conditions. STAR comprises a base on which is installed the IFA, the cylindrical stainless steel hull, the compressed air system for the test section emptying and refilling, and the instrumentation for temperature and level measurements. The commissioning tests or pre-operational check, consist of several preliminary tests to verify experimental procedures, the difficulties during assembling of STAR in the pool, the difficulties in control the emptying and refilling velocities, as well as, the repeatability capacity, tests of equipment, valves and systems and tests of instrumentation and data acquisition system. Safety, accuracy and easiness of operation will be checked. (author)

  17. Evaluation of the Pseudalert/Quanti-Tray MPN Test for the Rapid Enumeration of Pseudomonas aeruginosa in Swimming Pool and Spa Pool Waters.

    Science.gov (United States)

    Sartory, David P; Brewer, Megan; Beswick, Agnieszka; Steggles, Darron

    2015-12-01

    This study assessed the performance of a new most probable number test (Pseudalert/Quanti-Tray) for the enumeration of Pseudomonas aeruginosa from swimming pool and spa pool waters by comparing it to the international and national membrane filtration-based culture methods for P. aeruginosa: ISO 16266:2006 and UK The Microbiology of Drinking Water-Part 8 (MoDW Part 8) which both use Pseudomonas CN agar. The comparison was based on the calculation of mean relative differences between the two methods conducted according to ISO 17994:2014. Using both routine pool water samples (149 from 8 laboratories) and artificially contaminated samples (309 from 7 laboratories), paired counts from each sample and enumeration method were analysed. For routine samples, there were insufficient data for a conclusive assessment, but the data do indicate at least equivalent performance of Pseudalert/Quanti-Tray to the reference methods. For the artificially contaminated samples, the data also did not result in a statistically conclusive assessment but did indicate potentially better performance of Pseudalert/Quanti-Tray. Combining the data from the routine samples and artificially contaminated samples resulted in an ISO 17994 outcome that the two methods were not statistically significantly different. Thus, the Pseudalert/Quanti-Tray method is an acceptable alternative to ISO 16266 and MoDW Part 8. The Pseudalert/Quanti-Tray method has the advantage in that it does not require confirmation testing, and of providing confirmed counts within 24-28 h incubation compared to 40-48 h or longer for the ISO 16266 and MoDW Part 8 methods.

  18. Assessment of susceptibility of Type 304 stainless steel to intergranular stress corrosion cracking in simulated Savannah River Reactor environments

    Energy Technology Data Exchange (ETDEWEB)

    Ondrejcin, R.S.; Caskey, C.R. Jr.

    1989-12-01

    Intergranular stress corrosion cracking (IGSCC) of Type 304 stainless steel rate tests (CERT) of specimens machined was evaluated by constant extension from Savannah River Plant (SRP) decontaminated process water piping. Results from 12 preliminary CERT tests verified that IGSCC occurred over a wide range of simulated SRP envirorments. 73 specimens were tested in two statistical experimental designs of the central composite class. In one design, testing was done in environments containing hydrogen peroxide; in the other design, hydrogen peroxide was omitted but oxygen was added to the environment. Prediction equations relating IGSCC to temperature and environmental variables were formulated. Temperature was the most important independent variable. IGSCC was severe at 100 to 120C and a threshold temperature between 40C and 55C was identified below which IGSCC did not occur. In environments containing hydrogen peroxide, as in SRP operation, a reduction in chloride concentration from 30 to 2 ppB also significantly reduced IGSCC. Reduction in sulfate concentration from 50 to 7 ppB was effective in reducing IGSCC provided the chloride concentration was 30 ppB or less and temperature was 95C or higher. Presence of hydrogen peroxide in the environment increased IGSCC except when chloride concentration was 11 ppB or less. Actual concentrations of hydrogen peroxide, oxygen and carbon dioxide did not affect IGSCC. Large positive ECP values (+450 to +750 mV Standard Hydrogen Electrode (SHE)) in simulated SRP environments containing hydrogen peroxide and were good agreement with ECP measurements made in SRP reactors, indicating that the simulated environments are representative of SRP reactor environments. Overall CERT results suggest that the most effective method to reduce IGSCC is to reduce chloride and sulfate concentrations.

  19. 用于池式快堆系统分析的钠池三维模型开发%Development of Three-Dimensional Sodium Pool Model for System Analysis of Pool-Type Liquid Metal Fast Breeder Reactor

    Institute of Scientific and Technical Information of China (English)

    隋丹婷; 陆道纲; 张盼

    2012-01-01

    由于池式快堆钠池内的热工水力学特性对反应堆的安全运行有重要影响,本文采用基于交错网格的SIMPLE算法开发直角坐标系和柱坐标系下钠池三维计算软件.应用CFX软件进行验证之后,完成了三维流场分析程序与系统分析软件SAC-CFR的耦合,并用耦合后的程序分析日本文殊快堆45%功率稳态运行工况上腔室内的流场分布,初步验证了堆芯上腔三维化的SAC-CFR用于系统分析的有效性,为进一步开发事故模型、非能动余热排出系统模型做准备.%As the thermal-hydraulic characteristic in sodium pool is crucial for safety operation of liquid metal fast breeder reactor (LMFBR), a three-dimensional sodium pool thermal-hydraulic analysis code was developed based on SIMPLE algorithm on stagger grid under Cartesian coordinates and cylindrical coordinates. After the validation with CFX, coupling between the analysis code and SAC-CFR was completed) and then the coupled code was applied to the flow field analysis in upper plenum of Monju Plant at 45% thermal power steady-state operation condition, which preliminary shows the effectiveness of the system analysis with coupled code and makes preparations for further development of accident analysis model and passive residual heat removal system.

  20. Deterministic Modeling of the High Temperature Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ortensi, J.; Cogliati, J. J.; Pope, M. A.; Ferrer, R. M.; Ougouag, A. M.

    2010-06-01

    Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine INL’s current prismatic reactor deterministic analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 column thin annular core, and the fully loaded core critical condition with 30 columns. Special emphasis is devoted to the annular core modeling, which shares more characteristics with the NGNP base design. The DRAGON code is used in this study because it offers significant ease and versatility in modeling prismatic designs. Despite some geometric limitations, the code performs quite well compared to other lattice physics codes. DRAGON can generate transport solutions via collision probability (CP), method of characteristics (MOC), and discrete ordinates (Sn). A fine group cross section library based on the SHEM 281 energy structure is used in the DRAGON calculations. HEXPEDITE is the hexagonal z full core solver used in this study and is based on the Green’s Function solution of the transverse integrated equations. In addition, two Monte Carlo (MC) based codes, MCNP5 and PSG2/SERPENT, provide benchmarking capability for the DRAGON and the nodal diffusion solver codes. The results from this study show a consistent bias of 2–3% for the core multiplication factor. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement stems from the fact that during the experiments the

  1. Testing of an advanced thermochemical conversion reactor system

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    This report presents the results of work conducted by MTCI to verify and confirm experimentally the ability of the MTCI gasification process to effectively generate a high-quality, medium-Btu gas from a wider variety of feedstock and waste than that attainable in air-blown, direct gasification systems. The system's overall simplicity, due to the compact nature of the pulse combustor, and the high heat transfer rates attainable within the pulsating flow resonance tubes, provide a decided and near-term potential economic advantage for the MTCI indirect gasification system. The primary objective of this project was the design, construction, and testing of a Process Design Verification System for an indirectly heated, thermochemical fluid-bed reactor and a pulse combustor an an integrated system that can process alternative renewable sources of energy such as biomass, black liquor, municipal solid waste and waste hydrocarbons, including heavy oils into a useful product gas. The test objectives for the biomass portion of this program were to establish definitive performance data on biomass feedstocks covering a wide range of feedstock qualities and characteristics. The test objectives for the black liquor portion of this program were to verify the operation of the indirect gasifier on commercial black liquor containing 65 percent solids at several temperature levels and to characterize the bed carbon content, bed solids particle size and sulfur distribution as a function of gasification conditions. 6 refs., 59 figs., 29 tabs.

  2. Miniaturized Charpy test for reactor pressure vessel embrittlement characterization

    Energy Technology Data Exchange (ETDEWEB)

    Manahan, M.P. Sr. [MPM Research and Consulting, Lemont, PA (United States)

    1999-10-01

    Modifications were made to a conventional Charpy machine to accommodate the miniaturized Charpy V-Notch (MCVN) specimens which were fabricated from an archived reactor pressure vessel (RPV) steel. Over 100 dynamic MCVN tests were performed and compared to the results from conventional Charpy V-Notch (CVN) tests to demonstrate the efficacy of the miniature specimen test. The optimized sidegrooved MCVN specimens exhibit transitional fracture behavior over essentially the same temperature range as the CVN specimens which indicates that the stress fields in the MCVN specimens reasonably simulate those of the CVN specimens and this fact has been observed in finite element calculations. This result demonstrates a significant breakthrough since it is now possible to measure the ductile-brittle transition temperature (DBTT) using miniature specimens with only small correction factors, and for some materials as in the present study, without the need for any correction factor at all. This development simplifies data interpretation and will facilitate future regulatory acceptance. The non-sidegrooved specimens yield energy-temperature data which is significantly shifted downward in temperature (non-conservative) as a result of the loss of constraint which accompanies size reduction.

  3. Development of Pool-type Sodium-cooled Fast Reactor System Analysis Code%池式钠冷快堆系统分析程序开发

    Institute of Scientific and Technical Information of China (English)

    王晋; 张东辉; 胡文军

    2016-01-01

    针对池式钠冷快堆的特点,在对快堆系统的水力模型、热工模型和中子动力学模型进行详细分类和建模的基础上,利用 FORTRAN95语言开发了可用于池式钠冷快堆事故分析的系统分析程序(FASYS程序)。以中国实验快堆为计算对象对FASYS程序模型进行了初步验证,所获得的结果和试验值与其他系统程序计算值符合良好,证明了所开发的系统分析程序的正确性。%According to the characteristics of pool‐type sodium‐cooled fast reactor ,and with the fast reactor hydraulic model , thermal model and neutron kinetics model thoroughly classified and developed ,a fast reactor system analysis code (FASYS code) was developed by FORTRAN95 language for pool‐type sodium‐cooled fast reactor acci‐dent analysis .Transient conditions in CEFR were calculated with FASYS code and the results were used for code validation .The calculation results are consistent with the test data and other fast reactor system analysis code results , and the correctness of the FASYS code is proved .

  4. New insight into the microtexture of chalks from NMR analysis

    DEFF Research Database (Denmark)

    Faÿ-Gomord, Ophélie; Soete, Jeroen; Katika, Konstantina

    2016-01-01

    quality chalks independently of their sedimentological and/or diagenetic history. The study aims to develop an NMR-based approach to characterize a broad range of chalk samples. The provided laboratory low-field NMR chalk classification can be used as a guide to interpret NMR logging data...... size and T2 logarithmic (T2lm) was calculated. It is apparent that tight chalks, whether their characteristics are sedimentological or diagenetic, yield smaller pore body sizes (T2lm well as narrower pore throats (average radius

  5. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-01-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  6. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  7. 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 of Pool Heaters P Appendix P to Subpart B of Part 430 Energy DEPARTMENT OF ENERGY ENERGY CONSERVATION ENERGY CONSERVATION PROGRAM FOR CONSUMER PRODUCTS Test Procedures Pt. 430, Subpt. B, App. P Appendix P...

  8. CERCA LEU fuel assemblies testing in Maria Reactor - safety analysis summary and testing program scope.

    Energy Technology Data Exchange (ETDEWEB)

    Pytel, K.; Mieleszczenko, W.; Lechniak, J.; Moldysz, A.; Andrzejewski, K.; Kulikowska, T.; Marcinkowska, A.; Garner, P. L.; Hanan, N. A.; Nuclear Engineering Division; Institute of Atomic Energy (Poland)

    2010-03-01

    The presented paper contains neutronic and thermal-hydraulic (for steady and unsteady states) calculation results prepared to support annex to Safety Analysis Report for MARIA reactor in order to obtain approval for program of testing low-enriched uranium (LEU) lead test fuel assemblies (LTFA) manufactured by CERCA. This includes presentation of the limits and operational constraints to be in effect during the fuel testing investigations. Also, the scope of testing program (which began in August 2009), including additional measurements and monitoring procedures, is described.

  9. Design considerations of the irradiation test vehicle for the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.C.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    1997-08-01

    An irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) is being jointly developed by the Lockheed Martin Idaho Technologies Company (LMIT) and the U.S. Fusion Program. The vehicle is intended for neutron irradiation testing of candidate structural materials, including vanadium-based alloys, silicon carbide composites, and low activation steels. It could possibly be used for U.S./Japanese collaboration in the Jupiter Program. The first test train is scheduled to be completed by September 1998. In this report, we present the functional requirements for the vehicle and a preliminary design that satisfies these requirements.

  10. Calabash Chalk's Geophagy Affects Gestating Rats' Behavior and the Histomorphology of the Cerebral Cortex

    Directory of Open Access Journals (Sweden)

    Moses B. Ekong

    2014-01-01

    Full Text Available Introduction. Calabash chalk contains heavy metals, and this lead to this study on the effect of this chalk on the behavior and the histomorphology of the cerebral cortex of gestating rats. Material & Methods. 24 female rats were equally divided into 4 groups and were mated at preostrous with the males. The day after mating was designated as day 1 of gestation. On gestation days 7–20, groups 1, 2, 3, and 4 animals were treated with 1 mL of distilled water, and 1 mL (200 mg/kg, 2 mL (400 mg/kg, and 3 mL (600 mg/kg of calabash chalk suspension, respectively. On pregnancy day 21, behavioral tests using the open field and the light/dark mazes were carried out and the animals subsequently euthanized and their brains were routinely processed. Results. There was no difference in ambulatory activities, but group 4 animals had more (P<0.05 transition frequency and were more averse to the dark in the light and dark field, while sections of the cerebral cortex showed a higher (P<0.05 cellular population, hypertrophied pyramidal cells, and vacuolations in the treatment groups. Conclusion. Calabash chalk may have anxiolytic effect especially at high dose in the light and dark field but not in the open field and can stimulate maternal cerebral cortical cellular changes.

  11. Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Bradley K. Heath

    2014-03-01

    This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

  12. 49 CFR Appendix B to Part 179 - Procedures for Simulated Pool and Torch-Fire Testing

    Science.gov (United States)

    2010-10-01

    ... used to simulate a torch fire under paragraph 3a(2) of this appendix. (2) The back of the bare plate... 49 Transportation 2 2010-10-01 2010-10-01 false Procedures for Simulated Pool and Torch-Fire... SPECIFICATIONS FOR TANK CARS Pt. 179, App. B Appendix B to Part 179—Procedures for Simulated Pool and...

  13. The Effective Convectivity Model for Simulation and Analysis of Melt Pool Heat Transfer in a Light Water Reactor Pressure Vessel Lower Head

    Energy Technology Data Exchange (ETDEWEB)

    Tran, Chi Thanh

    2009-09-15

    Severe accidents in a Light Water Reactor (LWR) have been a subject of intense research for the last three decades. The research in this area aims to reach understanding of the inherent physical phenomena and reduce the uncertainties in their quantification, with the ultimate goal of developing models that can be applied to safety analysis of nuclear reactors, and to evaluation of the proposed accident management schemes for mitigating the consequences of severe accidents. In a hypothetical severe accident there is likelihood that the core materials will be relocated to the lower plenum and form a decay-heated debris bed (debris cake) or a melt pool. Interactions of core debris or melt with the reactor structures depend to a large extent on the debris bed or melt pool thermal hydraulics. In case of inadequate cooling, the excessive heat would drive the structures' overheating and ablation, and hence govern the vessel failure mode and timing. In turn, threats to containment integrity associated with potential ex-vessel steam explosions and ex-vessel debris uncoolability depend on the composition, superheat, and amount of molten corium available for discharge upon the vessel failure. That is why predictions of transient melt pool heat transfer in the reactor lower head, subsequent vessel failure modes and melt characteristics upon the discharge are of paramount importance for plant safety assessment. The main purpose of the present study is to develop a method for reliable prediction of melt pool thermal hydraulics, namely to establish a computational platform for cost-effective, sufficiently-accurate numerical simulations and analyses of core Melt-Structure-Water Interactions in the LWR lower head during a postulated severe core-melting accident. To achieve the goal, an approach to efficient use of Computational Fluid Dynamics (CFD) has been proposed to guide and support the development of models suitable for accident analysis. The CFD method, on the one hand

  14. Multi-Purpose Thermal Hydraulic Loop: Advanced Reactor Technology Integral System Test (ARTIST) Facility for Support of Advanced Reactor Technologies

    Energy Technology Data Exchange (ETDEWEB)

    James E. O' Brien; Piyush Sabharwall; SuJong Yoon

    2001-11-01

    Effective and robust high temperature heat transfer systems are fundamental to the successful deployment of advanced reactors for both power generation and non-electric applications. Plant designs often include an intermediate heat transfer loop (IHTL) with heat exchangers at either end to deliver thermal energy to the application while providing isolation of the primary reactor system. In order to address technical feasibility concerns and challenges a new high-temperature multi-fluid, multi-loop test facility “Advanced Reactor Technology Integral System Test facility” (ARTIST) is under development at the Idaho National Laboratory. The facility will include three flow loops: high-temperature helium, molten salt, and steam/water. Details of some of the design aspects and challenges of this facility, which is currently in the conceptual design phase, are discussed

  15. Safety Evaluation Report Restart of K-Reactor Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    1991-10-01

    In April 1991, the Department of Energy (DOE) issued DOE/DP-0084T, Safety Evaluation Report Restart of K-Reactor Savannah River Site.'' The Safety Evaluation Report (SER) documents the results of DOE reviews and evaluations of the programmatic aspects of a large number of issues necessary to be satisfactorily addressed before restart. The issues were evaluated for compliance with the restart criteria included in the SER. The results of those evaluations determined that the restart criteria had been satisfied for some of the issues. However, for most of the issues at least part of the applicable restart criteria had not been found to be satisfied at the time the evaluations were prepared. For those issues, open or confirmatory items were identified that required resolution. In August 1991, DOE issued DOE/DP-0090T, Safety Evaluation Report Restart of K-Reactor Savannah River Site Supplement 1.'' That document was the first Supplement to the April 1991 SER, and documented the resolution of 62 of the open items identified in the SER. This document is the second Supplement to the April 1991 SER. This second SER Supplement documents the resolution of additional open times identified in the SER, and includes a complete list of all remaining SER open items. The resolution of those remaining open items will be documented in future SER Supplements. Resolution of all open items for an issue indicates that its associated restart criteria have been satisfied, and that DOE concludes that the programmatic aspects of the issue have been satisfactorily addressed.

  16. Evaluation of radcal gamma thermometers for in-core monitoring of Savannah River Site production reactors

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W.; Crowley, J.L. [DELTA M Corp., Oak Ridge, TN (United States); Croft, W.D. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1991-12-31

    The Savannah River Site (SRS) recently obtained a quantity of Radcal Gamma Thermometer Assemblies (RGTAs) for in-core monitoring of local power in their production reactors. The RGTAs, manufactured by DELTA M Corporation in Oak Ridge, Tennessee, contained seven Self Calibrating Gamma Thermometer (SCGT) sensors within a 7.26 mm diameter, 3.06 m length with a total length of 5.6 m. All RGTAs contained an isolated segmented heater cable for in-situ calibration. Each SCGT sensor was subjected to a 40 point calibration at discrete power levels from 0.5 to 6 watts per gram (w/g) under both joule and cable power. Calibration equations were developed from this to predict reactor power at each sensor. Additionally three units were calibrated at combined joule and cable heating conditions from 0.5 to 2.5 w/g cable and 0.5 to 6 w/g joule. A statistical analysis of all data was used to derive prediction equations that enable SRS engineers to precisely track any changes in sensor calibration throughout the lifetime of the instruments. This paper presents the detailed configuration of the 36 units manufactured for SRS, reviews the calibration results, and discusses the utility and accuracy of the statistically derived prediction equations for in-situ calibration.

  17. Evaluation of radcal gamma thermometers for in-core monitoring of Savannah River Site production reactors

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W.; Crowley, J.L. (DELTA M Corp., Oak Ridge, TN (United States)); Croft, W.D. (Westinghouse Savannah River Co., Aiken, SC (United States))

    1991-01-01

    The Savannah River Site (SRS) recently obtained a quantity of Radcal Gamma Thermometer Assemblies (RGTAs) for in-core monitoring of local power in their production reactors. The RGTAs, manufactured by DELTA M Corporation in Oak Ridge, Tennessee, contained seven Self Calibrating Gamma Thermometer (SCGT) sensors within a 7.26 mm diameter, 3.06 m length with a total length of 5.6 m. All RGTAs contained an isolated segmented heater cable for in-situ calibration. Each SCGT sensor was subjected to a 40 point calibration at discrete power levels from 0.5 to 6 watts per gram (w/g) under both joule and cable power. Calibration equations were developed from this to predict reactor power at each sensor. Additionally three units were calibrated at combined joule and cable heating conditions from 0.5 to 2.5 w/g cable and 0.5 to 6 w/g joule. A statistical analysis of all data was used to derive prediction equations that enable SRS engineers to precisely track any changes in sensor calibration throughout the lifetime of the instruments. This paper presents the detailed configuration of the 36 units manufactured for SRS, reviews the calibration results, and discusses the utility and accuracy of the statistically derived prediction equations for in-situ calibration.

  18. Modelling of turbulent hydrocarbon combustion. Test of different reactor concepts for describing the interactions between turbulence and chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, C.; Kremer, H. [Ruhr-Universitaet Bochum, Lehrstuhl fuer Energieanlagentechnik, Bochum (Germany); Kilpinen, P.; Hupa, M. [Aabo Akademi, Turku (Finland). Combustion Chemistry Research Group

    1997-12-31

    The detailed modelling of turbulent reactive flows with CFD-codes is a major challenge in combustion science. One method of combining highly developed turbulence models and detailed chemistry in CFD-codes is the application of reactor based turbulence chemistry interaction models. In this work the influence of different reactor concepts on methane and NO{sub x} chemistry in turbulent reactive flows was investigated. Besides the classical reactor approaches, a plug flow reactor (PFR) and a perfectly stirred reactor (PSR), the Eddy-Dissipation Combustion Model (EDX) and the Eddy Dissipation Concept (EDC) were included. Based on a detailed reaction scheme and a simplified 2-step mechanism studies were performed in a simplified computational grid consisting of 5 cells. The investigations cover a temperature range from 1273 K to 1673 K and consider fuel-rich and fuel-lean gas mixtures as well as turbulent and highly turbulent flow conditions. All test cases investigated in this study showed a strong influence of the reactor residence time on the species conversion processes. Due to this characteristic strong deviations were found for the species trends resulting from the different reactor approaches. However, this influence was only concentrated on the `near burner region` and after 4-5 cells hardly any deviation and residence time dependence could be found. The importance of the residence time dependence increased when the species conversion was accelerated as it is the case for overstoichiometric combustion conditions and increased temperatures. The study focused furthermore on the fine structure in the EDC. Unlike the classical approach this part of the cell was modelled as a PFR instead of a PSR. For high temperature conditions there was hardly any difference between both reactor types. However, decreasing the temperature led to obvious deviations. Finally, the effect of the selective species transport between the cells on the conversion process was investigated

  19. Development of automatic Ultrasonic testing equipment for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kor R.; Kim, Jae H.; Lee, Jae C.

    1996-06-01

    The selected weld areas of a reactor pressure vessel and adjacent piping are examined by the remote mechanized ultrasonic testing (MUT) equipment. Since the MUT equipment was purchased from southwest Research Institute (SwRI) in April 1985, 15 inservice inspections and 5 preservice inspections are performed with this MUT equipment. However due to the old age of the equipment and frequent movements to plant sites, the reliability of examination was recently decreased rapidly and it is very difficult to keep spare parts. In order to resolve these problems and to meet the strong request from plant sites, we intend to develop a new 3-axis control system including hardware and software. With this control system, we expect more efficient and reliable examination of the nozzle to shell weld areas, which is specified in ASME Code Section XI. The new 3-axis control system hardware and software were designed and development of our own control system, the advanced technologies of computer control mechanism were established and examination reliability of the nozzle to shell weld area was improved. With the development of our 3-axis control system for PaR ISI-2 computer control system, the reliability of nozzle to shell weld area examination has been improved. The established technologies from the development and detailed analysis of existing control system, are expected to be applied to the similar control systems in nuclear power plants. (author). 12 refs., 4 tabs., 33 figs.

  20. Experimental investigation of heat transfer during severe accident of a Pressurized Heavy Water Reactor with simulated decay heat generation in molten pool inside calandria vessel

    Energy Technology Data Exchange (ETDEWEB)

    Prasad, Sumit Vishnu, E-mail: svprasad@barc.gov.in; Nayak, Arun Kumar, E-mail: arunths@barc.gov.in

    2016-07-15

    Highlights: • Scaled test facility simulating the calandria vessel and calandria vault water of PHWR with simulated decay heat was built. • Experiments conducted with simulant material at about 1200 °C. • Experimental result shows that melt coolability and growth rate of crust thickness are affected by presence of decay heat. • No gap was observed between the crust and vessel on opening. • Result shows that vessel integrity is intact with presence of water inside water tank in both cases. - Abstract: The present study focuses on experimental investigation in a scaled facility of an Indian PHWR to investigate the coolability of molten corium with simulated decay heat in the simulated calandria vessel. Molten borosilicate glass was used as the simulant due to its comparable heat transfer characteristics similar to prototypic material. About 60 kg of the molten material was poured into the test section at about 1200 °C. Decay heat in the melt pool was simulated using four high watt heaters cartridges, each having 9.2 kW. The temperature distributions inside the molten pool, across the vessel wall thickness and vault water were measured. Experimental results obtained are compared with the results obtained previously for no decay heat case. The results indicated that presence of decay heat seriously affects the coolability behaviour and formation of crust in the melt pool. The location and magnitude of maximum heat flux and surface temperature of the vessel also are affected in the presence of decay heat.

  1. Effect of Fluid Dynamic Viscosity on the Strength of Chalk

    DEFF Research Database (Denmark)

    Hedegaard, K.; Fabricius, Ida Lykke

    The mechanical strength of high porosity and weakly cemented chalk is affected by the fluid in the pores. In this study, the effect of the dynamic viscosity of non-polar fluids has been measured on outcrop chalk from Sigerslev Quarry, Stevns, Denmark. The outcome is that the measured strength...

  2. The Beauty of the Beasts in Chalk Pastels

    Science.gov (United States)

    Skophammer, Karen

    2010-01-01

    In this article, the author describes how her seventh-grade art students captured an image of a stuffed animal in the "whole-to-part" drawing technique using chalk pastels. Shading with chalk pastels can give a gradual change in value from dark to light. The shading and color changes the mood of the original drawing, and adds texture, too. Chalk…

  3. Stylolites, porosity, depositional texture, and silicates in chalk facies sediments

    DEFF Research Database (Denmark)

    Fabricius, Ida Lykke; Borre, Mai K.

    2007-01-01

    Comparison of chalk on the Ontong Java Plateau and chalk in the Central North Sea indicates that, whereas pressure dissolution is controlled by effective burial stress, pore-filling cementation is controlled by temperature. Effective burial stress is caused by the weight of all overlying water an...

  4. RJH, a new test reactor in Europe; Le RJH - un nouveau reacteur d'essai en europe

    Energy Technology Data Exchange (ETDEWEB)

    Iracane, D. [CEA Saclay, Dir. de l' Energie Nucleaire (DEN), 91 - Gif sur Yvette (France)

    2005-07-01

    Material test reactors (MTR) are now ageing in Europe and they cannot secure the experimental needs over next decades. In this context, a new MTR, named Jules Horowitz reactor (RJH), operated as an international user-facility, is under development on the Cea's site of Cadarache (France). The design studies will end in 2007, the construction stage will follow and RJH commissioning is scheduled in 2014. Its construction costs are estimated to 500 million euros. RJH is a pool reactor of 100 MWth, its core will be inserted in a pressurized vessel with a primary circuit assuring water flow through forced convection. The core inlet-outlet temperature is about 25-40 Celsius degrees. RJH core is designed to use a high density - low enrichment UMo nuclear fuel (8 gU/cm{sup 3}, enrichment rate: 19.75%). Experimental devices located in the core will benefit from neutron fluxes ranging from 2.5 10{sup 14} n/cm{sup 2}.s to 5.10{sup 14} n/cm{sup 2}.s (E > 1 MeV). RJH is designed to manage simultaneously 10 experiments in the core and as many in the reflector. (A.C.)

  5. 77 FR 68133 - Guidance for Industry: Use of Nucleic Acid Tests on Pooled and Individual Samples From Donors of...

    Science.gov (United States)

    2012-11-15

    ...The Food and Drug Administration (FDA) is announcing the availability of a document entitled ``Guidance for Industry: Use of Nucleic Acid Tests on Pooled and Individual Samples from Donors of Whole Blood and Blood Components, including Source Plasma, to Reduce the Risk of Transmission of Hepatitis B Virus,'' dated October 2012. The guidance document provides recommendations on the use of FDA-......

  6. 76 FR 72950 - Draft Guidance for Industry: Use of Nucleic Acid Tests on Pooled and Individual Samples From...

    Science.gov (United States)

    2011-11-28

    ...The Food and Drug Administration (FDA) is announcing the availability of a draft document entitled ``Guidance for Industry: Use of Nucleic Acid Tests (NAT) on Pooled and Individual Samples from Donors of Whole Blood and Blood Components (including Recovered Plasma, Source Plasma and Source Leukocytes) to Adequately and Appropriately Reduce the Risk of Transmission of Hepatitis B Virus (HBV),......

  7. Water weakening of chalk explaied from a fluid-solid friction factor

    DEFF Research Database (Denmark)

    Andreassen, Katrine Alling; Fabricius, Ida Lykke

    2010-01-01

    were tested at temperatures from 20°C to 130°C with the following pore fluids: fresh water, synthetic seawater of different chemical compositions, methanol, glycol, and oil of varying viscosity. The data was evaluated according to failure lines and yield envelopes for all fluids and temperatures while...... to the macroscale failure and pore collapse properties. The Biot critical frequency incorporates the porosity, permeability, fluid density and fluid viscosity, where the latter is highly temperature dependent – it does not include the applied sound velocity frequency. The listed parameters are usually determined...... during laboratory tests and the fluid viscosity and density may be found in tabulated references. There exist a number of previously published laboratory test results on chalk which was collected from Brazilian, unconfined compression and triaxial tests. The data spans four different chalk types which...

  8. ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING DEACTIVATION AND DECOMMISSIONING OF REACTOR VESSELS AT THE SAVANNAH RIVER SITE

    Energy Technology Data Exchange (ETDEWEB)

    Wiersma, B.; Serrato, M.; Langton, C.

    2010-11-10

    The R- and P-reactor vessels at the Savannah River Site (SRS) are being prepared for deactivation and decommissioning (D&D). D&D activities will consist primarily of physically isolating and stabilizing the reactor vessel by filling it with a grout material. The reactor vessels contain aluminum alloy materials, which pose a concern in that aluminum corrodes rapidly when it comes in contact with the alkaline grout. A product of the corrosion reaction is hydrogen gas and therefore potential flammability issues were assessed. A model was developed to calculate the hydrogen generation rate as the reactor is being filled with the grout material. Three options existed for the type of grout material for D&D of the reactor vessels. The grout formulation options included ceramicrete (pH 6-8), a calcium aluminate sulfate (CAS) based cement (pH 10), or Portland cement grout (pH 12.4). Corrosion data for aluminum in concrete were utilized as input for the model. The calculations considered such factors as the surface area of the aluminum components, the open cross-sectional area of the reactor vessel, the rate at which the grout is added to the reactor vessel, and temperature. Given the hydrogen generation rate, the hydrogen concentration in the vapor space of the reactor vessel above the grout was calculated. This concentration was compared to the lower flammability limit for hydrogen. The assessment concluded that either ceramicrete or the CAS grout may be used to safely grout the P-reactor vessel. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Portland cement grout, on the other hand, for the same range of process parameters did not provide a margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations in the P-reactor vessel. Therefore, it was recommended that this grout not be utilized for this task. On the other hand, the R-reactor vessel

  9. Adaptation of Crack Growth Detection Techniques to US Material Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; Sebastien P. Teysseyre; Kurt L. Davis; Joy L. Rempe; Gordon Kohse; Yakov Ostrovsky; David M. Carpenter

    2014-04-01

    A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some materials testing reactors (MTRs) outside the U.S., such as the Halden Boiling Water Reactor (HBWR), have deployed a technique to measure crack growth propagation during irradiation. This technique incorporates a compact loading mechanism to stress the specimen during irradiation. A crack in the specimen is monitored using the Direct Current Potential Drop (DCPD) method. A project is underway to develop and demonstrate the performance of a similar type of test rig for use in U.S. MTRs. The first year of this three year project was devoted to designing, analyzing, fabricating, and bench top testing a mechanism capable of applying a controlled stress to specimens while they are irradiated in a pressurized water loop (simulating PWR reactor conditions). During the second year, the mechanism will be tested in autoclaves containing high pressure, high temperature water with representative water chemistries. In addition, necessary documentation and safety reviews for testing in a reactor environment will be completed. In the third year, the assembly will be tested in the Massachusetts Institute of Technology Reactor (MITR) and Post Irradiation Examinations (PIE) will be performed.

  10. Development of System Analysis Code for Pool-Type Fast Reactor Under Transient Operation%池式快堆系统瞬态分析软件开发

    Institute of Scientific and Technical Information of China (English)

    陆道纲; 隋丹婷

    2012-01-01

    为实现快堆系统分析软件国产化,在已开发的适用于稳态计算的池式快堆系统分析软件SAC-CFR的基础上,进一步开发了系统各部件的瞬态模型、控制系统和保护系统模型、瞬态工况热工水力学的求解逻辑,完成瞬态计算功能的开发.通过对日本文殊快堆45%功率汽机跳闸工况进行建模分析,验证了SAC-CFR用于系统瞬态分析的有效性,为进一步开发非能动余热排出系统分析模型打下了基础.%Aiming at developing system analysis code independently, the system analysis code for pool-type fast reactor in China (SAC-CFR) under transient operation was developed with further development of component transient model, plant control and protection system model, calculation logic for system transient thermal-hydraulic analysis based on the former SAC-CFR version applicable to steady state analysis. The transient started from turbine trip test at 45 % thermal output in the Monju Plant was analyzed with the developed SAC-CFR. A good agreement between the calculated results and the test data was obtained. SAC-CFR is now ready to incorporate passive residual heat removal model for China Experimental Fast Reactor.

  11. Simulation of decay heat removal by natural convection in a pool type fast reactor model-ramona-with coupled 1D/2D thermal hydraulic code system

    Energy Technology Data Exchange (ETDEWEB)

    Kasinathan, N.; Rajakumar, A.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    1995-09-01

    Post shutdown decay heat removal is an important safety requirement in any nuclear system. In order to improve the reliability of this function, Liquid metal (sodium) cooled fast breeder reactors (LMFBR) are equipped with redundant hot pool dipped immersion coolers connected to natural draught air cooled heat exchangers through intermediate sodium circuits. During decay heat removal, flow through the core, immersion cooler primary side and in the intermediate sodium circuits are also through natural convection. In order to establish the viability and validate computer codes used in making predictions, a 1:20 scale experimental model called RAMONA with water as coolant has been built and experimental simulation of decay heat removal situation has been performed at KfK Karlsruhe. Results of two such experiments have been compiled and published as benchmarks. This paper brings out the results of the numerical simulation of one of the benchmark case through a 1D/2D coupled code system, DHDYN-1D/THYC-2D and the salient features of the comparisons. Brief description of the formulations of the codes are also included.

  12. Controls on the spatial and temporal variability of Rn-222 in riparian groundwater in a lowland Chalk catchment.

    OpenAIRE

    Mullinger, Neil J.; Pates, Jackie M.; Binley, Andrew M.; Crook, N. P.

    2009-01-01

    Radon is a powerful tracer of stream-aquifer interactions. However, it is important to consider the source and behaviour of radon in groundwater when interpreting observations of river radon in relation to groundwater discharge. Here we characterise the variability in groundwater radon concentrations in the riparian zone of a Chalk catchment. Groundwater 222Rn (radon) concentrations were determined in riparian zone boreholes at two sites in the Lambourn catchment, Berkshire, UK, over a two ye...

  13. Project size and common pool size: An empirical test using Danish municipal mergers

    DEFF Research Database (Denmark)

    Hansen, Sune Welling

    The paper examines the proposition that project size tends to increase with common pool size from the law of 1 over n (Weingast et al, 1981). This remains under-investigated and a recent study conducted by Primo & Snyder (2008) argues, and empirically substantiates, a reverse law of 1 over n....... In this paper, the proposition is examined using recent municipal mergers in Denmark as a crucial case under less favorable conditions. The paper finds positive, statistically and economically significant effects of common pool size in the ultimate year of the treatment period. These results are consistent...

  14. IRRADIATION TESTING OF THE RERTR FUEL MINIPLATES WITH BURNABLE ABSORBERS IN THE ADVANCED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    I. Glagolenko; D. Wachs; N. Woolstenhulme; G. Chang; B. Rabin; C. Clark; T. Wiencek

    2010-10-01

    Based on the results of the reactor physics assessment, conversion of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) can be potentially accomplished in two ways, by either using U-10Mo monolithic or U-7Mo dispersion type plates in the ATR fuel element. Both designs, however, would require incorporation of the burnable absorber in several plates of the fuel element to compensate for the excess reactivity and to flatten the radial power profile. Several different types of burnable absorbers were considered initially, but only borated compounds, such as B4C, ZrB2 and Al-B alloys, were selected for testing primarily due to the length of the ATR fuel cycle and fuel manufacturing constraints. To assess and compare irradiation performance of the U-Mo fuels with different burnable absorbers we have designed and manufactured 28 RERTR miniplates (20 fueled and 8 non-fueled) containing fore-mentioned borated compounds. These miniplates will be tested in the ATR as part of the RERTR-13 experiment, which is described in this paper. Detailed plate design, compositions and irradiations conditions are discussed.

  15. The creation of digital thematic soil maps at the regional level (with the map of soil carbon pools in the Usa River basin as an example)

    Science.gov (United States)

    Pastukhov, A. V.; Kaverin, D. A.; Shchanov, V. M.

    2016-09-01

    A digital map of soil carbon pools was created for the forest-tundra ecotone in the Usa River basin with the use of ERDAS Imagine 2014 and ArcGIS 10.2 software. Supervised classification and thematic interpretation of satellite images and digital terrain models with the use of a georeferenced database on soil profiles were applied. Expert assessment of the natural diversity and representativeness of random samples for different soil groups was performed, and the minimal necessary size of the statistical sample was determined.

  16. Study of fast reactor safety test facilities. Preliminary report

    Energy Technology Data Exchange (ETDEWEB)

    Bell, G.I.; Boudreau, J.E.; McLaughlin, T.; Palmer, R.G.; Starkovich, V.; Stein, W.E.; Stevenson, M.G.; Yarnell, Y.L.

    1975-05-01

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods. (DG)

  17. Chalk Formations as Natural Barriers towards Radionuclide Migration

    DEFF Research Database (Denmark)

    Pedersen, Walther Batsberg; Carlsen, Lars; Jensen, Bror Skytte

    1985-01-01

    A series of chalk samples from the cretaceous formation overlying the Erslev salt dome have been studied in order to establish permeabilities, porosities, dispersion-, diffusion-, and sorption characteristics of the chalk. The chalk was found to be porous (∊≈0.4), however, of rather low...... permeability (k≈10-7 cm/sec). It was found that the material exhibits a retarding effect on migration of cationic species as Cs+, Sr2+ , Co2+, and Eu3+, whereas ionic species as Cl- move with the water front. The geochemical implications are discussed...

  18. Research on Power Ramp Testing Method for PWR Fuel Rod at Research Reactor

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    In order to develop high performance fuel assembly for domestic nuclear power plant, it is necessary to master some fundamental test technology. So the research on the power ramp testing methods is proposed. A tentative power ramp test for short PWR fuel rod has been conducted at the heavy water research reactor (HWRR) in China Institute of Atomic Energy (CIAE) in May of 2001. The in-pile test rig was placed into the central channel of the reactor . The test rig consists of pressure pipe assembly, thimble, solid neutron absorbing screen and its driving parts, etc.. The test

  19. Groundwater modeling of the proposed new production reactor site, Savannah River Site, South Carolina

    Energy Technology Data Exchange (ETDEWEB)

    Looney, B.B.; Haselow, J.S.; Andersen, P.F.; Spalding, C.P.; Davis, D.H.

    1990-01-05

    This report addresses groundwater modeling performed to support the Environmental Impact Statement (EIS) that is being prepared by the Department of Energy (DOE). The EIS pertains to construction and operation of a new production reactor (NPR) that is under consideration for the Savannah River Site (SRS). Three primary issues are addressed by the modeling analysis: (1) groundwater availability, (2) changes in vertical hydraulic gradients as a result of groundwater pumpage, and (3) migration of potential contaminants from the NPR site. The modeling indicates that the maximum pumpage to be used, 1000 gpm, will induce only minor drawdown across SRS. Pumpage of this magnitude will have a limited effect on the upward gradient from the Cretaceous into the Tertiary near Upper Three Runs Creek. Potentiometric surface maps generated from modeled results indicate that horizontal flow in the water table is either towards Four Mile Creek to the north or to Pen Branch on the south. Particle tracking analysis indicates that the primary flow paths are vertical into the Lower Tertiary Zone, with very little lateral migration. Total travel times from the NPR site to the edge of the model (approximately 3 miles) is on the order of 50 years. The flow direction of water in the Lower Tertiary Zone is relatively well defined due to the regional extent of the flow system. The Pen Branch Fault does not influence contaminant migration for this particular site because it is in the opposite direction of Lower Tertiary Zone groundwater flow. 20 refs., 27 figs., 2 tabs.

  20. Estimated recurrence frequencies for initiating accident categories associated with the Clinch River Breeder Reactor Plant design

    Energy Technology Data Exchange (ETDEWEB)

    Copus, E R

    1982-04-01

    Estimated recurrence frequencies for each of twenty-five generic LMFBR initiating accident categories were quantified using the Clinch River Breeder Reactor Plant (CRBRP) design. These estimates were obtained using simplified systems fault trees and functional event tree models from the Accident Delineation Study Phase I Final Report coupled with order-of-magnitude estimates for the initiator-dependent failure probabilities of the individual CRBRP engineered safety systems. Twelve distinct protected accident categories where SCRAM is assumed to be successful are estimated to occur at a combined rate of 10/sup -3/ times per year while thirteen unprotected accident categories in which SCRAM fails are estimated to occur at a combined rate on the order of 10/sup -5/ times per year. These estimates are thought to be representative despite the fact that human performance factors, maintenance and repair, as well as input common cause uncertainties, were not treated explicitly. The overall results indicate that for the CRBRP design no single accident category appears to be dominant, nor can any be totally eliminated from further investigation in the areas of accident phenomenology for in-core events and post-accident phenomenology for containment.

  1. An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Cetiner, Sacit M [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

    2009-11-01

    This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.

  2. Reactor

    Science.gov (United States)

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  3. Evaluation of the aptitude for the service of the pool of the TRIGA Mark III reactor of the National Institute of Nuclear Research of Mexico; Evaluacion de la aptitud para el servicio de la piscina del reactor TRIGA Mark III del Instituto Nacional de Investigaciones Nucleares de Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Merino C, J.; Gachuz M, M.; Diaz S, A.; Arganis J, C.; Gonzalez R, C.; Nava G, T.; Medina R, M.J. [Departamento de Sintesis y Caracterizacion de Materiales del ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2001-07-01

    This work describes the evaluation of the structural integrity of the pool of the TRIGA Mark III reactor of the National Institute of Nuclear Research of Mexico, which was realized in July 2001, as an element to determine those actions for preventive and corrective maintenance which owner must do it for a safety and efficient operation of the component in the next years. (Author)

  4. Fatigue Test of Domestic Manufactured Reactor Pressure Vessel Steel

    Institute of Scientific and Technical Information of China (English)

    ZHONG; Wei-hua; TONG; Zhen-feng; NING; Guang-sheng; YU; Bin-tao

    2013-01-01

    The CAP1400 will be built by our country,after the self-dependent innovation work on the imported technology of AP1000,which is a 3rd generation NPP.Now,the design of CAP1400 key equipment is ongoing,and the fatigue design of the domestic manufactured key equipment,such as reactor pressure vessel(RPV),is found to be a main problem in the design work,as the fatigue data is lacked.Thus the

  5. Coupling channel evolution monitoring and RFID tracking in a large, wandering, gravel-bed river: Insights into sediment routing on geomorphic continuity through a riffle-pool sequence

    Science.gov (United States)

    Chapuis, Margot; Dufour, Simon; Provansal, Mireille; Couvert, Bernard; de Linares, Matthieu

    2015-02-01

    Bedload transport and bedform mobility in large gravel-bed rivers are not easily monitored, especially during floods. Large reaches present difficulties in bed access during flows for flow measurements. Because of these logistical issues, the current knowledge about bedload transport processes and bedform mobility lacks field-based information, while this missing information would precisely match river management needs. The lack of information linking channel evolution and particle displacements is even more striking in wandering reaches. The Durance River is a large, wandering, gravel-bed river (catchment area: 14,280 km2; mean width: 240 m), located in the southern French Alps and highly impacted by flow diversion and gravel mining. In order to improve current understanding of the link between sediment transport processes and river bed morphodynamics, we set up a sediment particle survey in the channel using Radio Frequency Identification (RFID) tracking and topographic surveys (GPS RTK and scour chains) for a 4-year recurrence interval flood. By combining topographic changes before and after a flood, intraflood erosion/deposition patterns from scour chains, differential routing of tracer particles, and spatial distribution of bed shear stress through a complex reach, this paper aims to define the critical shear stress for significant sediment mobility in this setting. Gravel tracking highlights displacement patterns in agreement with bar downstream migration and transport of particles across the riffle within this single flood event. Because no velocity measurements were possible during flood, a TELEMAC three-dimensional model helped interpret particle displacements by estimating spatial distribution of shear stresses and flow directions at peak flow. Although RFID tracking in a large, wandering, gravel-bed river does have some technical limitations (burial, recovery process time-consuming), it provides useful information on sediment routing through a riffle-pool

  6. How burial diagenesis of chalk sediments controls sonic velocity and porosity

    DEFF Research Database (Denmark)

    Fabricius, Ida Lykke

    2003-01-01

    to the progress of burial diagenesis of chalk, which is revised as follows: Newly deposited carbonate ooze and mixed sediments range in porosity from 60 to 80%, depending on the prevalence of hollow microfossils. Despite the high porosity, these sediments are not in suspension, as reflected in IFs of 0.......1 or higher. Upon burial, the sediments lose porosity by mechanical compaction, and concurrently, the calcite particles recrystallize into progressively more equant shapes. High compaction rates may keep the particles in relative motion, whereas low compaction rates allow the formation of contact cement......, whereby IF increases and chalk forms. Rock mechanical tests show that when compaction requires more than in-situ stress, porosity reduction is arrested. During subsequent burial, crystals and pores grow in size as a consequence of the continuing recrystallization. ne lack of porosity loss during...

  7. 10 CFR 830 Major Modification Determination for the Advanced Test Reactor Remote Monitoring and Management Capability

    Energy Technology Data Exchange (ETDEWEB)

    Bohachek, Randolph Charles [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-09-01

    The Advanced Test Reactor (ATR; TRA-670), which is located in the ATR Complex at Idaho National Laboratory, was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. While ATR is safely fulfilling current mission requirements, assessments are continuing. These assessments intend to identify areas to provide defense–in-depth and improve safety for ATR. One of the assessments performed by an independent group of nuclear industry experts recommended that a remote accident management capability be provided. The report stated that: “contemporary practice in commercial power reactors is to provide a remote shutdown station or stations to allow shutdown of the reactor and management of long-term cooling of the reactor (i.e., management of reactivity, inventory, and cooling) should the main control room be disabled (e.g., due to a fire in the control room or affecting the control room).” This project will install remote reactor monitoring and management capabilities for ATR. Remote capabilities will allow for post scram reactor management and monitoring in the event the main Reactor Control Room (RCR) must be evacuated.

  8. Evaluation of the pool chemistry model in MELCOR by considering the organic iodine formation within the pool using the phebus FPT-1 test data

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Hwa; Kim, Dong Ha [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2004-07-01

    In the severe accident, the understanding on the chemical form of iodine and its behavior is important for estimating the source terms because it has not only a active chemical features but also hazard to the public. From the current understanding on the early phase of severe accident, it is known that the most of the injected iodine into the containment is the aerosol form of metal iodide such as cesium iodide and silver iodide. The remaining fraction is the gaseous form of iodine such as molecular iodine, elemental iodine or hydrogen iodide. These aerosol and gaseous iodine can be removed by the engineered safety features such as the spray system. Also the natural depletion processes like condensing, settling or deposit on the surface remove the most of the suspended aerosol and gaseous iodine and finally, cause them to accumulate within the pool on the bottom floor. As time goes on, in the late phase of severe accident, the iodine pool chemistry phenomena starts under the strong radiation condition. The dissolved iodine within the pool can be transformed to the volatile iodine such as molecular iodine and organic iodine due to its strong chemical activity and various kind of immersed impurities such as alcohol or grease. This transformed volatile iodine can be partitioned to the atmosphere from the inside of pool. Consequently, an equilibrium concentration is reached and this suspended gaseous form of iodine becomes the main threat to the public when the containment fails or leaks in the late phase of severe accident. Therefore, decreasing the equilibrium concentration of the suspended gaseous iodine and the trapping of iodine within the pool are the key parameters of the iodine management strategy in the late phase of severe accident. The Phebus FPT-1 showed that the silver could play an important role in trapping the iodine in the sump by forming the insoluble silver iodide from I{sub 2} or I. The formation of volatile iodine depends on the PH, pool

  9. NaK pool-boiler solar receiver durability bench test. Volume 2: Metallurgical analysis

    Science.gov (United States)

    Goods, S. H.; Bradshaw, R. W.

    1995-01-01

    The principal materials used in the construction of a NaK based pool-boiler were analyzed. The device, operated for 7500 hours, accumulated 1000 thermal cycles to a peak temperature of 750 C. Haynes 230, used to fabricate the pool-boiler vessel, was found to perform satisfactorily. Air-side corrosion of the pool-boiler vessel was insignificant. Internal surface of the alloy exhibited some NaK-induced elemental dissolution; this dissolution was somewhat more extensive where the alloy was exposed to the liquid metal compared to regions exposed only to NaK vapor; however, the corresponding metal loss in all regions was inconsequential, never exceeding more than a few microns. Autogenous seam welds of the alloy responded in a similar fashion, exhibiting only minimal metal loss over the course of the experiment. While there was 50% loss in ductility of the alloy there remained adequate ductility for the anticipated operating environment. An enhanced boiling nucleation surface comprised of stainless steel powder brazed to the vessel ID showed no change in its structure. It remained intact, showing no cracking after repeated thermal cycling. Other materials used in the experiment showed more extensive degradation after exposure to the NaK. IN 600, used to fabricate thermowells, exhibited extensive surface and intergranular dissolution. Grain boundary dissolution was sufficiently severe in one of the thermowells to cause an air leak, resulting in experiment termination. BNi-3, a brazing alloy used to join the pool-boiler vessel, endcaps and thermowells, showed some dissolution where it was exposed to the NaK as well as thermal aging effects. However, all brazes remained structurally sound. A nickel metal ribbon showed catastrophic dissolution, resulting in the formation of deep (greater than 30 (mu)m) pits and cavities. A zirconium metal foil used to getter oxygen from the NaK became extremely brittle.

  10. Kolmanda aastatuhande piraadid / kommenteerinud Peter Chalk ja Gordan Van Hook

    Index Scriptorium Estoniae

    2009-01-01

    Piraatlusest Somaalia piirkonnas ja rahvusvahelistest dokumentidest piraatluse vastu võitlemiseks 21. sajandil. Kommenteerivad uurimiskeskuse RAND Corporation vanempoliitanalüütik Peter Chalk ja transpordikompanii Maersk Line innovatsiooni ja arenduse vanemdirektor Gordan Van Hook

  11. Coccomyxa actinabiotis sp. nov. (Trebouxiophyceae, Chlorophyta), a new green microalga living in the spent fuel cooling pool of a nuclear reactor.

    Science.gov (United States)

    Rivasseau, Corinne; Farhi, Emmanuel; Compagnon, Estelle; de Gouvion Saint Cyr, Diane; van Lis, Robert; Falconet, Denis; Kuntz, Marcel; Atteia, Ariane; Couté, Alain

    2016-10-01

    Life can thrive in extreme environments where inhospitable conditions prevail. Organisms which resist, for example, acidity, pressure, low or high temperature, have been found in harsh environments. Most of them are bacteria and archaea. The bacterium Deinococcus radiodurans is considered to be a champion among all living organisms, surviving extreme ionizing radiation levels. We have discovered a new extremophile eukaryotic organism that possesses a resistance to ionizing radiations similar to that of D. radiodurans. This microorganism, an autotrophic freshwater green microalga, lives in a peculiar environment, namely the cooling pool of a nuclear reactor containing spent nuclear fuels, where it is continuously submitted to nutritive, metallic, and radiative stress. We investigated its morphology and its ultrastructure by light, fluorescence and electron microscopy as well as its biochemical properties. Its resistance to UV and gamma radiation was assessed. When submitted to different dose rates of the order of some tens of mGy · h(-1) to several thousands of Gy · h(-1) , the microalga revealed to be able to survive intense gamma-rays irradiation, up to 2,000 times the dose lethal to human. The nuclear genome region spanning the genes for small subunit ribosomal RNA-Internal Transcribed Spacer (ITS) 1-5.8S rRNA-ITS2-28S rRNA (beginning) was sequenced (4,065 bp). The phylogenetic position of the microalga was inferred from the 18S rRNA gene. All the revealed characteristics make the alga a new species of the genus Coccomyxa in the class Trebouxiophyceae, which we name Coccomyxa actinabiotis sp. nov.

  12. Moderated heat pipe thermionic reactor (MOHTR) module development and test

    Science.gov (United States)

    Merrigan, Michael A.; Trujillo, Vincent L.

    1992-01-01

    The Moderated Heat Pipe Thermionic Reactor (MOHTR) thermionic space reactor design combines the low risk technology associated with the Thermionic Fuel Element (TFE) Verification Program with the high reliability heat transfer capability of liquid metal heat pipe technology. The resulting design concept, capable of implementation over the power range of 10 to 100 kWe, offers efficiency and reliability with reduced risk of single point failures. The union of TFE and heat pipe technology is achieved by imbedding TFEs and heat pipes in a beryllium matrix to which they are thermally coupled by brazing or by liquid metal (NaK or Na) bonding. The reactor employs an array of TFE modules, each comprising a TFE, a zirconium hydride (ZrH) cylinder for neutron moderation, and heat pipes for transport of heat from the collector surface of the TFE to the waste heat radiator. An advantage of the design is the low temperature drop from the collector surface to the radiating surface. This is a result of the elimination of electrical insulation from the heat transport path through electrical isolation of the modules. The module used in this study consisted of a beryllium core, and electrical cartridge heater simulating the TFE, and three heat pipes to dissipate the waste heat. The investigation was focused on the thermal performance of the assembly, including evaluation of the sodium and braze bonding options for minimizing the thermal resistance between the elements, the temperature distribution in the beryllium matrix, and the heat pipe performance. Continuing subjects of the investigation include performance of the heat pipes through start-up transients, during normal operation, and in a single heat pipe failure mode. Secondary objectives of the investigation include correlation of analytic models for the thermionic element and module including the effects of gap thermal conductances at the modules electrically insulated surfaces.

  13. 1992 Contaminant Survey of the Upper Mississippi River National Wildlife and Fish Refuge, Pools 12 and 14

    Data.gov (United States)

    US Fish and Wildlife Service, Department of the Interior — Sediments and tissue samples from five areas in and near the Upper Mississippi River National Wildlife and Fish Refuge were analyzed for inorganic and organic...

  14. Validation of RELAP5 model of experimental test rig simulating the natural convection in MTR research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Khedr, A.; Abdel-Latif, Salwa H. [Nuclear and Radiological Regulatory Authority, Cairo (Egypt); Abdel-Hadi, Eed A. [Benha Univ., Cairo (Egypt). Shobra Faculty of Engineering; D' Auria, F. [Pisa Univ. (Italy)

    2016-03-15

    In an attempt to understand the built-up of natural circulation in MTR pool type upward flow research reactors after loss of power, an experimental test rig was built to simulate the loop of natural circulation in MTR reactors. The test rig consisting of two vertically oriented branches, in one of them the core is simulated by two rectangular, electrically heated, parallel channels. The other branch simulates the part of the return pipe that participates in the development of core natural circulation. In the first phase of the work, many experimental runs at different conditions of channel's power and branch's initial temperatures are performed. The channel's coolant and surface temperatures were measured. The measurements and their interpretation were published by the first three authors. In the present work the thermal hydraulic behavior of the test rig is complemented by theoretical analysis using RELAP5 Mod 3.3 system code. The analysis consisting of two parts; in the first part RELAP5 model is validated against the measured values and in the second part some of the other not measured hydraulic parameters are predicted and analyzed. The test rig is typically nodalized and an input dick is prepared. In spite of the low pressure of the test rig, the results show that RELAP5 qualitatively predicts the thermal hydraulic behaviour and the accompanied phenomenon of flow inversion of such facilities. Quantitatively, there is a difference between the predicted and measured values especially the channel's surface temperature. This difference may be return to the uncertainties in initial conditions of experimental runs, the position of the thermocouples which buried inside the heat structure, and the heat transfer package in RELAP5.

  15. The Jules Horowitz Reactor - A new High Performance European Material Testing Reactor open to International Users Present Status and Objectives

    Energy Technology Data Exchange (ETDEWEB)

    Iracane, Daniel; Bignan, Gilles [CEA Atomic Energy Commission Saclay Batiment 121- 91191 Gif Sur Yvette (France); Lindbaeck, Jan-Erik; Blomgren, Jan [VATTENFALL AB Nuclear Power Jaemtlandsgatan 99 SE-16287 Stockholm (Sweden)

    2010-07-01

    The development of sustainable nuclear energy requires R and D on fuel and material behaviour under irradiation with a high level of performance in order to meet the needs and challenges for the benefit of industry, research and public bodies. These stakes require a sustainable and secured access to an up-to-date high performance Material Testing Reactor. Following a broad survey within the European Research Area, the international community agreed that the need for Material Test Reactors in support of nuclear power plant safety and operation will continue in the context of sustainable nuclear energy. The Jules Horowitz Reactor project (JHR) copes with this context. JHR is designed as a user facility addressing the needs of the international community. This means: - flexibility with irradiation loops able to reproduce a large variation in operation conditions of different power reactor technologies, - high flux capacity to address Generations II, III, and IV needs. JHR is designed, built and operated as an international user facility because: - Given the maturity and globalization of the industry, domestic tools have no more the required level of economic and technical efficiency. Meanwhile, countries with nuclear energy need an access to high performance irradiation experimental capabilities to support technical skill and guarantee the competitiveness and safety of nuclear energy. - Many research items related to safety or public policy (waste management, etc.) require international cooperation to share costs and benefits of resulting consensus. JHR design is optimised for offering high performance material and fuel irradiation capability for the coming decades. This project is driven and funded by an international consortium gathering vendors, utilities and public stakeholders. This consortium has been set up in March 2007 when the construction began. The construction is in progress and the start of operation is scheduled for 2014. The JHR is a research

  16. ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO US MATERIAL TEST REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; Sebastien P. Teysseyre; Kurt L. Davis; Gordon Kohse; Yakov Ostrovsky; David M. Carpenter; Joy L. Rempe

    2015-04-01

    A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some test reactors outside the United States, such as the Halden Boiling Water Reactor (HBWR), have developed techniques to measure crack growth propagation during irradiation. The basic approach is to use a custom-designed compact loading mechanism to stress the specimen during irradiation, while the crack in the specimen is monitored in-situ using the Direct Current Potential Drop (DCPD) method. In 2012 the US Department of Energy commissioned the Idaho National Laboratory and the MIT Nuclear Reactor Laboratory (MIT NRL) to take the basic concepts developed at the HBWR and adapt them to a test rig capable of conducting in-pile IASCC tests in US Material Test Reactors. The first two and half years of the project consisted of designing and testing the loader mechanism, testing individual components of the in-pile rig and electronic support equipment, and autoclave testing of the rig design prior to insertion in the MIT Reactor. The load was applied to the specimen by means of a scissor like mechanism, actuated by a miniature metal bellows driven by pneumatic pressure and sized to fit within the small in-core irradiation volume. In addition to the loader design, technical challenges included developing robust connections to the specimen for the applied current and voltage measurements, appropriate ceramic insulating materials that can endure the LWR environment, dealing with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, due primarily to fast neutron damage, which change the specimen resistance without additional crack growth. The project culminated with an in

  17. Production test PTA-002, increased graphite temperature limit -- B, C and D Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Russell, A.

    1965-12-17

    The fundamental objective of the graphite temperature limit is to prevent excessive oxidation of the graphite moderator blocks with carbon dioxide and water vapor in the reactor atmosphere. Laboratory tests have shown that 10% uniform oxidation of graphite results in a loss in strength of approximately 50%. Production Test IP-725 was conducted at F Reactor for a period of six months at graphite temperatures approximately 50 and 100 C higher than the present graphite temperature limit of 650 C. The results from the F Reactor test suggest that an increase in the graphite temperature limit from 650 C to 700 C is technically feasible from the standpoint of oxidation of the graphite moderator with CO{sub 2}. Any significant additional increase was shown to lead to excessively high oxidation rates and is therefore not considered feasible. The objective of this test, therefore, is to extend the higher temperature investigations to B, C, and D Reactors. For the duration of this test, the graphite temperature limit will be increased from 650 C and 700 C, corresponding to an increase in the graphite stringer temperature limit from 735 C to 790 C. The test is expected to last for approximately six months but may be terminated early on any or all the reactors.

  18. Diagenetic Variations between Upper Cretaceous Outcrop and Deeply Buried Reservoir Chalks of the North Sea Area

    DEFF Research Database (Denmark)

    Hjuler, Morten Leth; Fabricius, Ida Lykke

    2007-01-01

    In the central North Sea Basin hydrocarbon-bearing chalks are deeply buried (2-3 km) whereas chalks in the rim areas are cropping out in the surrounding countries. The differing diagenetic histories between buried and outcrop chalk result in different rock properties, which is of great importance...... when simulating reservoir conditions using outcrop chalks as models. In general deeply buried reservoir chalks show significant overgrowth as witnessed by reshaping of particles together with strengthening of particle contacts. Most outcrop chalks are moderately affected with looser inter...

  19. Safety Issues at the DOE Test and Research Reactors. A Report to the U.S. Department of Energy.

    Science.gov (United States)

    National Academy of Sciences - National Research Council, Washington, DC. Commission on Physical Sciences, Mathematics, and Resources.

    This report provides an assessment of safety issues at the Department of Energy (DOE) test and research reactors. Part A identifies six safety issues of the reactors. These issues include the safety design philosophy, the conduct of safety reviews, the performance of probabilistic risk assessments, the reliance on reactor operators, the fragmented…

  20. Limitations of eddy current testing in a fast reactor environment

    Science.gov (United States)

    Wu, Tao; Bowler, John R.

    2016-02-01

    The feasibility of using eddy current probes for detecting flaws in fast nuclear reactor structures has been investigated with the aim of detecting defects immersed in electrically conductive coolant including under liquid sodium during standby. For the inspections to be viable, there is a need to use an encapsulated sensor system that can be move into position with the aid of visualization tools. The initial objective being to locate the surface to be investigated using, for example, a combination of electromagnetic sensors and sonar. Here we focus on one feature of the task in which eddy current probe impedance variations due to interaction with the external surface of a tube are evaluated in order to monitor the probe location and orientation during inspection.

  1. Markovian reliability analysis under uncertainty with an application on the shutdown system of the Clinch River Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Papazoglou, I A; Gyftopoulos, E P

    1978-09-01

    A methodology for the assessment of the uncertainties about the reliability of nuclear reactor systems described by Markov models is developed, and the uncertainties about the probability of loss of coolable core geometry (LCG) of the Clinch River Breeder Reactor (CRBR) due to shutdown system failures, are assessed. Uncertainties are expressed by assuming the failure rates, the repair rates and all other input variables of reliability analysis as random variables, distributed according to known probability density functions (pdf). The pdf of the reliability is then calculated by the moment matching technique. Two methods have been employed for the determination of the moments of the reliability: the Monte Carlo simulation; and the Taylor-series expansion. These methods are adopted to Markovian problems and compared for accuracy and efficiency.

  2. Bioavailability of Environmental Contaminants on the Upper Mississippi River National Wildlife and Fish Refuge Associated with Water Level Management on Upper Mississippi River Pool 8

    Data.gov (United States)

    US Fish and Wildlife Service, Department of the Interior — The U.S. Army Corps of Engineers coordinated with a variety of state and federal resource management agencies, the river transportation industry, and the public to...

  3. Sequential probability ratio tests for reactor signal validation and sensor surveillance applications

    Energy Technology Data Exchange (ETDEWEB)

    Humenik, K. (Maryland Univ., Baltimore, MD (USA)); Gross, K.C. (Argonne National Lab., IL (USA))

    1989-11-09

    This paper examines the properties of sequential probability ratio tests (SPRT's) and the application of these tests to nuclear power reactor operation. Recently SPRT's have been applied to delayed-neutron (DN) signal data analysis using actual reactor data from the Experimental Breeder Reactor-II, which is operated by Argonne National Laboratory. The implementation of this research as part of an expert system is described. Mathematical properties of the SPRT are investigated, and theoretical results are validated with tests that use DN-signal data taken from the EBR-II in Idaho. Variations of the basic SPRT and applications to general signal validation are also explored. 16 refs., 3 figs.

  4. Preliminary Investigations: Archaeology and Sediment Geomorphology, Navigation Pool 12, Upper Mississippi River. Volume II. Technical Documents and Site Data Sheets.

    Science.gov (United States)

    1981-01-01

    34The Grand River, Koshkonong, Green Bay, and Lake Winnebago PRases-- Eight Hundred Years of Eastern Wisconsin Onebta Prehistory:" Foreign Language ...data Hominid Paleontology North American PrehistoryNorth American Indians [ I(*indicates Graduate course) Adult Education Courses Taught: ISite Survey

  5. Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

    Energy Technology Data Exchange (ETDEWEB)

    M. Chen; CM Regan; D. Noe

    2006-01-09

    Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

  6. Comparison of the Becton Dickinson strand displacement amplification and Cobas Amplicor Roche PCR for the detection of Chlamydia trachomatis: pooling versus individual tests

    DEFF Research Database (Denmark)

    Bang, D; Angelsø, Lene; Schirakow, Bente

    2003-01-01

    The objective of the study was to examine the influence of pooling Chlamydia trachomatis specimens. We compared Becton Dickinson ProbeTec strand displacement amplification (SDA) with Cobas Amplicor Roche (PCR). With PCR as the standard, SDA performed equally well in single-sample testing....... For pooled PCR samples (compared to individual PCR), we found a sensitivity of 100% and a specificity of 98.9%. For pooled SDA tests (compared to individual SDA), we found a sensitivity of 86.5% and a specificity of 98.9%. Our conclusion is that 2-sucrose phosphate buffer (2-SP) can be used for individual...

  7. Postirradiation examination of recycle test elements from the Peach Bottom Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tiegs, T.N.; Long, E.L. Jr.

    1978-12-01

    The Recycle Test Elements were a series of tests of High-Temperature Gas-Cooled Reactor fuels irradiated in Core 2 of the Peach Bottom Unit 1 Reactor. They tested a wide variety of fissile and fertile fuel types of prime interest when the tests were designed. The fuel types included UO/sub 2/, UC/sub 2/, (2Th,U)O/sub 2/, (4Th,U)O/sub 2/, ThC/sub 2/, and ThO/sub 2/. The mixed thorium--uranium oxides and the pure thorium oxide were tested as Biso-coated particles only, while the others were tested as both Biso- and Triso-coated particles. The Biso coatings on the fissile kernels contained the fission products inadequately but on the fertile kernels they did so acceptably. The results from accelerated and real-time tests on the particle types agreed well.

  8. Half a decade of mini-pool nucleic acid testing: Cost-effective way for improving blood safety in India

    Directory of Open Access Journals (Sweden)

    Shivaram Chandrashekar

    2014-01-01

    Full Text Available Background and Objectives: It is well established that Nucleic acid testing (NAT reduces window phase of transfusion transmissible infections (TTI and helps improve blood safety. NAT testing can be done individually or in pools. The objectives of this study were to determine the utility, feasibility and cost effectiveness of an in-house minipool-NAT(MP-NAT. Materials and Methods: Blood donors were screened by history, tested by ELISA and sero-negative samples were subjected to an in-house NAT by using reverse transcriptase-polymerase chain reaction (RT-PCR. Testing was done in mini-pools of size eight (8. Positive pools were repeated with individual samples. Results: During the study period of Oct 2005-Sept 2010 (5 years all blood donors (n=53729 were screened by ELISA. Of which 469 (0.87% were positive for HIV-1, HBV or HCV. Sero-negative samples (n=53260 were screened by in-house MP-NAT. HIV-NAT yield was 1/53260 (n=1 and HBV NAT yield (n=2 was 1/26630. Conclusion: NAT yield was lower than other India studies possibly due to the lower sero-reactivity amongst our donors. Nevertheless it intercepted 9 lives including the components prepared. The in-house assay met our objective of improving blood safety at nominal cost and showed that it is feasible to set up small molecular biology units in medium-large sized blood banks and deliver blood within 24-48 hours. The utility of NAT (NAT yield will vary based on the donor population, the type of serological test used, the nature of kit employed and the sensitivity of NAT test used. The limitations of our in-house MP-NAT consisted of stringent sample preparation requirements, with labor and time involved. The benefits of our MP-NAT were that it acted as a second level of check for ELISA tests, was relatively inexpensive compared to ID-NAT and did not need sophisticated equipment.

  9. Cold Model Study and Commercial Test on Novel Vapor-Liquid Distributor of Hydroprocessing Reactor

    Institute of Scientific and Technical Information of China (English)

    Wang Shaobing; Zhang Zhanzhu; Wu Defei; Guo Qingming

    2007-01-01

    A novel vapor-liquid distributor was developed on the basis of sufficient study on the existing distributors applied in hydroprocessing reactors.The cold model test data showed that the fluid distribution performance of the novel vapor-liquid distributor was evidently better than the traditional one.Commercial tests of the new distributor were carried out in the 300 kt/a gas oil hydrotreating reactor at SINOPEC Changling Branch Company,showing that the new vapor-liquid distributor could improve the fluid distribution,promote the hydrotreating efficiency and lead to better performance than the traditional one.

  10. UMRS LTRMP 2010/11 LCU Mapping -- Pool 20

    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 2010...

  11. UMRS LTRMP 2010/11 LCU Mapping -- Pool 25

    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...

  12. UMRS LTRMP 2010/11 LCU Mapping -- Pool 22

    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 2010...

  13. UMRS LTRMP 2010/11 LCU Mapping -- Pool 19

    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 2010...

  14. UMRS LTRMP 2010/11 LCU Mapping -- Pool 21

    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...

  15. UMRS LTRMP 2010/11 LCU Mapping -- Pool 3

    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...

  16. UMRS LTRMP 2010/11 LCU Mapping -- Pool 6

    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...

  17. UMRS LTRMP 2010/11 LCU Mapping -- Pool 14

    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 2010...

  18. UMRS LTRMP 2010/11 LCU Mapping -- Pool 25

    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 2010...

  19. UMRS LTRMP 2010/11 LCU Mapping -- Pool 5

    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 2010...

  20. UMRS LTRMP 2010/11 LCU Mapping -- Pool 24

    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 2010...

  1. UMRS LTRMP 2010/11 LCU Mapping -- Pool 4

    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 2010...

  2. UMRS LTRMP 2010/11 LCU Mapping -- Pool 12

    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 2010...

  3. UMRS LTRMP 2010/11 LCU Mapping -- Pool 18

    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 2010...

  4. UMRS LTRMP 2010/11 LCU Mapping -- Pool 8

    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 2010...

  5. UMRS LTRMP 2010/11 LCU Mapping -- Pool 26

    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 2010...

  6. UMRS LTRMP 2010/11 LCU Mapping -- Pool 13

    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 2010...

  7. UMRS LTRMP 2010/11 LCU Mapping -- Pool 24

    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 -- Pool 4

    Data.gov (United States)

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

  9. UMRS LTRMP 2010/11 LCU Mapping -- Pool 5a

    Data.gov (United States)

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

  10. UMRS LTRMP 2010/11 LCU Mapping -- Pool 7

    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...

  11. UMRS LTRMP 2010/11 LCU Mapping -- Pool 18

    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...

  12. UMRS LTRMP 2010/11 LCU Mapping -- Pool 26

    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...

  13. UMRS LTRMP 2010/11 LCU Mapping -- Pool 19

    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...

  14. UMRS LTRMP 2010/11 LCU Mapping -- Pool 8

    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...

  15. UMRS LTRMP 2010/11 LCU Mapping -- Pool 10

    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...

  16. UMRS LTRMP 2010/11 LCU Mapping -- Pool 5

    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...

  17. UMRS LTRMP 2010/11 LCU Mapping -- Pool 22

    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...

  18. UMRS LTRMP 2010/11 LCU Mapping -- Pool 20

    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...

  19. UMRS LTRMP 2010/11 LCU Mapping -- Pool 09

    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...

  20. UMRS LTRMP 2010/11 LCU Mapping -- Pool 14

    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...

  1. UMRS LTRMP 2010/11 LCU Mapping -- Pool 12

    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...

  2. UMRS LTRMP 2010/11 LCU Mapping -- Pool 13

    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...

  3. Fusion Reactor and Fusion Reactor Materials:Concept Design of the ITER Test Blanket Modules

    Institute of Scientific and Technical Information of China (English)

    HUANGJinhua; LIZaixing; ZHUYukun; HUGang

    2003-01-01

    Performances required: prospect to be adopted in DEMO. Shielding for V.V. and TFC in ITER. Design principles: the peak temperature and stress should not exceed technical limits. The structure of test blanket modules (TBM) should be simple for easy fabrication, and TBM should be robust for reliability.

  4. Environmental Impact Study of the Northern Section of the Upper Mississippi River, Upper and Lower St. Anthony Falls Pool.

    Science.gov (United States)

    1973-11-01

    Icress Rorippa obtusa Obtuse yellow cress Unidentified sp. P I CUCURBITACEAE Sicyos angulatus Bur-cucumber1 CYPERACEAE Carex aenca Sedge Carex annectens...Bitter- White oak Little bluestem Reed-canary- Sedges flackbcrry nut White pine Nodding grama grass Milkweed Green ash hickory Sugar maple Northern Rice...Aster Cottonwood Hackbcrry Paper birch dropseed cutgrass Blue-joint Silver maple Ironwood Ironwood Hairy grama River sedge grass Slippery Bur oak Red

  5. High Temperature Stress Analysis on 61-pin Test Assembly for Reactor Core Sub-channel Flow Test

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dongwon; Kim, Hyungmo; Lee, Hyeongyeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In this study, a high temperature heat transfer and stress analysis of a 61-pin test fuel assembly scaled down from the full scale 217-pin sub-assembly was conducted. The reactor core subchannel flow characteristic test will be conducted to evaluate uncertainties in computer codes used for reactor core thermal hydraulic design. Stress analysis for a 61-pin fuel assembly scaled down from Prototype Generation IV Sodium-cooled Fast Reactor was conducted and structural integrity in terms of load controlled stress limits was conducted. In this study, The evaluations on load-controlled stress limits for a 61-pin test fuel assembly to be used for reactor core subchannel flow distribution tests were conducted assuming that the test assembly is installed in a Prototype Generation IV Sodium-cooled fast reactor core. The 61-pin test assembly has the geometric similarity on P/D and H/D with PGSFR and material of fuel assembly is austenitic stainless steel 316L. The stress analysis results showed that 4.05MPa under primary load occurred at mid part of the test assembly and it was shown that the value of 4.05Mpa was far smaller than the code allowable of 127MPa. , it was shown that the stress intensity due to due to primary load is very small. The stress analysis results under primary and secondary loads showed that maximum stress intensity of 84.08MPa occurred at upper flange tangent to outer casing and the value was well within the code allowable of 268.8MPa. Integrity evaluations based on strain limits and creep-fatigue damage are underway according to the elevated design codes.

  6. Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

    Energy Technology Data Exchange (ETDEWEB)

    Mourao, Rogerio P.; Leite da Silva, Luiz [Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte (Brazil); Miranda, Carlos A.; Mattar Neto, Miguel [Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo (Brazil); Quintana, Jose F.A.; Saliba, Roberto O. [Comision Nacional de Energia Atomica, Bariloche (Argentina); Novara, Oscar E. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

    2013-07-01

    Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away from reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)

  7. Advanced high-pressure bench-scale reactor for testing with hot corrosive gases

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, J.; Bachta, R.P.; Wangerow, J.R. (Inst. of Gas Technology, Chicago, IL (United States)); Mojtahedi, W.; Salo, K. (Enviropower Inc., Espoo (Finland))

    1994-01-01

    A bench-scale, high-pressure/high-temperature fluidized-bed reactor (HPTR) system is described that is capable of operating at a maximum temperature and pressure of 1,000 C and 30 bar in a corrosive atmosphere. The design of the unit is based on a double-shell balanced-pressure system. All the hot parts of the reactor that are wetted by the corrosive (and/or reactive) gases and the entire sampling line are constructed of inert material to prevent corrosion and loss of the reactant gases. The unit has been used for over 200 high-pressure hot coal gas desulfurization tests at 20 bars and up to 750 C without any experimental problem and with excellent sulfur balance, indicating that this reactor system is ideal for testing with reactive and corrosive gases at elevated pressures and temperatures.

  8. Prospects of Microbial Enhanced Oil Recovery  in Danish chalk rocks

    DEFF Research Database (Denmark)

    Rudyk, Svetlana Nikolayevna; Jørgensen, Leif Wagner; Bah Awasi, Ismail

    % of gas was produced in the presence of chalk in microbial solution and just 20% in the process of microbial fermentation with molasses (without chalk). The microbial solution improved the permeability of three chalk samples by 8.1%, 16.4% and 2658% respectively during 14 days of exposure having formed...... a big hole in the core sample in the latter case....

  9. Evaluating and planning the radioactive waste options for dismantling the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rule, K.; Scott, J.; Larson, S. [Princeton Plasma Physics Lab., NJ (United States)] [and others

    1995-12-31

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a kind tritium fusion research reactor, and is planned to be decommissioned within the next several years. This is the largest fusion reactor in the world and as a result of deuterium-tritum reactions is tritium contaminated and activated from 14 Mev neutrons. This presents many unusual challenges when dismantling, packaging and disposing its components and ancillary systems. Special containers are being designed to accommodate the vacuum vessel, neutral beams, and tritium delivery and processing systems. A team of experienced professionals performed a detailed field study to evaluate the requirements and appropriate methods for packaging the radioactive materials. This team focused on several current and innovative methods for waste minimization that provides the oppurtunmost cost effective manner to package and dispose of the waste. This study also produces a functional time-phased schedule which conjoins the waste volume, weight, costs and container requirements with the detailed project activity schedule for the entire project scope. This study and project will be the first demonstration of the decommissioning of a tritium fusion test reactor. The radioactive waste disposal aspects of this project are instrumental in demonstrating the viability of a fusion power reactor with regard to its environmental impact and ultimate success.

  10. Performance tests of a small hydrogen reactor based on Mg-Al pellets

    Energy Technology Data Exchange (ETDEWEB)

    Capurso, Giovanni, E-mail: giovanni.capurso@studenti.unipd.it [Dipartimento di Ingegneria Meccanica, Settore Materiali, Universita di Padova, via Marzolo 9, 35131 Padova (Italy); Agresti, Filippo [Dipartimento di Ingegneria Meccanica, Settore Materiali, Universita di Padova, via Marzolo 9, 35131 Padova (Italy); Russo, Sergio Lo [Dipartimento di Fisica and CNISM, Universita di Padova, via Marzolo 8, 35131 Padova (Italy); Maddalena, Amedeo; Principi, Giovanni [Dipartimento di Ingegneria Meccanica, Settore Materiali, Universita di Padova, via Marzolo 9, 35131 Padova (Italy); Cavallari, Andrea; Guardamagna, Cristina [ERSE s.p.a., via Rubattino 54, 20134 Milano (Italy)

    2011-09-15

    On the basis of a previously acquired experience on scaling up issues concerning the use of magnesium hydride as a base material for solid-state hydrogen storage, a small reactor was designed and tested in different operating conditions. It contains about 10 g of catalyzed magnesium hydride powder mixed with 5 wt.% aluminium powder and pressed in the form of cylindrical pellets and the heat flow is managed by means of an oil circulation system. Carbon paper is used to ensure good heat conductivity between the pellets and the inner wall of the reactor and between one pellet and another. A number of hydrogen absorption and desorption cycles at different temperatures and pressures was carried out to compare the behaviour of the small reactor with the laboratory data obtained on small amounts (fractions of grams) of powdered and pelletized samples. Data acquisition for gas flow, pressure and temperature in different positions of the reactor allow a good understanding of internal dynamics. The results in terms of hydrogen absorption/desorption kinetics and of stability to ongoing cycles are stimulating, so that the tested small reactor can be considered as a basic element for further studies and improvements.

  11. Operation, test, research and development of the high temperature engineering test reactor (HTTR). FY1999-2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    The HTTR (High Temperature Engineering Test Reactor) with the thermal power of 30 MW and the reactor outlet coolant temperature of 850/950 degC is the first high temperature gas-cooled reactor (HTGR) in Japan, which uses coated fuel particle, graphite for core components, and helium gas for primary coolant. The HTTR, which locates at the south-west area of 50,000 m{sup 2} in the Oarai Research Establishment, had been constructed since 1991 before accomplishing the first criticality on November 10, 1998. Rise to power tests of the HTTR started in September, 1999 and the rated thermal power of 30 MW and the reactor outlet coolant temperature of 850 degC was attained in December 2001. JAERI received the certificate of pre-operation test, that is, the commissioning license for the HTTR in March 2002. This report summarizes operation, tests, maintenance, radiation control, and construction of components and facilities for the HTTR as well as R and Ds on HTGRs from FY1999 to 2001. (author)

  12. What are the governing processes during low-flows in a chalk catchment?

    Science.gov (United States)

    Lubega Musuuza, Jude; Coxon, Gemma; Hutton, Chris; Howden, Nicholas; Woods, Ross; Freer, Jim; Wagener, Thorsten

    2016-04-01

    Low flows are important because they lead to the prioritisation of different consumptive water usages, imposition of restrictions and bans, raising of water tariffs and higher production costs to industry. The partitioning of precipitation into evaporation, storage and runoff depends on the local variability in meteorological variables and site-specific characteristics e.g., topography, soils and vegetation. The response of chalk catchments to meteorological forcing especially precipitation is of particular interest because of the preferential flow through the weathered formation. This makes the observed stream discharge groundwater-dominated and hence, out of phase with precipitation. One relevant question is how sensitive the low flow characteristics of such a chalk catchment is to changes in climate and land use. It is thus important to understand all the factors that control low stream discharge periods. In this study we present the results from numerical sensitivity analysis experiments performed with a detailed physically-based model on the Kennet, a sub-catchment of the River Thames, in the UK during the historical drought years of the 1970's.

  13. Influence of effective stress coefficient on mechanical failure of chalk

    DEFF Research Database (Denmark)

    Alam, Mohammad Monzurul; Fabricius, Ida Lykke; Hjuler, M.L.

    2012-01-01

    , as this process could affect the grain contact cement. If this happens, the effective stress at the grain contacts in a reservoir will change according to the effective stress principle of Biot. In a p′-q space for failure analysis, we observed that a higher effective stress coefficient reduces the elastic region...... and vice versa. However, as the effective stress working on the rock decreases with increased effective stress coefficient, the reduction of elastic region will have less effect on pore collapse strength if we consider the change in the effective stress coefficient. This finding will help estimate a more......The Effective stress coefficient is a measure of how chalk grains are connected with each other. The stiffness of chalk may decrease if the amount of contact cements between the grains decreases, which may lead to an increase of the effective stress coefficient. We performed CO2 injection in chalk...

  14. Chemical and Mechanical processes during burial diagenesis of chalk

    DEFF Research Database (Denmark)

    Borre, Mai Kirstine; Lind, Ida

    1998-01-01

    or larger influence on the textural development. In the chalk interval below, compaction is not the only porosity reducing agent but it has a larger influence on texture than concurrent recrystallization. Below 850 m grain-bridging cementation becomes important resulting in a lithified limestone below 1100......Burial diagenesis of chalk is a combination of mechanical compaction and chemical recrystallization as well as cementation. We have predicted the characteristic trends in specific surface resulting from these processes. The specific surface is normally measured by nitrogen adsorption but is here...... in the Pacific, where a > 1 km thick package of chalk facies sediments accumulated from the Cretaceous to the present. In the upper 200-300 m the sediment is unconsolidated carbonate ooze, throughout this depth interval compaction is the principal porosity reducing agent, but recrystallization has an equal...

  15. Warm Water Oxidation Verification - Scoping and Stirred Reactor Tests

    Energy Technology Data Exchange (ETDEWEB)

    Braley, Jenifer C.; Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2011-06-15

    Scoping tests to evaluate the effects of agitation and pH adjustment on simulant sludge agglomeration and uranium metal oxidation at {approx}95 C were performed under Test Instructions(a,b) and as per sections 5.1 and 5.2 of this Test Plan prepared by AREVA. (c) The thermal testing occurred during the week of October 4-9, 2010. The results are reported here. For this testing, two uranium-containing simulant sludge types were evaluated: (1) a full uranium-containing K West (KW) container sludge simulant consisting of nine predominant sludge components; (2) a 50:50 uranium-mole basis mixture of uraninite [U(IV)] and metaschoepite [U(VI)]. This scoping study was conducted in support of the Sludge Treatment Project (STP) Phase 2 technology evaluation for the treatment and packaging of K-Basin sludge. The STP is managed by CH2M Hill Plateau Remediation Company (CHPRC) for the U.S. Department of Energy. Warm water ({approx}95 C) oxidation of sludge, followed by immobilization, has been proposed by AREVA and is one of the alternative flowsheets being considered to convert uranium metal to UO{sub 2} and eliminate H{sub 2} generation during final sludge disposition. Preliminary assessments of warm water oxidation have been conducted, and several issues have been identified that can best be evaluated through laboratory testing. The scoping evaluation documented here was specifically focused on the issue of the potential formation of high strength sludge agglomerates at the proposed 95 C process operating temperature. Prior hydrothermal tests conducted at 185 C produced significant physiochemical changes to genuine sludge, including the formation of monolithic concretions/agglomerates that exhibited shear strengths in excess of 100 kPa (Delegard et al. 2007).

  16. Digital System Reliability Test for the Evaluation of safety Critical Software of Digital Reactor Protection System

    Directory of Open Access Journals (Sweden)

    Hyun-Kook Shin

    2006-08-01

    Full Text Available A new Digital Reactor Protection System (DRPS based on VME bus Single Board Computer has been developed by KOPEC to prevent software Common Mode Failure(CMF inside digital system. The new DRPS has been proved to be an effective digital safety system to prevent CMF by Defense-in-Depth and Diversity (DID&D analysis. However, for practical use in Nuclear Power Plants, the performance test and the reliability test are essential for the digital system qualification. In this study, a single channel of DRPS prototype has been manufactured for the evaluation of DRPS capabilities. The integrated functional tests are performed and the system reliability is analyzed and tested. The results of reliability test show that the application software of DRPS has a very high reliability compared with the analog reactor protection systems.

  17. Experimental determination of nuclear parameters for RP-0 reactor core; Determinacion experimental de los parametros nucleares para el nucleo tipo MTR del reactor nuclear RP-0

    Energy Technology Data Exchange (ETDEWEB)

    Cajacuri, Rafael A. [Sao Paulo Univ., SP (Brazil). Inst. de Fisica

    2000-07-01

    In the nuclear reactor for investigations RP-0 which is in Lima, Peru, that is a open pool class reactor with 1 to 10 watts of power and as a nuclear fuel uranium 238 enriched to 20% constituted by elements of Material Testing Reactor fuel class. This has reflectors of graphite and moderator of water demineralized. In 1996/1997 was measured in this reactor the following parameters: position of the control bar that make critic the reactor, critic height of moderator, excess of reactivity of the nucleus, parameter of reactivity for vacuum, parameter of reactivity for temperature, reactivity of its control bar, levels of doses in the reactor. (author)

  18. Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

    2010-01-01

    A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

  19. Abbreviated sampling and analysis plan for planning decontamination and decommissioning at Test Reactor Area (TRA) facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-10-01

    The objective is to sample and analyze for the presence of gamma emitting isotopes and hazardous constituents within certain areas of the Test Reactor Area (TRA), prior to D and D activities. The TRA is composed of three major reactor facilities and three smaller reactors built in support of programs studying the performance of reactor materials and components under high neutron flux conditions. The Materials Testing Reactor (MTR) and Engineering Test Reactor (ETR) facilities are currently pending D/D. Work consists of pre-D and D sampling of designated TRA (primarily ETR) process areas. This report addresses only a limited subset of the samples which will eventually be required to characterize MTR and ETR and plan their D and D. Sampling which is addressed in this document is intended to support planned D and D work which is funded at the present time. Biased samples, based on process knowledge and plant configuration, are to be performed. The multiple process areas which may be potentially sampled will be initially characterized by obtaining data for upstream source areas which, based on facility configuration, would affect downstream and as yet unsampled, process areas. Sampling and analysis will be conducted to determine the level of gamma emitting isotopes and hazardous constituents present in designated areas within buildings TRA-612, 642, 643, 644, 645, 647, 648, 663; and in the soils surrounding Facility TRA-611. These data will be used to plan the D and D and help determine disposition of material by D and D personnel. Both MTR and ETR facilities will eventually be decommissioned by total dismantlement so that the area can be restored to its original condition.

  20. Test Plan for the Boiling Water Reactor Dry Cask Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-11-01

    The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the

  1. Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meeting have been cataloged separately.

  2. Tests of Lorentz and CPT Violation in the Medium Baseline Reactor Antineutrino Experiment

    CERN Document Server

    Li, Yu-Feng

    2014-01-01

    Tests of Lorentz and CPT violation in the medium baseline reactor antineutrino experiment are presented in the framework of the Standard Model Extension (SME). Both the spectral distortion and sidereal variation are employed to derive the limits of Lorentz violation (LV) coefficients. We do the numerical analysis of the sensitivity of LV coefficients by taking the Jiangmen Underground Neutrino Observatory (JUNO) as an illustration, which can improve the sensitivity by more than two orders of magnitude compared with the current limits from reactor antineutrino experiments.

  3. Development and Testing of a High Capacity Plasma Chemical Reactor in the Ukraine

    Energy Technology Data Exchange (ETDEWEB)

    Reilly, Raymond W.

    2012-07-30

    This project, Development and Testing of a High Capacity Plasma Chemical Reactor in the Ukraine was established at the Kharkiv Institute of Physics and Technology (KIPT). The associated CRADA was established with Campbell Applied Physics (CAP) located in El Dorado Hills, California. This project extends an earlier project involving both CAP and KIPT conducted under a separate CRADA. The initial project developed the basic Plasma Chemical Reactor (PCR) for generation of ozone gas. This project built upon the technology developed in the first project, greatly enhancing the output of the PCR while also improving reliability and system control.

  4. Mechanical properties of reactor pressure vessel steels studied by static and dynamic torsion tests

    Science.gov (United States)

    Munier, A.; Maamouri, M.; Schaller, R.; Mercier, O.

    1993-06-01

    Internal friction measurements and torsional plastic deformation tests have been performed in reactor pressure vessel steels (unirradiated, irradiated and irradiated/annealed specimens). The results of these experiments have been interpreted with help of transmission electron microscopy observations (conventional and in situ). It is shown how the interactions between screw dislocations and obstacles (Peierls valleys, impurities and precipitates) could explain the low temperature hardening and the irradiation embrittlement of ferritic steels. In addition, it appears that the nondestructive internal friction technique could be used advantageously to follow the evolution of the material properties under irradiation, as for instance the irradiation embrittlement of the reactor pressure vessel steels.

  5. Pilot-plant testing of magnetic filters for the N-Reactor primary cooling circuit

    Energy Technology Data Exchange (ETDEWEB)

    Emory, B.B.

    1982-01-01

    Data obtained during the laboratory loop test program using the high power HGMF indicates that removal efficiency for /sup 60/Co and subsequently the bulk of the crud, will be greater than 90% at field strength above .1 Tesla for the expanded metal mesh matrix. However, since /sup 54/Mn seems to exhibit paramagnetic behavior and the possibility of quantities of alpha iron forming during reactor shut down from oxygen inleakage, a field strength of .5 to 1 Tesla may be more appropriate for a full scale on-reactor installation. Crud loading of 50 gm per kg of matrix weight are readily obtainable and up to twice that amount has been reached.

  6. Evaluation of the Shielding Characteristics Test around the Reactor Core in the Prototype Fbr Monju

    Science.gov (United States)

    Usami, Shin; Suzuoki, Zenro; Deshimaru, Takehide; Nakashima, Fumiaki; Hikichi, Takuo

    2003-06-01

    In Monju, shielding measurements were made around the reactor core as a part of the system start-up tests in order to evaluate the design margins of the shielding performance, to demonstrate the validity of the shielding analysis method, and to acquire basic data for use in future FBR design. The measured reaction rates have been obtained radially from the core to the in-vessel storage rack and axially to the reactor vessel upper plenum. The measured values (E) were compared with the calculated values (C) obtained with the FBR shielding analysis system on the basis of the nuclear data library JENDL-3.2. Based upon these results, the design margins around the reactor core have been examined.

  7. Two-Compartment Photoelectrochemical Reactors Tested under various solar Light Concentration ratios

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez-Ibanez, P.; Malato, S. [Plataforma Solar de Almeria. CIEMAT (Spain)

    1999-07-01

    A new type of photo reactor, made of a cylindrical photo anode placed around an inner comportment, has been adapted to the compound parabolic (CPC's) and to the parabolic trough (PTC, Helioman) solar collectors. The photoelectrochemical performances of such two-compartment photo reactors are noticeably improved with respect to those previously obtained with photo reactors having flat photoanodes and a single compartment. The abatement of model pollutants shows up to threefold higher organic oxidation rates compared to Ti O{sub 2} slurries tested in the same experimental conditions. Clearly, charge separation is much better when an external electrochemical bias is applied to Ti/Ti O{sub 2} photoanodes under irradiation. (Author) 8 refs.

  8. Monitoring and Analysis of In-Pile Phenomena in Advanced Test Reactor using Acoustic Telemetry

    Energy Technology Data Exchange (ETDEWEB)

    Agarwal, Vivek [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Human Factors, Controls, and Statistics; Smith, James A. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Fuel Performance and Design; Jewell, James Keith [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Fuel Performance and Design

    2015-02-01

    The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. A number of research programs are developing acoustic-based sensing approach to take advantage of the acoustic transmission properties of reactor cores. Idaho National Laboratory has installed vibroacoustic receivers on and around the Advanced Test Reactor (ATR) containment vessel to take advantage of acoustically telemetered sensors such as thermoacoustic (TAC) transducers. The installation represents the first step in developing an acoustic telemetry infrastructure. This paper presents the theory of TAC, application of installed vibroacoustic receivers in monitoring the in-pile phenomena inside the ATR, and preliminary data processing results.

  9. Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

    Science.gov (United States)

    Garber, Anne E.; Dickens, Ricky E.

    2011-01-01

    The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

  10. Proceedings of the 1990 International Meeting on Reduced Enrichment for Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The global effort to reduce, and possibly, eliminate the international traffic in highly-enriched uranium caused by its use in research reactors requires extensive cooperation and free exchange of information among all participants. To foster this free exchange of information, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the thirteenth of a series which began in 1978. The common effort brought together, past, a large number of specialists from many countries. On hundred twenty-three participants from 26 countries, including scientists, reactor operators, and personnel from commercial fuel suppliers, research centers, and government organizations, convened in Newport, Rhode Island to discuss their results, their activities, and their plans relative to converting research reactors to low-enriched fuels. As more and more reactors convert to the use of low-enriched uranium, the emphasis of our effort has begun to shift from research and development to tasks more directly related to implementation of the new fuels and technologies that have been developed, and to refinements of those fuels and technologies. It is appropriate, for this reason, that the emphasis of this meeting was placed on safety and on conversion experiences. This individual papers in this report have been cataloged separately.

  11. Relationship of weed shiner and young-of-year bluegill and largemouth bass abundance to submersed aquatic vegetation in Navigation Pools 4, 8, and 13 of the Upper Mississippi River, 1998-2012

    Science.gov (United States)

    DeLain, Steven A.; Popp, Walter A.

    2014-01-01

    Aquatic vegetation provides food resources and shelter for many species of fish. This study found a significant relationship between increases in submersed aquatic vegetation (SAV) in four study reaches of the Upper Mississippi River (UMR) and increases in catch-per-unit-effort (CPUE) of weed shiners (Notropis texanus) and age-0 bluegills (Lepomis macrochirus) and largemouth bass (Micropterus salmoides) when all of the study reaches were treated collectively using Long Term Resource Monitoring Program (LTRMP) vegetation and fish data for 1998–2012. The selected fishes were more abundant in study reaches with higher SAV frequencies (Pool 8 and Lower Pool 4) and less abundant in reaches with lower SAV frequencies (Pool 13 and Upper Pool 4). When each study reach was examined independently, the relationship between SAV frequency and CPUE of the three species was not significant in most cases, the primary exception being weed shiners in Lower Pool 4. Results of this study indicate that the prevalence of SAV does affect relative abundance of these vegetation-associated fish species. However, the poor annual relationship between SAV frequency and age-0 relative abundance in individual study reaches indicates that several other factors also govern age-0 abundance. The data indicate that there may be a SAV frequency threshold in backwaters above which there is not a strong relationship with abundance of these fish species. This is indicated by the high annual CPUE variability of the three selected fishes in backwaters of Pool 8 and Lower Pool 4 when SAV exceeded certain frequencies.

  12. TESTING OF THE RADBALL TECHNOLOGY AT SAVANNAH RIVER NATIONAL LABORATORY

    Energy Technology Data Exchange (ETDEWEB)

    Farfan, E.; Foley, T.

    2010-02-10

    The United Kingdom's National Nuclear Laboratory (NNL) has developed a remote, nonelectrical, radiation-mapping device known as RadBall (patent pending), which offers a means to locate and quantify radiation hazards and sources within contaminated areas of the nuclear industry. Positive results from initial deployment trials in nuclear waste reprocessing plants at Sellafield in the United Kingdom and the anticipated future potential use of RadBall throughout the U.S. Department of Energy Complex have led to the NNL partnering with the Savannah River National Laboratory (SRNL) to further test, underpin, and strengthen the technical performance of the technology. The study completed at SRNL addresses key aspects of the testing of the RadBall technology. The first set of tests was performed at Savannah River Nuclear Solutions Health Physics Instrument Calibration Laboratory (HPICL) using various gamma-ray sources and an x-ray machine with known radiological characteristics. The objective of these preliminary tests was to identify the optimal dose and collimator thickness. The second set of tests involved a highly contaminated hot cell. The objective of this testing was to characterize a hot cell with unknown radiation sources. The RadBall calibration experiments and hot cell deployment were successful in that for each trial radiation tracks were visible. The deployment of RadBall can be accomplished in different ways depending on the size and characteristics of the contaminated area (e.g., a hot cell that already has a crane/manipulator available or highly contaminated room that requires the use of a remote control device with sensor and video equipment to position RadBall). This report also presents SRNL-designed RadBall accessories for future RadBall deployment (a harness, PODS, and robot).

  13. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    Energy Technology Data Exchange (ETDEWEB)

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition

  14. The 5th surveillance testing for Kori unit 2 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-03-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules V, R, P, T and N are 2.837E+18, 1.105E+19, 2.110E+19, 3.705E+19 and 4.831E+19n/cm{sup 2}, respectively. The bias factor, the ratio of measurement/calculation, was 0.918 for the 1st through 5th testing and the calculational uncertainty, 11.6% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.898E+19n/cm{sup 2} based on the end of 15th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 4.203E+19, 5.232E+19, 6.262E+19 and 7.291E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 49 refs., 35 figs., 48 tabs. (Author)

  15. The 4th surveillance testing for Kori unit 3 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-10-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 4.983E+18, 1.641E+19, 3.158E+19, and 4.469E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.840 for the 1st through 4th testing and the calculational uncertainty, 12% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.362E+19n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.481E+19, 4.209E+19, 5.144E+19 and 5.974E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 41 tabs. (Author)

  16. Psychometric evaluation of the EORTC computerized adaptive test (CAT) fatigue item pool

    DEFF Research Database (Denmark)

    Petersen, Morten Aa; Giesinger, Johannes M; Holzner, Bernhard;

    2013-01-01

    Fatigue is one of the most common symptoms associated with cancer and its treatment. To obtain a more precise and flexible measure of fatigue, the EORTC Quality of Life Group has developed a computerized adaptive test (CAT) measure of fatigue. This is part of an ongoing project developing a CAT v...

  17. Field Tests and Simulation of Lion-Head River Bridge

    Directory of Open Access Journals (Sweden)

    Yao-Min Fang

    2007-01-01

    Full Text Available Lion-Head River Bridge is a twin bridge in parallel position. The east-bounded was designed and constructed as a traditional prestress concrete box girder bridge with pot bearings; and the west-bounded was installed with seismic isolation devices of lead rubber bearings. The behavior of the isolated bridge is compared with that of the traditional bridge through several field tests including the ambient vibration test, the force vibration test induced by shakers, the free vibration test induced by a push and fast release system, and the truck test. The bridges suffered from various extents of damage due to the Chi-Chi and the Chi-I earthquakes of great strength during the construction and had been retrofitted. The damage was reflected by the change of the bridges' natural frequencies obtained from the ambient vibration tests. The models of the two bridges are simulated by the finite element method based on the original design drawings. Soil-structure interaction was also scrutinized in this study. The simulation was then modified based on the results from the field tests. Dynamic parameters of bridges are identified and compared with those from theoretical simulation. The efficiency is also verified to be better for an isolated bridge.

  18. Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR)

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; Gerry L. McCormick; Shannon J. Corrigan

    2010-06-01

    2010 International Congress on Advances in Nuclear Power Plants (ICAPP’10) ANS Annual Meeting Imbedded Topical San Diego, CA June 13 – 17, 2010 Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR) Author: A. Joseph Palmer, Mechanical Engineer, Irradiation Test Programs, 208-526-8700, Joe.Palmer@INL.gov Affiliation: Idaho National Laboratory P.O. Box 1625, MS-3840 Idaho Falls, ID 83415 INL/CON-10-17680 ABSTRACT Most test reactors are equipped with shuttle facilities (sometimes called rabbit tubes) whereby small capsules can be inserted into the reactor and retrieved during power operations. With the installation of Hydraulic Shuttle Irradiation System (HSIS) this capability has been restored to the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The general design and operating principles of this system were patterned after the hydraulic rabbit at Oak Ridge National Laboratory’s (ORNL) High Flux Isotope Reactor (HFIR), which has operated successfully for many years. Using primary coolant as the motive medium the HSIS system is designed to simultaneously transport fourteen shuttle capsules, each 16 mm OD x 57 mm long, to and from the B-7 position of the reactor. The B-7 position is one of the higher flux positions in the reactor with typical thermal and fast (>1 Mev) fluxes of 2.8E+14 n/cm2/sec and 1.9E+14 n/cm2/sec respectively. The available space inside each shuttle is approximately 14 mm diameter x 50 mm long. The shuttle containers are made from titanium which was selected for its low neutron activation properties and durability. Shuttles can be irradiated for time periods ranging from a few minutes to several months. The Send and Receive Station (SRS) for the HSIS is located 2.5 m deep in the ATR canal which allows irradiated shuttles to be easily moved from the SRS to a wet loaded cask, or transport pig. The HSIS system first irradiated (empty) shuttles in September 2009 and has since completed

  19. Nano sized clay detected on chalk particle surfaces

    DEFF Research Database (Denmark)

    Skovbjerg, Lone; Hassenkam, Tue; Makovicky, Emil

    2012-01-01

    that in calcite saturated water, both the polar and the nonpolar functional groups adhere to the nano sized clay particles but not to calcite. This is fundamentally important information for the development of conceptual and chemical models to explain wettability alterations in chalk reservoirs...

  20. Quantitative 1D saturation profiles on chalk by NMR

    DEFF Research Database (Denmark)

    Olsen, Dan; Topp, Simon; Stensgaard, Anders;

    1996-01-01

    Quantitative one-dimensional saturation profiles showing the distribution of water and oil in chalk core samples are calculated from NMR measurements utilizing a 1D CSI spectroscopy pulse sequence. Saturation profiles may be acquired under conditions of fluid flow through the sample. Results reveal...

  1. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for AHTRs

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Quiping [The Ohio State Univ., Columbus, OH (United States); Sun, Xiaodong [The Ohio State Univ., Columbus, OH (United States); Chtistensen, Richard [The Ohio State Univ., Columbus, OH (United States); Blue, Thomas [The Ohio State Univ., Columbus, OH (United States); Yoder, Graydon [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-05-08

    The principal objective of this research is to test and model the heat transfer performance and reliability of the Direct Reactor Auxiliary Cooling System (DRACS) for AHTRs. In addition, component testing of fluidic diodes is to be performed to examine the performance and viability of several existing fluidic diode designs. An extensive database related to the thermal performance of the heat exchangers involved will be obtained, which will be used to benchmark a computer code for the DRACS design and to evaluate and improve, if needed, existing heat transfer models of interest. The database will also be valuable for assessing the viability of the DRACS concept and benchmarking any related computer codes in the future. The experience of making a liquid fluoride salt test facility available, with lessons learned, will greatly benefit the development of the Fluoride Salt-cooled High-temperature Reactor (FHR) and eventually the AHTR programs.

  2. In-reactor tests of the nuclear light bulb rocket concept

    Science.gov (United States)

    Gauntt, R. O.; Slutz, S. A.; Latham, T. S.; Roman, W. C.; Rogers, R. J.

    1992-07-01

    An overview is given of the closed-cycle Gas Core Nuclear Rocket outlining scenarios for its use in short-duration Mars missions and results of Nuclear Light Bulb (NLB) tests. Isothermal and nonnuclear tests are described which confirmed the fundamental concepts behind the NLB. NLB reference-engine performance characteristics are given for hypothetical engines that could be used for manned Mars missions. Vehicle/propulsion sizing is based on a Mars mission with three trans-Mars impulse burns, capture and escape burns, and a total mission duration of 600 days. The engine would have a specific impulse of 1870 seconds, a 412-kN thrust, and a thrust/weight ratio of 1.3. Reactor tests including small-scale in-reactor tests are shown to be prerequisites for studying: (1) fluid mechanical confinement of the gaseous nuclear fuel; (2) buffer gas separation and circulation; and (3) the minimization of transparent wall-heat loading. The reactor tests are shown to be critical for establishing the feasibility of the NLB concept.

  3. The 5th surveillance testing for Kori unit 1 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-08-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed primarily by Korea Atomic Energy Research Institute and Westinhouse corporation partially involved in testing and calculation data evaluation in order to obtain reliable test result. Fast neutron fluences for capsule V, T, S, R and P were 5.087E+18, 1.115E+19, 1.228E+19, 2.988E+19, and 3.938E+19n/cm2, respectively. The bias factor, the ratio of calculation/measurement, was 0.940 for the 1st through 5th testing and the calculational uncertainty, 7% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.9846E+19n/cm{sup 2} based on the end of 17th fuel cycle and it was predicted that the fluences of vessel inside surface at 24, 32, 40 and 48EFPY would reach 3.0593E+19, 4.0695E+19, 5.0797E+19 and 6.0900E+19n/cm{sup 2} based on the current calculation. PTS analysis for Kori unit 1 showed that 27.93EFPY was the threshold value for 300 deg F requirement. 71 refs., 33 figs., 52 tabs. (Author)

  4. Light Water Reactor Sustainability Program Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study

    Energy Technology Data Exchange (ETDEWEB)

    Curtis Smith; David Schwieder; Cherie Phelan; Anh Bui; Paul Bayless

    2012-08-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about LWR design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the RISMC Pathway R&D is to support plant decisions for risk-informed margins management with the aim to improve economics, reliability, and sustain safety of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. This report describes the RISMC methodology demonstration where the Advanced Test Reactor (ATR) was used as a test-bed for purposes of determining safety margins. As part of the demonstration, we describe how both the thermal-hydraulics and probabilistic safety calculations are integrated and used to quantify margin management strategies.

  5. Ekofisk chalk: core measurements, stochastic reconstruction, network modeling and simulation

    Energy Technology Data Exchange (ETDEWEB)

    Talukdar, Saifullah

    2002-07-01

    This dissertation deals with (1) experimental measurements on petrophysical, reservoir engineering and morphological properties of Ekofisk chalk, (2) numerical simulation of core flood experiments to analyze and improve relative permeability data, (3) stochastic reconstruction of chalk samples from limited morphological information, (4) extraction of pore space parameters from the reconstructed samples, development of network model using pore space information, and computation of petrophysical and reservoir engineering properties from network model, and (5) development of 2D and 3D idealized fractured reservoir models and verification of the applicability of several widely used conventional up scaling techniques in fractured reservoir simulation. Experiments have been conducted on eight Ekofisk chalk samples and porosity, absolute permeability, formation factor, and oil-water relative permeability, capillary pressure and resistivity index are measured at laboratory conditions. Mercury porosimetry data and backscatter scanning electron microscope images have also been acquired for the samples. A numerical simulation technique involving history matching of the production profiles is employed to improve the relative permeability curves and to analyze hysteresis of the Ekofisk chalk samples. The technique was found to be a powerful tool to supplement the uncertainties in experimental measurements. Porosity and correlation statistics obtained from backscatter scanning electron microscope images are used to reconstruct microstructures of chalk and particulate media. The reconstruction technique involves a simulated annealing algorithm, which can be constrained by an arbitrary number of morphological parameters. This flexibility of the algorithm is exploited to successfully reconstruct particulate media and chalk samples using more than one correlation functions. A technique based on conditional simulated annealing has been introduced for exact reproduction of vuggy

  6. Proving test on the seismic reliability of nuclear power plant: PWR reactor containment vessel

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Hiroshi; Yoshikawa, Teiichi; Ohno, Tokue; Yoshikawa, Eiji.

    1989-01-01

    Seismic reliability proving tests of nuclear power plant facilities are carried out by the Nuclear Power Engineering Test Center, using the large-scale, high-performance vibration table of Tadotsu Engineering Laboratory, and sponsored by the Ministry of International Trade and Industry. In 1982, the seismic reliability proving test of a PWR containment vessel was conducted using a test component of reduced scale 1/3.7. As a result of this test, the test component proved to have structural soundness against earthquakes, and at the same time its stable function was proved by leak tests which were carried out before and after the vibration test. In 1983, the detailed analysis and evaluation of these test results were carried out, and the analysis methods for evaluating strength against earthquakes were established. The seismic analysis and evaluation on the actual containment vessel were then performed using these analysis methods, and the safety and reliability of the PWR reactor containment vessel were confirmed.

  7. Relationship Between Persistent Heavy Rain Events in the Huaihe River Valley and the Distribution Pattern of Convective Activities in the Tropical Western Pacific Warm Pool

    Institute of Scientific and Technical Information of China (English)

    BAO Ming

    2008-01-01

    Using daily outgoing long-wave radiation(OLR)data from the National Oceanic and Atmospheric Ad-ministration(NOAA)and the National Center for Environmental Prediction/National Center for Atmo-spheric Research(NCEP/NCAR)reanalysis data of geopotential height fields for 1979-2006,the relation-ship between persistent heavy rain events(PHREs)in the Huaihe River valley(HRV)and the distribution pattern of convective activity in the tropical western Pacific warm pool(WPWP)is investigated.Based on nine cases of PHREs in the HRV,common characteristics of the West Pacific subtropical high(WPSH)show that the northern edge of the WPSH continues to lie in the HRV and is associated with the persistent "north weak south strong" distribution pattern of convective activities in the WPWP.Composite analysis of OLR leading the circulation indicates that the response of the WPSH to OLR anomaly patterns lags by about 1-2 days.In order to explain the reason for the effects of the distribution pattern of convective activities in the WPWP on the persistent northern edge of the WPSH in the HRV,four typical persistent heavy and light rain events in the Yangtze River valley(YRV)are contrasted with the PHREs in the HRV.The comparison indicates that when the distribution pattern of the convective activities anomaly behaves in a weak(strong)manner across the whole WPWP, persistent heavy(light)rain tends to occur in the YRV.When the distribution pattern of the convective activities anomaly behaves according to the "north weak south strong" pattern in the WPWP,persistent heavy rain tends to occur in the HRV.The effects of the "north weak south strong" distribution pattern of convective activities on PHREs in the HRV are not obvious over the seasonal mean timescale,perhaps due to the non-extreme status of convective activities in the WPWP.

  8. You can pool faecal samples from individual pigs to test for Porcine Circovirus Type 2 and Lawsonia intracellularis using real-time PCRs

    DEFF Research Database (Denmark)

    Holyoake, Patricia K.; Hjulsager, Charlotte Kristiane; Larsen, Lars Erik;

    Introduction Real-time PCR tests have been developed to detect and quantify Porcine Circovirus type 2 (PCV2) and Lawsonia intracellularis in pigs’ faeces. Pooling of individual faecal samples is often used to reduce the costs of diagnostic testing. The objective of this study was to evaluate any ...

  9. Nucleate pool boiling investigation on a silicon test section with micro-fabricated cavities

    Energy Technology Data Exchange (ETDEWEB)

    Sanna, A.; Kenning, D.B.R.; Karayiannis, T.G. [Brunel University, Uxbridge (United Kingdom); Hutter, C.; Sefiane, K. [University of Edinburgh (United Kingdom); Nelson, R.A. [Los Alamos National Laboratory, Los Alamos, NM (United States)

    2009-07-01

    The basic mechanisms of nucleate boiling are still not completely understood, in spite of the many numerical and experimental studies dedicated to the topic. The use of a hybrid code allows reasonable computational times for simulations of a solid plate with a large population of artificial micro-cavities with fixed distribution. This paper analyses the guidelines for the design, through numerical simulations, of the location and sizes of micro-fabricated cavities on a new silicon test section immersed in FC-72 at the saturation temperature for different pressures with an imposed heat flux applied at the back of the plate. Particular focus is on variations of wall temperature around nucleation sites. (author)

  10. A feasibility and optimization study to determine cooling time and burnup of advanced test reactor fuels using a nondestructive technique

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Jorge [Univ. of Utah, Salt Lake City, UT (United States)

    2013-12-01

    The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method

  11. Initial river test of a monostatic RiverSonde streamflow measurement system

    Science.gov (United States)

    Teague, C.C.; Barrick, D.E.; Lilleboe, P.M.; Cheng, R.T.; ,

    2003-01-01

    A field experiment was conducted on May 7-8, 2002 using a CODAR RiverSonde UHF radar system at Vernalis, California on the San Joaquin River. The monostatic radar configuration on one bank of the river, with the antennas looking both upriver and downriver, provided very high-quality data. Estimates of both along-river and cross-river surface current were generated using several models, including one based on normal-mode analysis. Along-river surface velocities ranged from about 0.6 m/s at the river banks to about 1.0 m/s near the middle of the river. Average cross-river surface velocities were 0.02 m/s or less.

  12. Criticality Safety Evaluation for the Advanced Test Reactor U-Mo Demonstration Elements

    Energy Technology Data Exchange (ETDEWEB)

    Leland M. Montierth

    2010-12-01

    The Reduced Enrichment Research Test Reactors (RERTR) fuel development program is developing a high uranium density fuel based on a (LEU) uranium-molybdenum alloy. Testing of prototypic RERTR fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. Two RERTR-Full Size Demonstration fuel elements based on the ATR-Reduced YA elements (all but one plate fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). The two fuel elements will be irradiated in alternating cycles such that only one element is loaded in the reactor at a time. Existing criticality analyses have analyzed Standard (HEU) ATR elements (all plates fueled) from which controls have been derived. This criticality safety evaluation (CSE) documents analysis that determines the reactivity of the Demonstration fuel elements relative to HEU ATR elements and shows that the Demonstration elements are bound by the Standard HEU ATR elements and existing HEU ATR element controls are applicable to the Demonstration elements.

  13. The risk of shortage of radioelements at medical use must not lead to overlook the reactors safety that produce them; Le risque de penurie de radioelements a usage medical ne doit pas conduire a faire l'impasse sur la surete des reacteurs qui les produisent

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2009-07-01

    As the reactors supplying the world production of radioelements for medical use have over 40 years of operation, the nuclear safety authority alerts the stake holders on the necessity to prevent the conflicts between public health and nuclear safety in the production of these elements; Asn estimates that the solution is not to extend the lifetime of the reactors but goes for a new international concerted approach. The most of the present production comes from five old reactors: N.R.U. at Chalk river (Canada, 40%), H.F.R. at Petten (Netherlands, 30%), Safari at Pelindaba (South Africa, 10%) B.R.2 at Mol (Belgium, 9%) and Osiris at Saclay (France, 5%). In this context, Asn organised in january 2009 a seminar on the safety-availability of facilities of radio-isotopes production with safety authorities of the concerned countries. Nea organised a seminar on the radiopharmaceuticals supply at the end of january 2009. (N.C.)

  14. Vortex Diode Analysis and Testing for Fluoride Salt-Cooled High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoder Jr, Graydon L [ORNL; Elkassabgi, Yousri M. [Texas A& M University, Kingsville; De Leon, Gerardo I. [Texas A& M University, Kingsville; Fetterly, Caitlin N. [Texas A& M University, Kingsville; Ramos, Jorge A. [Texas A& M University, Kingsville; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK)

    2012-02-01

    Fluidic diodes are presently being considered for use in several fluoride salt-cooled high-temperature reactor designs. A fluidic diode is a passive device that acts as a leaky check valve. These devices are installed in emergency heat removal systems that are designed to passively remove reactor decay heat using natural circulation. The direct reactor auxiliary cooling system (DRACS) uses DRACS salt-to-salt heat exchangers (DHXs) that operate in a path parallel to the core flow. Because of this geometry, under normal operating conditions some flow bypasses the core and flows through the DHX. A flow diode, operating in reverse direction, is-used to minimize this flow when the primary coolant pumps are in operation, while allowing forward flow through the DHX under natural circulation conditions. The DRACSs reject the core decay heat to the environment under loss-of-flow accident conditions and as such are a reactor safety feature. Fluidic diodes have not previously been used in an operating reactor system, and therefore their characteristics must be quantified to ensure successful operation. This report parametrically examines multiple design parameters of a vortex-type fluidic diode to determine the size of diode needed to reject a particular amount of decay heat. Additional calculations were performed to size a scaled diode that could be tested in the Oak Ridge National Laboratory Liquid Salt Flow Loop. These parametric studies have shown that a 152.4 mm diode could be used as a test article in that facility. A design for this diode is developed, and changes to the loop that will be necessary to test the diode are discussed. Initial testing of a scaled flow diode has been carried out in a water loop. The 150 mm diode design discussed above was modified to improve performance, and the final design tested was a 171.45 mm diameter vortex diode. The results of this testing indicate that diodicities of about 20 can be obtained for diodes of this size. Experimental

  15. A preliminary neutronic evaluation of high temperature engineering test reactor using the SCALE6 code

    Science.gov (United States)

    Tanure, L. P. A. R.; Sousa, R. V.; Costa, D. F.; Cardoso, F.; Veloso, M. A. F.; Pereira, C.

    2014-02-01

    Neutronic parameters of some fourth generation nuclear reactors have been investigated at the Departamento de Engenharia Nuclear/UFMG. Previous studies show the possibility to increase the transmutation capabilities of these fourth generation systems to achieve significant reduction concerning transuranic elements in spent fuel. To validate the studies, a benchmark on core physics analysis, related to initial testing of the High Temperature Engineering Test Reactor and provided by International Atomic Energy Agency (IAEA) was simulated using the Standardized Computer Analysis for Licensing Evaluation (SCALE). The CSAS6/KENO-VI control sequence and the 44-group ENDF/B-V 0 cross-section neutron library were used to evaluate the keff (effective multiplication factor) and the result presents good agreement with experimental value.

  16. Feasibility of conducting a dynamic helium charging experiment for vanadium alloys in the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.; Strain, R.V.; Smith, D.L. [Argonne National Lab., IL (United States); Matsui, H. [Tohoku Univ. (Japan)

    1996-10-01

    The feasibility of conducting a dynamic helium charging experiment (DHCE) for vanadium alloys in the water-cooled Advanced Test Reactor (ATR) is being investigated as part of the U.S./Monbusho collaboration. Preliminary findings suggest that such an experiment is feasible, with certain constraints. Creating a suitable irradiation position in the ATR, designing an effective thermal neutron filter, incorporating thermocouples for limited specimen temperature monitoring, and handling of tritium during various phases of the assembly and reactor operation all appear to be feasible. An issue that would require special attention, however, is tritium permeation loss through the capsule wall at the higher design temperatures (>{approx}600{degrees}C). If permeation is excessive, the reduced amount of tritium entering the test specimens would limit the helium generation rates in them. At the lower design temperatures (<{approx}425{degrees}C), sodium, instead of lithium, may have to be used as the bond material to overcome the tritium solubility limitation.

  17. Field test and evaluation of the passive neutron coincidence collar for prototype fast reactor fuel subassemblies

    Energy Technology Data Exchange (ETDEWEB)

    Menlove, H.O.; Keddar, A.

    1982-08-01

    The passive neutron Coincidence Collar, which was developed for the verification of plutonium content in fast reactor fuel subassemblies, has been field tested using Prototype Fast Reactor fuel. For passive applications, the system measures the /sup 240/Pu-effective mass from the spontaneous fission rate, and in addition, a self-interrogation technique is used to determine the fissile content in the subassembly. Both the passive and active modes were evaluated at the Windscale Works in the United Kingdom. The results of the tests gave a standard deviation 0.75% for the passive count and 3 to 7% for the active measurement for a 1000-s counting time. The unit will be used in the future for the verification of plutonium in fresh fuel assemblies.

  18. Steam-air mixture condensation in a subcooled water pool

    Science.gov (United States)

    Norman, Timothy Linhurst

    2007-12-01

    In any conceptual reactor design under postulated accidental conditions, one parameter that is considered as being highly ranked in determining the thermal-hydraulic conditions of the reactor safety components is the system pressure. To obtain a satisfactory prediction of steam partial pressure, within reasonable uncertainty in the gas space of a confined SP (suppression pool) bounded to the steam source of the break flow, one must establish a means by which local phenomena associated with steam direct contact condensation in the subcooled water pool can be fully addressed to predict the global component thermal response. For this purpose a scaled down, reduced pressure, suppression pool was designed and built to study condensation and mixing phenomena. The scaled test facility represented an idealized trapezoidal cross section, 1/10 sector of the SP with scaled height ratio of 1/4.5 and volume ratio of 1/400. The design and test conditions were based on a hierarchical scaling principle that preserves the transfer of mass, momentum, energy and condensation phenomena. Distributed thermocouples within the pool provided a means to quantify the pool thermal response. The test loop was not only instrumented with thermocouples for monitoring pool stratification but also with high speed photography for flow visualization from which to build a comprehensive database to identify the regions of the pool that were thermally stratified or mixed. Data were obtained for different pool initial subcooling and steam/air mixture flow rates. Dimensionless boundary maps were plotted from several experimental runs of pure steam injection to determine conditions when the pool transits from being homogeneously mixed to being thermally stratified. Steam-air mixture injection cases for single horizontal venting indicated that above a pool temperature of 40°C with airmass flow rates below 0.1 g/s the pool can attain thermal stratification. Models of a single phase liquid

  19. Highly Perturbed Operational Test Configurations at the WSMR Fast Burst Reactor

    Directory of Open Access Journals (Sweden)

    Flanders T. Michael

    2016-01-01

    Full Text Available The White Sands Missile Range (WSMR MoLLY-G reactor has a long history of producing a well characterized environment for testing electronic systems/devices in fission environments. As an unmoderated, unreflected, bare critical assembly, it provides a slightly degraded fission spectrum with a 1/E tail. For radiation hardness testing of electronics, the neutron fluence is usually reported as the 1-MeV Equivalent Neutron Fluence for Silicon. In this paper, we examine additional neutron environments and characterizations ranging from low intensity neutron fields to more extreme modifications of our normal test environment.

  20. Dynamic parameters test of Haiyang Nuclear Power Engineering in reactor areas, China

    Science.gov (United States)

    Zhou, N.; Zhao, S.; Sun, L.

    2012-12-01

    Haiyang Nuclear Power Project is located in Haiyang city, China. It consists of 6×1000MW AP1000 Nuclear Power generator sets. The dynamic parameters of the rockmass are essential for the design of the nuclear power plant. No.1 and No.2 reactor area are taken as research target in this paper. Sonic logging, single hole and cross-hole wave velocity are carried out respectively on the site. There are four types of rock lithology within the measured depth. They are siltstone, fine sandstone, shale and allgovite. The total depth of sonic logging is 409.8m and 2049 test points. The sound wave velocity of the rocks are respectively 5521 m/s, 5576m/s, 5318 m/s and 5576 m/s. Accroding to the statistic data, among medium weathered fine sandstone, fairly broken is majority, broken and relatively integrity are second, part of integrity. Medium weathered siltstone, relatively integrity is mojority, fairly broken is second. Medium weathered shale, fairly broken is majority, broken and relatively integrity for the next and part of integrity. Slight weathered fine sandstone, siltstone, shale and allgovite, integrity is the mojority, relatively integrity for the next, part of fairly broken.The single hole wave velocity tests are set in two boreholesin No.1 reactor area and No.2 reactor area respectively. The test depths of two holes are 2-24m, and the others are 2-40m. The wave velocity data are calculated at different depth in each holes and dynamic parameters. According to the test statistic data, the wave velocity and the dynamic parameter values of rockmass are distinctly influenced by the weathering degree. The test results are list in table 1. 3 groups of cross hole wave velocity tests are set for No.1 and 2 reactor area, No.1 reactor area: B16, B16-1, B20(Direction:175°, depth: 100m); B10, B10-1, B11(269°, 40m); B21, B21-1, B17(154°, 40m); with HB16, HB10, HB21 as trigger holes; No.2 reactor area: B47, B47-1, HB51(176°, 100m); B40, B40-1, B41(272°, 40m); B42, B42-1, B