WorldWideScience

Sample records for total reactor volume

  1. Evaluation of Productivity of Zymotis Solid-State Bioreactor Based on Total Reactor Volume

    Directory of Open Access Journals (Sweden)

    Oscar F. von Meien

    2002-01-01

    Full Text Available In this work a method of analyzing the performance of solid-state fermentation bioreactors is described. The method is used to investigate the optimal value for the spacing between the cooling plates of the Zymotis bioreactor, using simulated fermentation data supplied by a mathematical model. The Zymotis bioreactor has good potential for those solid-state fermentation processes in which the substrate bed must remain static. The current work addresses two design parameters introduced by the presence of the internal heat transfer plates: the width of the heat transfer plate, which is governed by the amount of heat to be removed and the pressure drop of the cooling water, and the spacing between these heat transfer plates. In order to analyze the performance of the bioreactor a productivity term is introduced that takes into account the volume occupied within the bioreactor by the heat transfer plates. As part of this analysis, it is shown that, for logistic growth kinetics, the time at which the biomass reaches 90 % of its maximum possible value is a good estimate of the optimum harvesting time for maximizing productivity. Application of the productivity analysis to the simulated fermentation results suggests that, with typical fast growing fungi ( = 0.324 h–1, the optimal spacing between heat transfer plates is of the order of 6 cm. The general applicability of this approach to evaluate the productivity of solid-state bioreactors is demonstrated.

  2. Total volume versus bouts

    DEFF Research Database (Denmark)

    Chinapaw, Mai; Klakk, Heidi; Møller, Niels Christian

    2018-01-01

    BACKGROUND/OBJECTIVES: Examine the prospective relationship of total volume versus bouts of sedentary behaviour (SB) and moderate-to-vigorous physical activity (MVPA) with cardiometabolic risk in children. In addition, the moderating effects of weight status and MVPA were explored. SUBJECTS....../METHODS: Longitudinal study including 454 primary school children (mean age 10.3 years). Total volume and bouts (i.e. ≥10 min consecutive minutes) of MVPA and SB were assessed by accelerometry in Nov 2009/Jan 2010 (T1) and Aug/Oct 2010 (T2). Triglycerides, total cholesterol/HDL cholesterol ratio (TC:HDLC ratio......, with or without mutual adjustments between MVPA and SB. The moderating effects of weight status and MVPA (for SB only) were examined by adding interaction terms. RESULTS: Children engaged daily in about 60 min of total MVPA and 0-15 min/week in MVPA bouts. Mean total sedentary time was around 7 h/day with over 3...

  3. Neutron behavior, reactor control, and reactor heat transfer. Volume four

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant)

  4. Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two

    International Nuclear Information System (INIS)

    Glasstone, S.; Sesonske, A.

    1994-01-01

    This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future

  5. Nuclear reactor engineering: Reactor design basics. Fourth edition, Volume One

    International Nuclear Information System (INIS)

    Glasstone, S.; Sesonske, A.

    1994-01-01

    This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in design and operation of nuclear power plants. Extensively updated, the fourth edition includes new material on reactor safety and risk analysis, regulation, fuel management, waste management, and operational aspects of nuclear power. This volume contains the following: energy from nuclear fission; nuclear reactions and radiations; neutron transport; nuclear design basics; nuclear reactor kinetics and control; radiation protection and shielding; and reactor materials

  6. The Calculation Of Total Radioactivity Of Kartini Reactor Fuel Element

    International Nuclear Information System (INIS)

    Budisantoso, Edi Trijono; Sardjono, Y.

    1996-01-01

    The total radioactivity of Kartini reactor fuel element has been calculated by using ORIGEN2. In this case, the total radioactivity is the sum of alpha, beta, and gamma radioactivity from activation products nuclides, actinide nuclides and fission products nuclides in the fuel element. The calculation was based on irradiation history of fuel in the reactor core. The fuel element no 3203 has location history at D, E, and F core zone. The result is expressed in graphics form of total radioactivity and photon radiations as function of irradiation time and decay time. It can be concluded that the Kartini reactor fuel element in zone D, E, and F has total radioactivity range from 10 Curie to 3000 Curie. This range is for radioactivity after decaying for 84 days and that after reactor shut down. This radioactivity is happened in the fuel element for every reactor operation and decayed until the fuel burn up reach 39.31 MWh. The total radioactivity emitted photon at the power of 0.02 Watt until 10 Watt

  7. Total tree, merchantable stem and branch volume models for ...

    African Journals Online (AJOL)

    Total tree, merchantable stem and branch volume models for miombo woodlands of Malawi. Daud J Kachamba, Tron Eid. Abstract. The objective of this study was to develop general (multispecies) models for prediction of total tree, merchantable stem and branch volume including options with diameter at breast height (dbh) ...

  8. Advances in Reactor Physics, Mathematics and Computation. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, Volume 2, are divided into 7 sessions bearing on: - session 7: Deterministic transport methods 1 (7 conferences), - session 8: Interpretation and analysis of reactor instrumentation (6 conferences), - session 9: High speed computing applied to reactor operations (5 conferences), - session 10: Diffusion theory and kinetics (7 conferences), - session 11: Fast reactor design, validation and operating experience (8 conferences), - session 12: Deterministic transport methods 2 (7 conferences), - session 13: Application of expert systems to physical aspects of reactor design and operation.

  9. Advances in Reactor Physics, Mathematics and Computation. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume one, are divided into 6 sessions bearing on: - session 1: Advances in computational methods including utilization of parallel processing and vectorization (7 conferences) - session 2: Fast, epithermal, reactor physics, calculation, versus measurements (9 conferences) - session 3: New fast and thermal reactor designs (9 conferences) - session 4: Thermal radiation and charged particles transport (7 conferences) - session 5: Super computers (7 conferences) - session 6: Thermal reactor design, validation and operating experience (8 conferences).

  10. Sodium fast reactor safety and licensing research plan. Volume II.

    Energy Technology Data Exchange (ETDEWEB)

    Ludewig, H. (Brokhaven National Laboratory, Upton, NY); Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A. (Argonne National Laboratory, Argonne, IL); Phillips, J.; Zeyen, R. (Institute for Energy Petten, Saint-Paul-lez-Durance, France); Clement, B. (IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France); Garner, Frank (Radiation Effects Consulting, Richland, WA); Walters, Leon (Advanced Reactor Concepts, Los Alamos, NM); Wright, Steve; Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Suo-Anttila, Ahti Jorma; Denning, Richard (Ohio State University, Columbus, OH); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki, Japan); Ohno, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Miyhara, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Yacout, Abdellatif (Argonne National Laboratory, Argonne, IL); Farmer, M. (Argonne National Laboratory, Argonne, IL); Wade, D. (Argonne National Laboratory, Argonne, IL); Grandy, C. (Argonne National Laboratory, Argonne, IL); Schmidt, R.; Cahalen, J. (Argonne National Laboratory, Argonne, IL); Olivier, Tara Jean; Budnitz, R. (Lawrence Berkeley National Laboratory, Berkeley, CA); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache, Cea, France); Natesan, Ken (Argonne National Laboratory, Argonne, IL); Carbajo, Juan J. (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin-Madison, Madison, WI); Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN); Bari, R. (Brokhaven National Laboratory, Upton, NY); Porter D. (Idaho National Laboratory, Idaho Falls, ID); Lambert, J. (Argonne National Laboratory, Argonne, IL); Hayes, S. (Idaho National Laboratory, Idaho Falls, ID); Sackett, J. (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.

    2012-05-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  11. Sodium fast reactor safety and licensing research plan - Volume II

    International Nuclear Information System (INIS)

    Ludewig, H.; Powers, D.A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A.; Phillips, J.; Zeyen, R.; Clement, B.; Garner, Frank; Walters, Leon; Wright, Steve; Ott, Larry J.; Suo-Anttila, Ahti Jorma; Denning, Richard; Ohshima, Hiroyuki; Ohno, S.; Miyhara, S.; Yacout, Abdellatif; Farmer, M.; Wade, D.; Grandy, C.; Schmidt, R.; Cahalen, J.; Olivier, Tara Jean; Budnitz, R.; Tobita, Yoshiharu; Serre, Frederic; Natesan, Ken; Carbajo, Juan J.; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Thomas, Justin; Wei, Tom; Sofu, Tanju; Flanagan, George F.; Bari, R.; Porter D.

    2012-01-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  12. Nuclear powerplant standardization: light water reactors. Volume 2. Appendixes

    International Nuclear Information System (INIS)

    1981-06-01

    This volume contains working papers written for OTA to assist in preparation of the report, NUCLEAR POWERPLANT STANDARDIZATION: LIGHT WATER REACTORS. Included in the appendixes are the following: the current state of standardization, an application of the principles of the Naval Reactors Program to commercial reactors; the NRC and standardization, impacts of nuclear powerplant standardization on public health and safety, descriptions of current control room designs and Duke Power's letter, Admiral Rickover's testimony, a history of standardization in the NRC, and details on the impact of standardization on public health and safety

  13. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    International Nuclear Information System (INIS)

    Greene, S.R.; Spellman, D.J.; Bevard, B.B.

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative

  14. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    Energy Technology Data Exchange (ETDEWEB)

    Greene, S.R.; Spellman, D.J.; Bevard, B.B. [and others

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative.

  15. Advances in Reactor physics, mathematics and computation. Volume 3

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume 3, are divided into sessions bearing on: - poster sessions on benchmark and codes: 35 conferences - review of status of assembly spectrum codes: 9 conferences - Numerical methods in fluid mechanics and thermal hydraulics: 16 conferences - stochastic transport and methods: 7 conferences.

  16. Atlas of total body radionuclide imaging. Volume I and II

    International Nuclear Information System (INIS)

    Fordham, E.W.; Ali, A.; Turner, D.A.; Charters, J.

    1982-01-01

    This two-volume work on total body imaging may well be regarded by future historians of nuclear medicine as representing the high points in the art of total body imaging in clinical nuclear medicine. With regard to information content and volume, it is the largest collection of well-interpreted, beautifully reproduced, total body images available to date. The primary goal of this atlas is to demonstrate patterns of abnormality in both typical and less typical variations. This goal is accomplished with many well-described examples of technical artifacts, of normal variants, of common and of rare diseases, and of pitfalls in interpretations. Volume I is entirely dedicated to skeletal imaging with Tc-99m labeled phosphates or phosphonates. The volume is divided into 22 chapters, which include chapters on methodology and instrumentation, chapters on the important bone diseases and other topics such as a treatise on false-negative and false-positive scans, and soft tissue and urinary tract abnormalities recognizable on bone scintigrams

  17. Research program plan: reactor vessels. Volume 1

    International Nuclear Information System (INIS)

    Vagins, M.; Taboada, A.

    1985-07-01

    The ability of the licensing staff of the NRC to make decisions concerning the present and continuing safety of nuclear reactor pressure vessels under both normal and abnormal operating conditions is dependent upon the existence of verified analysis methods and a solid background of applicable experimental data. It is the role of this program to provide both the analytical methods and the experimental data needed. Specifically, this program develops fracture mechanics analysis methods and design criteria for predicting the stress levels and flaw sizes required for crack initiation, propagation, and arrest in LWR pressure vessels under all known and postulated operations conditions. To do this, not only must the methods be developed but they must be experimentally validated. Further, the materials data necessary for input to these analytical methods must be developed. Thus, in addition to methods development and large scale experimental verification this program also develops data to show that slow-load fracture toughness, rapid-load fracture toughness, and crack arrest toughness obtained from small laboratory specimens are truly representative of the toughness characteristics of the material behavior in pressure vessels in both the unirradiated and the irradiated conditions

  18. FMDP reactor alternative summary report: Volume 4, Evolutionary LWR alternative

    International Nuclear Information System (INIS)

    1996-09-01

    Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] have become surplus to national defense needs both in the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES ampersand H) consequences if surplus fissile materials are not properly managed. The purpose of this report is to provide schedule, cost, and technical information that will be used to support the Record of Process (ROD). Following the screening process, DOE/MD via its national laboratories initiated a more detailed analysis activity to further evaluate each of the ten plutonium disposition alternatives that survived the screening process. Three ''Alternative Teams,'' chartered by DOE and comprised of technical experts from across the DOE national laboratory complex, conducted these analyses. One team was chartered for each of the major disposition classes (borehole, immobilization, and reactors). During the last year and a half, the Fissile Materials Disposition Program (FMDP) Reactor Alternative Team (RxAT) has conducted extensive analyses of the cost, schedule, technical maturity, S ampersand S, and other characteristics of reactor-based plutonium disposition. The results of the RxAT's analyses of the existing LWR, CANDU, and partially complete LWR alternatives are documented in Volumes 1-3 of this report. This document (Volume 4) summarizes the results of these analyses for the ELWR-based plutonium disposition option

  19. FMDP reactor alternative summary report: Volume 4, Evolutionary LWR alternative

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] have become surplus to national defense needs both in the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES&H) consequences if surplus fissile materials are not properly managed. The purpose of this report is to provide schedule, cost, and technical information that will be used to support the Record of Process (ROD). Following the screening process, DOE/MD via its national laboratories initiated a more detailed analysis activity to further evaluate each of the ten plutonium disposition alternatives that survived the screening process. Three ``Alternative Teams,`` chartered by DOE and comprised of technical experts from across the DOE national laboratory complex, conducted these analyses. One team was chartered for each of the major disposition classes (borehole, immobilization, and reactors). During the last year and a half, the Fissile Materials Disposition Program (FMDP) Reactor Alternative Team (RxAT) has conducted extensive analyses of the cost, schedule, technical maturity, S&S, and other characteristics of reactor-based plutonium disposition. The results of the RxAT`s analyses of the existing LWR, CANDU, and partially complete LWR alternatives are documented in Volumes 1-3 of this report. This document (Volume 4) summarizes the results of these analyses for the ELWR-based plutonium disposition option.

  20. Proceedings of the international topical meeting on advanced reactors safety: Volume 2

    International Nuclear Information System (INIS)

    1997-01-01

    In this volume, 89 papers are grouped under the following headings: advances in research/test reactor safety; advanced reactor accident management and emergency actions; advanced reactors instrumentation/controls/human factors; probabilistic risk/safety and reliability assessments; steam explosion research and issues; advanced reactor severe accident issues and research (analysis and assessments); advanced reactor thermal hydraulics; accelerator-driven source safety; liquid-metal reactor safety; structural assessments and issues; late papers

  1. Clinical associations of total kidney volume: the Framingham Heart Study.

    Science.gov (United States)

    Roseman, Daniel A; Hwang, Shih-Jen; Oyama-Manabe, Noriko; Chuang, Michael L; O'Donnell, Christopher J; Manning, Warren J; Fox, Caroline S

    2017-08-01

    Total kidney volume (TKV) is an imaging biomarker that may have diagnostic and prognostic utility. The relationships between kidney volume, renal function and cardiovascular disease (CVD) have not been characterized in a large community-dwelling population. This information is needed to advance the clinical application of TKV. We measured TKV in 1852 Framingham Heart Study participants (mean age 64.1 ± 9.2 years, 53% women) using magnetic resonance imaging. A healthy sample was used to define reference values. The associations between TKV, renal function and CVD risk factors were determined using multivariable logistic regression analysis. Overall, mean TKV was 278 ± 54 cm3 for women and 365 ± 66 cm3 for men. Risk factors for high TKV (>90% healthy referent size) were body surface area (BSA), diabetes, smoking and albuminuria, while age, female and estimated glomerular filtration rate (eGFR) kidney damage including albuminuria and eGFR <60 mL/min/1.73 m2, while high TKV is associated with diabetes and decreased odds of eGFR <60 mL/min/1.73 m2. Prospective studies are needed to characterize the natural progression and clinical consequences of TKV. Published by Oxford University Press on behalf of ERA-EDTA 2016. This work is written by US Government employees and is in the public domain in the US.

  2. Thirteenth water reactor safety research information meeting: proceedings Volume 1

    International Nuclear Information System (INIS)

    Weiss, A.J.

    1986-02-01

    This six-volume report contains 151 papers out of the 178 that were presented at the Thirteenth Water Reactor Safety Research Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 22-25, 1985. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included thirty-one different papers presented by researchers from Japan, Canada and eight European countries. The title of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume presents information on: risk analysis PRA application; severe accident sequence analysis; risk analysis/dependent failure analysis; and industry safety research

  3. Decommissioning of the AVR reactor, concept for the total dismantling

    International Nuclear Information System (INIS)

    Marnet, C.; Wimmers, M.; Birkhold, U.

    1998-01-01

    After more than 21 years of operation, the 15 MWe AVR experimental nuclear power plant with pebble bed high temperature gas-cooled reactor was shout down in 1988. Safestore decommissioning began in 1994. In order to completely dismantle the plant, a concept for Continued dismantling was developed according to which the plant could be dismantled in a step-wise procedure. After each step, there is the possibility to transform the plant into a new state of safe enclosure. The continued dismantling comprises three further steps following Safestore decommissioning: 1. Dismantling the reactor vessels with internals; 2. Dismantling the containment and the auxiliary units; 3. Gauging the buildings to radiation limit, release from the validity range of the AtG (Nuclear Act), and demolition of the buildings. For these steps, various technical procedures and concepts were developed, resulting in a reference concept in which the containment will essentially remain intact (in-situ concept). Over the top of the outer reactor vessel a disassembling area for remotely controlled tools will be erected that tightens on that vessel and can move down on the vessel according to the dismantling progress. (author)

  4. Removal of Total Coliforms, Thermotolerant Coliforms, and Helminth Eggs in Swine Production Wastewater Treated in Anaerobic and Aerobic Reactors

    Science.gov (United States)

    Zacarias Sylvestre, Silvia Helena; Lux Hoppe, Estevam Guilherme; de Oliveira, Roberto Alves

    2014-01-01

    The present work evaluated the performance of two treatment systems in reducing indicators of biological contamination in swine production wastewater. System I consisted of two upflow anaerobic sludge blanket (UASB) reactors, with 510 and 209 L in volume, being serially arranged. System II consisted of a UASB reactor, anaerobic filter, trickling filter, and decanter, being also organized in series, with volumes of 300, 190, 250, and 150 L, respectively. Hydraulic retention times (HRT) applied in the first UASB reactors were 40, 30, 20, and 11 h in systems I and II. The average removal efficiencies of total and thermotolerant coliforms in system I were 92.92% to 99.50% and 94.29% to 99.56%, respectively, and increased in system II to 99.45% to 99.91% and 99.52% to 99.93%, respectively. Average removal rates of helminth eggs in system I were 96.44% to 99.11%, reaching 100% as in system II. In reactor sludge, the counts of total and thermotolerant coliforms ranged between 105 and 109 MPN (100 mL)−1, while helminth eggs ranged from 0.86 to 9.27 eggs g−1 TS. PMID:24812560

  5. Future view of total energy system and reactor engineering and reactor physics

    International Nuclear Information System (INIS)

    Ozawa, T.

    1974-01-01

    This paper outlines the present status of fission reactors and fusion reactors. The conversion ratio of light water reactors is 0.5, and the efficiency is 32% because of relatively low temperature. Both pressurized water reactors and boiling water reactors are technically well developed, their performances are well known, and the fuel cycle is well developed, so that both reactors have monopolized power reactor market. But the reprocessing of spent fuel and the treatment of their hazards are inevitable, and the construction and enlargement of reprocessing facilities are indispensable. In LMFBR's tight sealing is easy because they are non-pressurized, and the efficiency is 41%. But liquid sodium is strongly activated and recirculated, so that chemical obstruction due to the breakage of recirculating pumps, pipings, and heat exchangers may occur, and the hazard of plutonium is large. Regarding controlled thermo-nuclear fusion reactors, because Lawson criterion must be satisfied, two methods of plasma confinement are now experimented. One is the plasma confinement by strong magnetic field of 50 KG to 100 KG, and the other is the confinement by the implosion method with high-power laser beam. The latter has much more uncertainties than the former, but recently both methods have made much progress. (Tai, I)

  6. Fast reactor safety: proceedings of the international topical meeting. Volume 1

    International Nuclear Information System (INIS)

    1985-07-01

    The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 1 include: impact of safety and licensing considerations on fast reactor design; safety aspects of innovative designs; intra-subassembly behavior; operational safety; design accommodation of seismic and other external events; natural circulation; safety design concepts; safety implications derived from operational plant data; decay heat removal; and assessment of HCDA consequences

  7. Fast reactor safety: proceedings of the international topical meeting. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1985-07-01

    The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 1 include: impact of safety and licensing considerations on fast reactor design; safety aspects of innovative designs; intra-subassembly behavior; operational safety; design accommodation of seismic and other external events; natural circulation; safety design concepts; safety implications derived from operational plant data; decay heat removal; and assessment of HCDA consequences.

  8. Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1992-04-01

    This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

  9. Criticality accident in uranium fuel processing plant. The estimation of the total number of fissions with related reactor physics parameters

    International Nuclear Information System (INIS)

    Nishina, Kojiro; Oyamatsu, Kazuhiro; Kondo, Shunsuke; Sekimoto, Hiroshi; Ishitani, Kazuki; Yamane, Yoshihiro; Miyoshi, Yoshinori

    2000-01-01

    This accident occurred when workers were pouring a uranium solution into a precipitation tank with handy operation against the established procedure and both the cylindrical diameter and the total mass exceeded the limited values. As a result, nuclear fission chain reactor in the solution reached not only a 'criticality' state continuing it independently but also an instantly forming criticality state exceed the criticality and increasing further nuclear fission number. The place occurring the accident at this time was not reactor but a place having not to form 'criticality' called by a processing process of uranium fuel. In such place, as because of relating to mechanism of chain reaction, it is required naturally for knowledge on the reactor physics, it is also necessary to understand chemical reaction in chemical process, and functions of tanks, valves and pumps mounted at the processes. For this purpose, some information on uranium concentration ratio, atomic density of nuclides largely affecting to chain reaction such as uranium, hydrogen, and so forth in the solution, shape, inner structure and size of container for the solution, and its temperature and total volume, were necessary for determining criticality volume of the accident uranium solution by using nuclear physics procedures. Here were described on estimation of energy emission in the JCO accident, estimation from analytical results on neutron and solution, calculation of various nuclear physics property estimation on the JCO precipitation tank at JAERI. (G.K.)

  10. Report on the Survey of the Design Review of New Reactor Applications. Volume 3: Reactor

    International Nuclear Information System (INIS)

    Downey, Steven; Monninger, John; Nevalainen, Janne; Lorin, Aurelie; ); Webster, Philip; Joyer, Philippe; Kawamura, Tomonori; Lankin, Mikhail; Kubanyi, Jozef; Haluska, Ladislav; Persic, Andreja; Reierson, Craig; Kang, Kyungmin; Kim, Walter

    2016-01-01

    At the tenth meeting of the CNRA Working Group on the Regulation of New Reactors (WGRNR) in March 2013, the Working Group agreed to present the responses to the Second Phase, or Design Phase, of the Licensing Process Survey as a multi-volume text. As such, each report will focus on one of the eleven general technical categories covered in the survey. The general technical categories were selected to conform to the topics covered in the International Atomic Energy Agency (IAEA) Safety Guide GS-G-4.1. This document, which is the third report on the results of the Design Phase Survey, focuses on the Reactor. The Reactor category includes the following technical topics: fuel system design, reactor internals and core support, nuclear design and core nuclear performance, thermal and hydraulic design, reactor materials, and functional design of reactivity control system. For each technical topic, the member countries described the information provided by the applicant, the scope and level of detail of the technical review, the technical basis for granting regulatory authorisation, the skill sets required and the level of effort needed to perform the review. Based on a comparison of the information provided by the member countries in response to the survey, the following observations were made: - Although the description of the information provided by the applicant differs in scope and level of detail among the member countries that provided responses, there are similarities in the information that is required. - All of the technical topics covered in the survey are reviewed in some manner by all of the regulatory authorities that provided responses. - Design review strategies most commonly used to confirm that the regulatory requirements have been met include document review and independent verification of calculations, computer codes, or models used to describe the design and performance of the core and the fuel. - It is common to consider operating experience and

  11. Operating reactors licensing actions summary. Volume 5, No. 2

    International Nuclear Information System (INIS)

    1985-04-01

    The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management. This summary report is published primarily for internal NRC use in managing the Operating Reactors Licensing Actions Program

  12. Volume reduction of reactor wastes by spray drying

    International Nuclear Information System (INIS)

    Gay, R.L.; Grantham, L.F.; McKenzie, D.E.

    1983-01-01

    Three simulated low-level reactor wastes were dried using a spray dryer-baghouse system. The three aqueous feedstocks were sodium sulfate waste characteristic of a BWR, boric acid waste characteristic of a PWR, and a waste mixture of ion exchange resins and filter aid. These slurries were spiked with nonradioactive iron, cobalt, and manganese (representing corrosion products) and nonradioactive cesium and iodine (representing fission products). The throughput for the 2.1-m-diameter spray dryer and baghouse system was 160-180 kg/h, which is comparable to the requirements for a full-scale commercial installation. A free-flowing, dry product was produced in all of the tests. The volume reduction factor ranged from 2.5 to 5.8; the baghouse decontamination factor was typically in the range of 10 3 to 10 4 . Using an overall system decontamination factor of 10 6 , the activity of the off-gas was calculated to be one to two orders of magnitude less than the nuclide release limit of the major active species, Cs-137

  13. Operating reactors licensing actions summary. Volume 5, No. 7

    International Nuclear Information System (INIS)

    1985-09-01

    The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

  14. Total Absorption Spectroscopy of Fission Fragments Relevant for Reactor Antineutrino Spectra and Decay Heat Calculations

    Directory of Open Access Journals (Sweden)

    Porta A.

    2016-01-01

    Full Text Available Beta decay of fission products is at the origin of decay heat and antineutrino emission in nuclear reactors. Decay heat represents about 7% of the reactor power during operation and strongly impacts reactor safety. Reactor antineutrino detection is used in several fundamental neutrino physics experiments and it can also be used for reactor monitoring and non-proliferation purposes. 92,93Rb are two fission products of importance in reactor antineutrino spectra and decay heat, but their β-decay properties are not well known. New measurements of 92,93Rb β-decay properties have been performed at the IGISOL facility (Jyväskylä, Finland using Total Absorption Spectroscopy (TAS. TAS is complementary to techniques based on Germanium detectors. It implies the use of a calorimeter to measure the total gamma intensity de-exciting each level in the daughter nucleus providing a direct measurement of the beta feeding. In these proceedings we present preliminary results for 93Rb, our measured beta feedings for 92Rb and we show the impact of these results on reactor antineutrino spectra and decay heat calculations.

  15. Total decay heat estimates in a proto-type fast reactor

    International Nuclear Information System (INIS)

    Sridharan, M.S.

    2003-01-01

    Full text: In this paper, total decay heat values generated in a proto-type fast reactor are estimated. These values are compared with those of certain fast reactors. Simple analytical fits are also obtained for these values which can serve as a handy and convenient tool in engineering design studies. These decay heat values taken as their ratio to the nominal operating power are, in general, applicable to any typical plutonium based fast reactor and are useful inputs to the design of decay-heat removal systems

  16. Operating reactors licensing actions summary. Volume 5, Number 1

    International Nuclear Information System (INIS)

    1985-03-01

    This document is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program

  17. Operating reactors licensing actions summary. Volume 5, No. 6

    International Nuclear Information System (INIS)

    1985-08-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management. This summary report is published for internal NRC use in managing the Operating Reactors Licensing Actions Program. Its content will change based on NRC management informational requirements

  18. Stereological quantification of tumor volume, mean nuclear volume and total number of melanoma cells correlated with morbidity and mortality

    DEFF Research Database (Denmark)

    Bønnelykke-Behrndtz, Marie Louise; Sørensen, Flemming Brandt; Damsgaard, Tine Engberg

    2008-01-01

    potential indicators of prognosis. Sixty patients who underwent surgery at the Department of Plastic Surgery, Aarhus University Hospital, from 1991 to 1994 were included in the study. Total tumor volume was estimated by the Cavalieri technique, total number of tumor cells by the optical dissector principle...... showed a significant impact on both disease-free survival (p=0.001) and mortality (p=0.009). In conclusion, tumor volume and total number of cancer cells were highly reproducible but did not add additional, independent prognostic information regarding the study population.......Stereological quantification of tumor volume, total number of tumor cells and mean nuclear volume provides unbiased data, regardless of the three-dimensional shape of the melanocytic lesion. The aim of the present study was to investigate whether these variables are reproducible and may represent...

  19. Proceedings of the 1984 DOE nuclear reactor and facility safety conference. Volume II

    Energy Technology Data Exchange (ETDEWEB)

    1984-01-01

    This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces.

  20. Operating manual for the High Flux Isotope Reactor. Volume I. Description of the facility

    International Nuclear Information System (INIS)

    1982-09-01

    This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report

  1. Operating manual for the High Flux Isotope Reactor. Volume I. Description of the facility

    Energy Technology Data Exchange (ETDEWEB)

    1982-09-01

    This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report.

  2. Proceedings of the 1984 DOE nuclear reactor and facility safety conference. Volume II

    International Nuclear Information System (INIS)

    1984-01-01

    This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces

  3. VIPRE-01. a thermal-hydraulic analysis code for reactor cores. Volume 1. Mathematical modeling

    International Nuclear Information System (INIS)

    Stewart, C.W.; Cuta, J.M.; Koontz, A.S.; Kelly, J.M.; Basehore, K.L.; George, T.L.; Rowe, D.S.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 1: Mathematical Modeling) explains the major thermal hydraulic models and supporting correlations in detail

  4. Operating reactors licensing actions summary. Volume 5, No. 9

    International Nuclear Information System (INIS)

    1985-11-01

    This document is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

  5. Operating reactors licensing actions summary. Volume 5, No. 8

    International Nuclear Information System (INIS)

    1985-10-01

    This summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

  6. Operating reactors licensing actions summary. Volume 4, No. 9

    International Nuclear Information System (INIS)

    1984-11-01

    This document is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the division of licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

  7. Study of advanced fission power reactor development for the United States. Volume I

    International Nuclear Information System (INIS)

    1976-01-01

    This volume summarizes the results and conclusions of an assessment of five advanced fission power reactor concepts in the context of potential nuclear power economies developed over the time period 1975 to 2020. The study was based on the premise that the LMFBR program has been determined to be the highest priority fission reactor program and it will proceed essentially as planned. Accepting this fact, the overall objective of the study was to provide evaluations of advanced fission reactor systems for input to evaluating the levels of research and development funding for fission power. Evaluation of the reactor systems included the following categories: (1) power plant performance, (2) fuel resource utilization; (3) fuel-cycle requirements; (4) economics; (5) environmental impact; (6) risk to the public; and (7) R and D requirements to achieve commercial status. The specific major objectives of the study were twofold: (1) to parametrically assess the impact of various reactor types for various levels of power demand through the year 2020 on fissile fuel utilization, economics, and the environment, based on varying but reasonable assumptions on the rates of installation; and (2) to qualitatively assess the practicality of the advanced reactor concepts, and their research and development. The reactor concepts examined were limited to the following: advanced high-temperature, gas-cooled reactor (HTGR) systems including the thorium/U-233 fuel cycle, gas turbine, and binary cycle (BIHTGR); gas-cooled fast breeder reactor (GCFR); molten salt breeder reactor (MSBR); light water breeder reactor (LWBR); and CANDU heavy water reactor

  8. SCALE-4 analysis of pressurized water reactor critical configurations. Volume 1: Summary

    International Nuclear Information System (INIS)

    DeHart, M.D.

    1995-03-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original fresh composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized water reactors (PWR). The analysis methodology utilized for all calculations in this report is based on the modules and data associated with the SCALE-4 code system. Each of the five volumes comprising this report provides an overview of the methodology applied. Subsequent volumes also describe in detail the approach taken in performing criticality calculations for these PWR configurations: Volume 2 describes criticality calculations for the Tennessee Valley Authority's Sequoyah Unit 2 reactor for Cycle 3; Volume 3 documents the analysis of Virginia Power's Surry Unit 1 reactor for the Cycle 2 core; Volume 4 documents the calculations performed based on GPU Nuclear Corporation's Three Mile Island Unit 1 Cycle 5 core; and, lastly, Volume 5 describes the analysis of Virginia Power's North Anna Unit 1 Cycle 5 core. Each of the reactor-specific volumes provides the details of calculations performed to determine the effective multiplication factor for each reactor core for one or more critical configurations using the SCALE-4 system; these results are summarized in this volume. Differences between the core designs and their possible impact on the criticality calculations are also discussed. Finally, results are presented for additional analyses performed to verify that solutions were sufficiently converged

  9. Fast reactor safety: proceedings of the international topical meeting. Volume 2

    International Nuclear Information System (INIS)

    1985-07-01

    The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 2 include: safety design concepts; operational transient experiments; analysis of seismic and external events; HCDA-related codes, analysis, and experiments; sodium fires; instrumentation and control/PPS design; whole-core accident analysis codes; and impact of safety design considerations on future LMFBR developments

  10. Fast reactor safety: proceedings of the international topical meeting. Volume 2. [R

    Energy Technology Data Exchange (ETDEWEB)

    1985-07-01

    The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 2 include: safety design concepts; operational transient experiments; analysis of seismic and external events; HCDA-related codes, analysis, and experiments; sodium fires; instrumentation and control/PPS design; whole-core accident analysis codes; and impact of safety design considerations on future LMFBR developments.

  11. FMDP Reactor Alternative Summary Report: Volume 3 - partially complete LWR alternative

    International Nuclear Information System (INIS)

    Greene, S.R.; Fisher, S.E.; Bevard, B.B.

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 3 of a four volume report summarizes the results of these analyses for the partially complete LWR (PCLWR) reactor based plutonium disposition alternative

  12. FMDP Reactor Alternative Summary Report: Volume 3 - partially complete LWR alternative

    Energy Technology Data Exchange (ETDEWEB)

    Greene, S.R.; Fisher, S.E.; Bevard, B.B. [and others

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 3 of a four volume report summarizes the results of these analyses for the partially complete LWR (PCLWR) reactor based plutonium disposition alternative.

  13. Analysis of the neutron flux in an annular pulsed reactor by using finite volume method

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Mário A.B. da; Narain, Rajendra; Bezerra, Jair de L., E-mail: mabs500@gmail.com, E-mail: narain@ufpe.br, E-mail: jairbezerra@gmail.com [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Centro de Tecnologia e Geociências. Departamento de Energia Nuclear

    2017-07-01

    Production of very intense neutron sources is important for basic nuclear physics and for material testing and isotope production. Nuclear reactors have been used as sources of intense neutron fluxes, although the achievement of such levels is limited by the inability to remove fission heat. Periodic pulsed reactors provide very intense fluxes by a rotating modulator near a subcritical core. A concept for the production of very intense neutron fluxes that combines features of periodic pulsed reactors and steady state reactors was proposed by Narain (1997). Such a concept is known as Very Intense Continuous High Flux Pulsed Reactor (VICHFPR) and was analyzed by using diffusion equation with moving boundary conditions and Finite Difference Method with Crank-Nicolson formalism. This research aims to analyze the flux distribution in the Very Intense Continuous Flux High Pulsed Reactor (VICHFPR) by using the Finite Volume Method and compares its results with those obtained by the previous computational method. (author)

  14. Analysis of the neutron flux in an annular pulsed reactor by using finite volume method

    International Nuclear Information System (INIS)

    Silva, Mário A.B. da; Narain, Rajendra; Bezerra, Jair de L.

    2017-01-01

    Production of very intense neutron sources is important for basic nuclear physics and for material testing and isotope production. Nuclear reactors have been used as sources of intense neutron fluxes, although the achievement of such levels is limited by the inability to remove fission heat. Periodic pulsed reactors provide very intense fluxes by a rotating modulator near a subcritical core. A concept for the production of very intense neutron fluxes that combines features of periodic pulsed reactors and steady state reactors was proposed by Narain (1997). Such a concept is known as Very Intense Continuous High Flux Pulsed Reactor (VICHFPR) and was analyzed by using diffusion equation with moving boundary conditions and Finite Difference Method with Crank-Nicolson formalism. This research aims to analyze the flux distribution in the Very Intense Continuous Flux High Pulsed Reactor (VICHFPR) by using the Finite Volume Method and compares its results with those obtained by the previous computational method. (author)

  15. The Osiris reactor. Descriptive report - Volume 1 - text

    International Nuclear Information System (INIS)

    1969-05-01

    Osiris is a pool type reactor with a 70 MW thermal power. Its main purpose is to irradiate under high flows of neutrons the materials of which future nuclear power stations are made. This report proposes a description of this pool reactor. A first part describes the functional aspects and general characteristics of all installations which are in principle definitely defined (premises, irradiation and experimentation equipment, water circuits, power supply, venting, controls). The second part addresses elements which are likely to be changed, and more particularly the reactor core: fuel elements and controls (uranium and boron load in different fuel element generations, experimental locations within the core), neutron transport aspects (calculation and experiment), and thermal aspects (power generation and removal) of the pile). The third part addresses the operation: operation cycles, stops, exploitation organisation [fr

  16. Relative blood volume changes underestimate total blood volume changes during hemodialysis

    NARCIS (Netherlands)

    Dasselaar, Judith J.; Lub-de Hooge, Marjolijn N.; Pruim, Jan; Nijnuis, Hugo; Wiersum, Anneke; de Jong, Paul E.; Huisman, Roel M.; Franssen, Casper F. M.

    Background: Measurements of relative blood volume changes (ARBV) during hemodialysis (HD) are based on hemoconcentration and assume uniform mixing of erythrocytes and plasma throughout the circulation. However, whole-body hematocrit (Ht) is lower than systemic Ht. During HD, a change in the ratio

  17. Safety report concerning the reactor Pegase - volume 1 - Description of the installation - volume 2 - Safety of the installations

    International Nuclear Information System (INIS)

    Lacour, J.

    1964-01-01

    In the first volume: This report is a description of the reactor Pegase, given with a view to examine the safety of the installations. The Cadarache site at which they are situated is briefly described, in particular because of the consequences on the techniques employed for building Pegase. A description is also given of the original aspects of the reactor. The independent loops which are designed for full-scale testing of fuel elements used in natural uranium-gas-graphite reactor systems are described in this report, together with their operational and control equipment. In the second volume: In the present report are examined the accidents which could cause damage to the Pegase reactor installation. Among possible causes of accidents considered are the seismicity of the region, an excessive power excursion of the reactor and a fracture in the sealing of an independent loop. Although all possible precautions have been taken to offset the effects of such accidents, their ultimate consequences are considered here. The importance is stressed of the security action and regulations which, added to the precautions taken for the construction, ensure the safety of the installations. (authors) [fr

  18. TOKMINA, Toroidal Magnetic Field Minimization for Tokamak Fusion Reactor. TOKMINA-2, Total Power for Tokamak Fusion Reactor

    International Nuclear Information System (INIS)

    Hatch, A.J.

    1975-01-01

    1 - Description of problem or function: TOKMINA finds the minimum magnetic field, Bm, required at the toroidal coil of a Tokamak type fusion reactor when the input is beta(ratio of plasma pressure to magnetic pressure), q(Kruskal-Shafranov plasma stability factor), and y(ratio of plasma radius to vacuum wall radius: rp/rw) and arrays of PT (total thermal power from both d-t and tritium breeding reactions), Pw (wall loading or power flux) and TB (thickness of blanket), following the method of Golovin, et al. TOKMINA2 finds the total power, PT, of such a fusion reactor, given a specified magnetic field, Bm, at the toroidal coil. 2 - Method of solution: TOKMINA: the aspect ratio(a) is minimized, giving a minimum value for Bm. TOKMINA2: a search is made for PT; the value of PT which minimizes Bm to the required value within 50 Gauss is chosen. 3 - Restrictions on the complexity of the problem: Input arrays presently are dimensioned at 20. This restriction can be overcome by changing a dimension card

  19. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 2. Conceptual balance of plant design

    Energy Technology Data Exchange (ETDEWEB)

    1979-11-01

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. This volume describes the conceptual balance-of-plant (BOP) design and was prepared by United Engineers and Constructors, Inc. of Philadelphia, Pennsylvania. The major emphasis of the BOP study was a preliminary design of an overall plant to provide a basis for future studies.

  20. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 2. Conceptual balance of plant design

    International Nuclear Information System (INIS)

    1979-11-01

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. This volume describes the conceptual balance-of-plant (BOP) design and was prepared by United Engineers and Constructors, Inc. of Philadelphia, Pennsylvania. The major emphasis of the BOP study was a preliminary design of an overall plant to provide a basis for future studies

  1. Efficiency of Worm Reactors in Reducing Sludge Volume in Activated Sludge Systems

    Directory of Open Access Journals (Sweden)

    Azam Naderi

    2017-01-01

    Full Text Available The activated sludge process is the most widely used on a global scale for the biological treatment of both domestic and industrial effluents. One problem associated with the process, however, is the high volume of sludge produced. Excess sludge treatment and disposal account for up to 60% of the total operating costs of urban wastewater treatment plants due to the stringent environmental regulations on excess sludge disposal. These strict requirements have encouraged a growing interest over the last few years in reducing sludge volumes produced at biological treatment plants and a number of physical, chemical, and mechanical methods have been accordingly developed for this purpose. The proposed methods are disadvantaged due to their rather high investment and operation costs. An alternative technology that avoids many of these limitations is the worm reactor. In this study, the characteristics of this technology are investigated while the related literature is reviewed to derive the optimal conditions for the operation of this process in different situations.

  2. The relationship of hospital charges and volume to surgical site infection after total hip replacement.

    Science.gov (United States)

    Boas, Rebecca; Ensor, Kelsey; Qian, Edward; Hutzler, Lorraine; Slover, James; Bosco, Joseph

    2015-05-01

    The purpose of this study was to analyze the effect of hospital volume and charges on the rate of surgical site infections for total hip replacements (THRs) in New York State (NYS). In NYS, higher volume hospitals have higher charges after THR. The study team analyzed 93,620 hip replacements performed in NYS between 2008 and 2011. Hospital charges increased significantly from $43,713 in 2008 to $50,652 in 2011 (P<.01). Compared with lower volume hospitals, patients who underwent THR at the highest volume hospitals had significantly lower surgical site infection rates (P=.003) and higher total hospital charges (P<.0001). The study team found that in the highest volume hospitals, preventing one surgical site infection was associated with $1.6 million dollars in increased charges. © 2014 by the American College of Medical Quality.

  3. Reactor safeguards system assessment and design. Volume I

    International Nuclear Information System (INIS)

    Varnado, G.B.; Ericson, D.M. Jr.; Daniel, S.L.; Bennett, H.A.; Hulme, B.L.

    1978-06-01

    This report describes the development and application of a methodology for evaluating the effectiveness of nuclear power reactor safeguards systems. Analytic techniques are used to identify the sabotage acts which could lead to release of radioactive material from a nuclear power plant, to determine the areas of a plant which must be protected to assure that significant release does not occur, to model the physical plant layout, and to evaluate the effectiveness of various safeguards systems. The methodology was used to identify those aspects of reactor safeguards systems which have the greatest effect on overall system performance and which, therefore, should be emphasized in the licensing process. With further refinements, the methodology can be used by the licensing reviewer to aid in assessing proposed or existing safeguards systems

  4. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 3. Appendix A. Equipment list

    International Nuclear Information System (INIS)

    1979-11-01

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. Volume 1 reports the overall plant and reactor system and was prepared by the General Electric Company. Core scoping studies were performed which evaluated the effects of annular and cylindrical core configurations, radial blanket zones, burnup, and ball heavy metal loadings. The reactor system, including the PCRV, was investigated for both the annular and cylindrical core configurations. Volume 3 is an Appendix containing the equipment list for the plant and was also prepared by United Engineers and Constructors, Inc. It tabulates the major components of the plant and describes each in terms of quantity, type, orientation, etc., to provide a basis for cost estimation

  5. The Effect of Structured Exercise Intervention on Intensity and Volume of Total Physical Activity

    Directory of Open Access Journals (Sweden)

    Niko Wasenius

    2014-12-01

    Full Text Available This study aimed to investigate the effects of a 12-week structured exercise intervention on total physical activity and its subcategories. Twenty-three overweight or obese middle aged men with impaired glucose regulation were randomized into a 12-week Nordic walking group, a power-type resistance training group, and a non-exercise control group. Physical activity was measured with questionnaires before the intervention (1–4 weeks and during the intervention (1–12 weeks and was expressed in metabolic equivalents of task. No significant change in the volume of total physical activity between or within the groups was observed (p > 0.050. The volume of total leisure-time physical activity (structured exercises + non-structured leisure-time physical activity increased significantly in the Nordic walking group (p 0.050 compared to the control group. In both exercise groups increase in the weekly volume of total leisure-time physical activity was inversely associated with the volume of non-leisure-time physical activities. In conclusion, structured exercise intervention did not increase the volume of total physical activity. Albeit, endurance training can increase the volume of high intensity physical activities, however it is associated with compensatory decrease in lower intensity physical activities. To achieve effective personalized exercise program, individuality in compensatory behavior should be recognised.

  6. VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 2. User's manual

    International Nuclear Information System (INIS)

    Cuta, J.M.; Koontz, A.S.; Stewart, C.W.; Montgomery, S.D.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear energy reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 2: User's Manual) describes the input requirements of VIPRE and its auxiliary programs, SPECSET, ASP and DECCON, and lists the input instructions for each code

  7. Tetrafluoroethane (R134a) hydrate formation within variable volume reactor accompanied by evaporation and condensation

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, K.; Choo, Y. S.; Hong, H. J.; Yoon, Y. S.; Song, M. H., E-mail: songm@dgu.edu [Department of Mechanical, Robotics, and Energy Engineering, Dongguk University, Seoul 100-715 (Korea, Republic of)

    2015-03-15

    Vast size hydrate formation reactors with fast conversion rate are required for the economic implementation of seawater desalination utilizing gas hydrate technology. The commercial target production rate is order of thousand tons of potable water per day per train. Various heat and mass transfer enhancement schemes including agitation, spraying, and bubbling have been examined to maximize the production capacities in scaled up design of hydrate formation reactors. The present experimental study focused on acquiring basic knowledge needed to design variable volume reactors to produce tetrafluoroethane hydrate slurry. Test vessel was composed of main cavity with fixed volume of 140 ml and auxiliary cavity with variable volume of 0 ∼ 64 ml. Temperatures at multiple locations within vessel and pressure were monitored while visual access was made through front window. Alternating evaporation and condensation induced by cyclic volume change provided agitation due to density differences among water and vapor, liquid and hydrate R134a as well as extended interface area, which improved hydrate formation kinetics coupled with latent heat release and absorption. Influences of coolant temperature, piston stroke/speed, and volume change period on hydrate formation kinetics were investigated. Suggestions of reactor design improvement for future experimental study are also made.

  8. Tetrafluoroethane (R134a) hydrate formation within variable volume reactor accompanied by evaporation and condensation

    International Nuclear Information System (INIS)

    Jeong, K.; Choo, Y. S.; Hong, H. J.; Yoon, Y. S.; Song, M. H.

    2015-01-01

    Vast size hydrate formation reactors with fast conversion rate are required for the economic implementation of seawater desalination utilizing gas hydrate technology. The commercial target production rate is order of thousand tons of potable water per day per train. Various heat and mass transfer enhancement schemes including agitation, spraying, and bubbling have been examined to maximize the production capacities in scaled up design of hydrate formation reactors. The present experimental study focused on acquiring basic knowledge needed to design variable volume reactors to produce tetrafluoroethane hydrate slurry. Test vessel was composed of main cavity with fixed volume of 140 ml and auxiliary cavity with variable volume of 0 ∼ 64 ml. Temperatures at multiple locations within vessel and pressure were monitored while visual access was made through front window. Alternating evaporation and condensation induced by cyclic volume change provided agitation due to density differences among water and vapor, liquid and hydrate R134a as well as extended interface area, which improved hydrate formation kinetics coupled with latent heat release and absorption. Influences of coolant temperature, piston stroke/speed, and volume change period on hydrate formation kinetics were investigated. Suggestions of reactor design improvement for future experimental study are also made

  9. FMDP reactor alternative summary report. Volume 1 - existing LWR alternative

    International Nuclear Information System (INIS)

    Greene, S.R.; Bevard, B.B.

    1996-01-01

    Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] are becoming surplus to national defense needs in both the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES ampersand H) consequences if surplus fissile materials are not properly managed. This document summarizes the results of analysis concerned with existing light water reactor plutonium disposition alternatives

  10. FMDP reactor alternative summary report. Volume 1 - existing LWR alternative

    Energy Technology Data Exchange (ETDEWEB)

    Greene, S.R.; Bevard, B.B. [and others

    1996-10-07

    Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] are becoming surplus to national defense needs in both the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES&H) consequences if surplus fissile materials are not properly managed. This document summarizes the results of analysis concerned with existing light water reactor plutonium disposition alternatives.

  11. Developmentally Sensitive Interaction Effects of Genes and the Social Environment on Total and Subcortical Brain Volumes.

    Science.gov (United States)

    Richards, Jennifer S; Arias Vásquez, Alejandro; Franke, Barbara; Hoekstra, Pieter J; Heslenfeld, Dirk J; Oosterlaan, Jaap; Faraone, Stephen V; Buitelaar, Jan K; Hartman, Catharina A

    2016-01-01

    Smaller total brain and subcortical volumes have been linked to psychopathology including attention-deficit/hyperactivity disorder (ADHD). Identifying mechanisms underlying these alterations, therefore, is of great importance. We investigated the role of gene-environment interactions (GxE) in interindividual variability of total gray matter (GM), caudate, and putamen volumes. Brain volumes were derived from structural magnetic resonance imaging scans in participants with (N = 312) and without ADHD (N = 437) from N = 402 families (age M = 17.00, SD = 3.60). GxE effects between DAT1, 5-HTT, and DRD4 and social environments (maternal expressed warmth and criticism; positive and deviant peer affiliation) as well as the possible moderating effect of age were examined using linear mixed modeling. We also tested whether findings depended on ADHD severity. Deviant peer affiliation was associated with lower caudate volume. Participants with low deviant peer affiliations had larger total GM volumes with increasing age. Likewise, developmentally sensitive GxE effects were found on total GM and putamen volume. For total GM, differential age effects were found for DAT1 9-repeat and HTTLPR L/L genotypes, depending on the amount of positive peer affiliation. For putamen volume, DRD4 7-repeat carriers and DAT1 10/10 homozygotes showed opposite age relations depending on positive peer affiliation and maternal criticism, respectively. All results were independent of ADHD severity. The presence of differential age-dependent GxE effects might explain the diverse and sometimes opposing results of environmental and genetic effects on brain volumes observed so far.

  12. Developmentally Sensitive Interaction Effects of Genes and the Social Environment on Total and Subcortical Brain Volumes.

    Directory of Open Access Journals (Sweden)

    Jennifer S Richards

    Full Text Available Smaller total brain and subcortical volumes have been linked to psychopathology including attention-deficit/hyperactivity disorder (ADHD. Identifying mechanisms underlying these alterations, therefore, is of great importance. We investigated the role of gene-environment interactions (GxE in interindividual variability of total gray matter (GM, caudate, and putamen volumes. Brain volumes were derived from structural magnetic resonance imaging scans in participants with (N = 312 and without ADHD (N = 437 from N = 402 families (age M = 17.00, SD = 3.60. GxE effects between DAT1, 5-HTT, and DRD4 and social environments (maternal expressed warmth and criticism; positive and deviant peer affiliation as well as the possible moderating effect of age were examined using linear mixed modeling. We also tested whether findings depended on ADHD severity. Deviant peer affiliation was associated with lower caudate volume. Participants with low deviant peer affiliations had larger total GM volumes with increasing age. Likewise, developmentally sensitive GxE effects were found on total GM and putamen volume. For total GM, differential age effects were found for DAT1 9-repeat and HTTLPR L/L genotypes, depending on the amount of positive peer affiliation. For putamen volume, DRD4 7-repeat carriers and DAT1 10/10 homozygotes showed opposite age relations depending on positive peer affiliation and maternal criticism, respectively. All results were independent of ADHD severity. The presence of differential age-dependent GxE effects might explain the diverse and sometimes opposing results of environmental and genetic effects on brain volumes observed so far.

  13. Tokamak experimental power reactor conceptual design. Volume II

    International Nuclear Information System (INIS)

    1976-08-01

    Volume II contains the following appendices: (1) summary of EPR design parameters, (2) impurity control, (3) plasma computational models, (4) structural support system, (5) materials considerations for the primary energy conversion system, (6) magnetics, (7) neutronics penetration analysis, (8) first wall stress analysis, (9) enrichment of isotopes of hydrogen by cryogenic distillation, and (10) noncircular plasma considerations

  14. Study of advanced fission power reactor development for the United States. Volume II

    International Nuclear Information System (INIS)

    1976-01-01

    This report presents the results of a multi-phase research study which had as its objective the comparative study of various advanced fission reactors and evaluation of alternate strategies for their development in the USA through the year 2020. By direction from NSF, ''advanced'' reactors were defined as those which met the dual requirements of (1) offering a significant improvement in fissile fuel utilization as compared to light-water reactors and (2) currently receiving U.S. Government funding. (A detailed study of the LMFBR was specifically excluded, but cursory baseline data were obtained from ERDA sources.) Included initially were the High-Temperature Gas-Cooled Reactor (HTGR), Gas-Cooled Fast Reactor (GCFR), Molten Salt Reactor (MSR), and Light-Water Breeder Reactor (LWBR). Subsequently, the CANDU Heavy Water Reactor (HWR) was included for comparison due to increased interest in its potential. This volume presents the reasoning process and analytical methods utilized to arrive at the conclusions for the overall study

  15. Fusion reactor control study. Volume 3. Tandem mirror reactors. Final report

    International Nuclear Information System (INIS)

    Chang, F.R.; DeCanio, F.; Fisher, J.L.; Madden, P.A.

    1982-03-01

    A study of the control requirements of the Tandem Mirror Reactor concept is reported. The study describes the development of a control simulator that is based upon a spatially averaged physics code of the reactor concept. The simulator portrays the evolution of the plasma through the complete reactor operating cycle; it includes models of the control and measurement system, thus allowing the exploration of various strategies for reactor control. Startup, shutdown, and control during the quasi-steady-state power producing phase were explored. Configurations are described which use a variety of control effectors including modulation of the refueling rate, beam current, and electron cyclotron resonance heating. Multivariable design techniques were used to design the control laws and compensators for the feedback controllers and presume the practical measurement of only a subset of the plasma and machine variables. Performance of the various controllers is explored using the nonlinear control simulator. Derivative control strategies using new or developed sensors and effectors appropriate to a power reactor environment are postulated, based upon the results of the control configurations tested. Research and development requirements for these controls are delineated

  16. Forecasting on the total volumes of Malaysia's imports and exports by multiple linear regression

    Science.gov (United States)

    Beh, W. L.; Yong, M. K. Au

    2017-04-01

    This study is to give an insight on the doubt of the important of macroeconomic variables that affecting the total volumes of Malaysia's imports and exports by using multiple linear regression (MLR) analysis. The time frame for this study will be determined by using quarterly data of the total volumes of Malaysia's imports and exports covering the period between 2000-2015. The macroeconomic variables will be limited to eleven variables which are the exchange rate of US Dollar with Malaysia Ringgit (USD-MYR), exchange rate of China Yuan with Malaysia Ringgit (RMB-MYR), exchange rate of European Euro with Malaysia Ringgit (EUR-MYR), exchange rate of Singapore Dollar with Malaysia Ringgit (SGD-MYR), crude oil prices, gold prices, producer price index (PPI), interest rate, consumer price index (CPI), industrial production index (IPI) and gross domestic product (GDP). This study has applied the Johansen Co-integration test to investigate the relationship among the total volumes to Malaysia's imports and exports. The result shows that crude oil prices, RMB-MYR, EUR-MYR and IPI play important roles in the total volumes of Malaysia's imports. Meanwhile crude oil price, USD-MYR and GDP play important roles in the total volumes of Malaysia's exports.

  17. Seismic design technology for Breeder Reactor structures. Volume 3: special topics in reactor structures

    International Nuclear Information System (INIS)

    Reddy, D.P.

    1983-04-01

    This volume is divided into six chapters: analysis techniques, equivalent damping values, probabilistic design factors, design verifications, equivalent response cycles for fatigue analysis, and seismic isolation

  18. Report on the Survey of the Design Review of New Reactor Applications. Volume 4: Reactor Coolant and Associated Systems

    International Nuclear Information System (INIS)

    Downey, Steven; Monninger, John; Nevalainen, Janne; Joyer, Philippe; Koley, Jaharlal; Kawamura, Tomonori; Chung, Yeon-Ki; Haluska, Ladislav; Persic, Andreja; Reierson, Craig; Monninger, John; Choi, Young-Joon; )

    2017-01-01

    At the tenth meeting of the Committee on Nuclear Regulatory Activities (CNRA) Working Group on the Regulation of New Reactors (WGRNR) in March 2013, the Working Group agreed to present the responses to the Second Phase, or Design Phase, of the licensing process survey as a multi-volume text. As such, each report will focus on one of the eleven general technical categories covered in the survey. The general technical categories were selected to conform to the topics covered in the International Atomic Energy Agency (IAEA) Safety Guide GS-G-4.1. This report provides a discussion of the survey responses related to the Reactor Coolant and Associated Systems category. The Reactor Coolant and Associated Systems category includes the following technical topics: overpressure protection, reactor coolant pressure boundary, reactor vessel, and design of the reactor coolant system. For each technical topic, the member countries described the information provided by the applicant, the scope and level of detail of the technical review, the technical basis for granting regulatory authorisation, the skill sets required and the level of effort needed to perform the review. Based on a comparison of the information provided by the member countries in response to the survey, the following observations were made: - Although the description of the information provided by the applicant differs in scope and level of detail among the member countries that provided responses, there are similarities in the information that is required. - All of the technical topics covered in the survey are reviewed in some manner by all of the regulatory authorities that provided responses. - It is common to consider operating experience and lessons learnt from the current fleet during the review process. - The most commonly and consistently identified technical expertise needed to perform design reviews related to this category are mechanical engineering and materials engineering. The complete survey

  19. Inertial Fusion Energy reactor design studies: Prometheus-L, Prometheus-H. Volume 2, Final report

    Energy Technology Data Exchange (ETDEWEB)

    Waganer, L.M.; Driemeyer, D.E.; Lee, V.D.

    1992-03-01

    This report contains a review of design studies for Inertial Confinement reactor. This second of three volumes discussions is some detail the following: Objectives, requirements, and assumptions; rationale for design option selection; key technical issues and R&D requirements; and conceptual design selection and description.

  20. Genetic Schizophrenia Risk Variants Jointly Modulate Total Brain and White Matter Volume

    DEFF Research Database (Denmark)

    Terwisscha van Scheltinga, Afke F; Bakker, Steven C; van Haren, Neeltje E M

    2013-01-01

    with total brain volume (R(2)=.048, p=1.6×10(-4)) and white matter volume (R(2)=.051, p=8.6×10(-5)) equally in patients and control subjects. The number of (independent) SNPs that substantially influenced both disease risk and white matter (n=2020) was much smaller than the entire set of SNPs that modulated...... modulating schizophrenia and brain volume. METHODS: Odds ratios for genome-wide SNP data were calculated in the sample collected by the Psychiatric Genome-wide Association Study Consortium (8690 schizophrenia patients and 11,831 control subjects, excluding subjects from the present study). These were used...

  1. Impact of the volume of gaseous phase in closed reactors on ANC results and modelling

    Science.gov (United States)

    Drapeau, Clémentine; Delolme, Cécile; Lassabatere, Laurent; Blanc, Denise

    2016-04-01

    The understanding of the geochemical behavior of polluted solid materials is often challenging and requires huge expenses of time and money. Nevertheless, given the increasing amounts of polluted solid materials and related risks for the environment, it is more and more crucial to understand the leaching of majors and trace metals elements from these matrices. In the designs of methods to quantify pollutant solubilization, the combination of experimental procedures with modeling approaches has recently gained attention. Among usual methods, some rely on the association of ANC and geochemical modeling. ANC experiments - Acid Neutralization Capacity - consists in adding known quantities of acid or base to a mixture of water and contaminated solid materials at a given liquid / solid ratio in closed reactors. Reactors are agitated for 48h and then pH, conductivity, redox potential, carbon, majors and heavy metal solubilized are quantified. However, in most cases, the amounts of matrix and water do not reach the total volume of reactors, leaving some space for air (gaseous phase). Despite this fact, no clear indication is given in standard procedures about the effect of this gaseous phase. Even worse, the gaseous phase is never accounted for when exploiting or modeling ANC data. The gaseous phase may exchange CO2 with the solution, which may, in turn, impact both pH and element release. This study lies within the most general framework for the use of geochemical modeling for the prediction of ANC results for the case of pure phases to real phase assemblages. In this study, we focus on the effect of the gaseous phase on ANC experiments on different mineral phases through geochemical modeling. To do so, we use PHREEQC code to model the evolution of pH and element release (including majors and heavy metals) when several matrices are put in contact with acid or base. We model the following scenarios for the gaseous phase: no gas, contact with the atmosphere (open system

  2. Energy-analysis of the total nuclear energy cycle based on light water reactors

    International Nuclear Information System (INIS)

    Kistemaker, J.

    1975-01-01

    The energy economy of the total nuclear energy cycle is investigated. Attention is paid to the importance of fossil fuel saving by using nuclear energy. The energy analysis is based on the construction and operation of power plants with an electric output of 1000MWe. Light water moderated reactors with a 2.7 - 3.2% enriched uranium core are considered. Additionally, the whole fuel cycle including ore winning and refining, enrichment and fuel element manufacturing and reprocessing has been taken into account. Neither radioactive waste storage problems nor safety problems related to the nuclear energy cycle and safeguarding have been dealt with, as exhaustive treatments can be found elswhere

  3. Code development of total sensitivity and uncertainty analysis for reactor physics calculations

    International Nuclear Information System (INIS)

    Wan, C.; Cao, L.; Wu, H.; Zu, T.; Shen, W.

    2015-01-01

    Sensitivity and uncertainty analysis are essential parts for reactor system to perform risk and policy analysis. In this study, total sensitivity and corresponding uncertainty analysis for responses of neutronics calculations have been accomplished and developed the S&U analysis code named UNICORN. The UNICORN code can consider the implicit effects of multigroup cross sections on the responses. The UNICORN code has been applied to typical pin-cell case in this paper, and can be proved correct by comparison the results with those of the TSUNAMI-1D code. (author)

  4. Code development of total sensitivity and uncertainty analysis for reactor physics calculations

    Energy Technology Data Exchange (ETDEWEB)

    Wan, C.; Cao, L.; Wu, H.; Zu, T., E-mail: chenghuiwan@stu.xjtu.edu.cn, E-mail: caolz@mail.xjtu.edu.cn, E-mail: hongchun@mail.xjtu.edu.cn, E-mail: tiejun@mail.xjtu.edu.cn [Xi' an Jiaotong Univ., School of Nuclear Science and Technology, Xi' an (China); Shen, W., E-mail: Wei.Shen@cnsc-ccsn.gc.ca [Xi' an Jiaotong Univ., School of Nuclear Science and Technology, Xi' an (China); Canadian Nuclear Safety Commission, Ottawa, ON (Canada)

    2015-07-01

    Sensitivity and uncertainty analysis are essential parts for reactor system to perform risk and policy analysis. In this study, total sensitivity and corresponding uncertainty analysis for responses of neutronics calculations have been accomplished and developed the S&U analysis code named UNICORN. The UNICORN code can consider the implicit effects of multigroup cross sections on the responses. The UNICORN code has been applied to typical pin-cell case in this paper, and can be proved correct by comparison the results with those of the TSUNAMI-1D code. (author)

  5. Sodium fast reactor safety and licensing research plan. Volume I.

    Energy Technology Data Exchange (ETDEWEB)

    Sofu, Tanju (Argonne National Laboratory, Argonne, IL); LaChance, Jeffrey L.; Bari, R. (Brokhaven National Laboratory Upton, NY); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.; Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN)

    2012-05-01

    This report proposes potential research priorities for the Department of Energy (DOE) with the intent of improving the licensability of the Sodium Fast Reactor (SFR). In support of this project, five panels were tasked with identifying potential safety-related gaps in available information, data, and models needed to support the licensing of a SFR. The areas examined were sodium technology, accident sequences and initiators, source term characterization, codes and methods, and fuels and materials. It is the intent of this report to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for the SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the Applied Technology designation from old documents. The second cross-cutting gap is the need for a robust Knowledge Management and Preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with Applied Technology and Knowledge Management.

  6. Tokamak experimental power reactor conceptual design. Volume I

    International Nuclear Information System (INIS)

    1976-08-01

    A conceptual design has been developed for a tokamak Experimental Power Reactor to operate at net electrical power conditions with a plant capacity factor of 50 percent for 10 years. The EPR operates in a pulsed mode at a frequency of approximately 1/min., with an approximate 75 percent duty cycle, is capable of producing approximately 72 MWe and requires 42 MWe. The annual tritium consumption is 16 kg. The EPR vacuum chamber is 6.25 m in major radius and 2.4 m in minor radius, is constructed of 2-cm thick stainless steel, and has 2-cm thick detachable, beryllium-coated coolant panels mounted on the interior. An 0.28 m stainless steel blanket and a shield ranging from 0.6 to 1.0 m surround the vacuum vessel. The coolant is H 2 O. Sixteen niobium-titanium superconducting toroidal-field coils provide a field of 10 T at the coil and 4.47 T at the plasma. Superconducting ohmic-heating and equilibrium-field coils provide 135 V-s to drive the plasma current. Plasma heating is accomplished by 12 neutral beam-injectors, which provide 60 MW. The energy transfer and storage system consists of a central superconducting storage ring, a homopolar energy storage unit, and a variety of inductor-converters

  7. Sodium fast reactor safety and licensing research plan - Volume I

    International Nuclear Information System (INIS)

    Sofu, Tanju; LaChance, Jeffrey L.; Bari, R.; Wigeland, Roald; Denman, Matthew R.; Flanagan, George F.

    2012-01-01

    This report proposes potential research priorities for the Department of Energy (DOE) with the intent of improving the licensability of the Sodium Fast Reactor (SFR). In support of this project, five panels were tasked with identifying potential safety-related gaps in available information, data, and models needed to support the licensing of a SFR. The areas examined were sodium technology, accident sequences and initiators, source term characterization, codes and methods, and fuels and materials. It is the intent of this report to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for the SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the Applied Technology designation from old documents. The second cross-cutting gap is the need for a robust Knowledge Management and Preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with Applied Technology and Knowledge Management.

  8. Total brain, cortical and white matter volumes in children previously treated with glucocorticoids

    DEFF Research Database (Denmark)

    Holm, Sara K; Madsen, Kathrine S; Vestergaard, Martin

    2018-01-01

    BACKGROUND: Perinatal exposure to glucocorticoids and elevated endogenous glucocorticoid-levels during childhood can have detrimental effects on the developing brain. Here, we examined the impact of glucocorticoid-treatment during childhood on brain volumes. METHODS: Thirty children and adolescents...... with rheumatic or nephrotic disease previously treated with glucocorticoids and 30 controls matched on age, sex, and parent education underwent magnetic resonance imaging (MRI) of the brain. Total cortical grey and white matter, brain, and intracranial volume, and total cortical thickness and surface area were...... were mainly driven by the children with rheumatic disease. Total cortical thickness and cortical surface area did not significantly differ between groups. We found no significant associations between glucocorticoid-treatment variables and volumetric measures. CONCLUSION: Observed smaller total brain...

  9. Skeletal and total body volumes of human fetuses: assessment of reference data by spiral CT

    International Nuclear Information System (INIS)

    Braillon, Pierre M.; Buenerd, Annie; Bouvier, Raymonde; Lapillonne, Alexandre

    2002-01-01

    Objective: To define reference data for skeletal and total body volumes of normal human fetuses. Materials and methods: Spiral CT was used to assess the skeletal and total body volumes of 31 normal human stillborn infants with gestational age (GA) and body weight (BW) ranging from 14 to 41.5 weeks and 22 to 3,760 g, respectively. CT scans (slice thickness 2.7 mm, pitch 0.7) were performed within the first 24 h after delivery. Precise bone and soft-tissue windows were defined from analysis of the density along the diaphysis of the fetal long bones and from the measurement of a phantom that mimics soft tissues. Lengths and volumes were obtained from 3D reconstructions. The femur lengths measured from CT images (FLct) were compared with those provided by US studies (FLus). Results: Significant correlations (r>0.9) were found between BW, measured volumes of the entire skeleton or head, long-bone lengths, biparietal diameter and GA. Strong linear correlations (r>0.98) were observed between FLct and FLus. Conclusions: Skeletal and total body volume values obtained using spiral CT were significantly correlated with fetal biometric measurements. These data could complement those obtained in obstetric investigations with US. (orig.)

  10. Measurement of thyroid volume, iodine concentration and total iodine content by CT and its clinical significance

    International Nuclear Information System (INIS)

    Nakaji, Shunsuke; Imanishi, Yoshimasa; Okamoto, Kyouko; Shinagawa, Toshihito

    2007-01-01

    Recently, Imanishi et al have developed new CT software for quantitative in vivo measurement of thyroid iodine. Using a CT system with the software, we measured volume, iodine concentration and total iodine content of thyroids in 63 controls and 435 patients with various diffuse thyroid diseases and thyroid nodules. In controls, all of them showed no difference between the sexes. Although the iodine concentration of the thyroid showed no difference among children, adults and seniles, the volume and total iodine content of the thyroid appeared smaller in children and seniles than in adults. In addition, although the volume and iodine concentration of the thyroid had two peaks in distribution, the total iodine content had almost normal distribution. Normal range of volume, iodine concentration and total iodine content in adults were 5.2-15.5 cm 3 , 0.28831-0.85919 mg/cm 3 and 2.35-11.69 mg, respectively. In thyroid nodule, there is no significant difference in volume, iodine concentration and total iodine content between benign and malignant nodules. All nodules with iodine concentration of less than 0.00007 mg/cm 3 were benign. No thyroid was higher in iodine concentration than the normal range although the thyroid was lower in 78.7% of patients with diffuse thyroid diseases. In all thyroids with increasing iodine concentration and total iodine content in medication course, thyroidal symptoms and signs were uncontrollable by the medication. In 43.8% of patients with long-period systemic diseases, the thyroid showed abnormality in any of the three. We concluded that quantitative in vivo measurement of thyroid iodine by CT could assist the diagnosis of thyroid diseases and decision of therapeutic methods. (author)

  11. Extracellular space, blood volume, and the early dumping syndrome after total gastrectomy

    Energy Technology Data Exchange (ETDEWEB)

    Miholic, J.; Reilmann, L.; Meyer, H.J.; Koerber, H.K.; Kotzerke, J.; Hecker, H. (Medzinische Hochschule Hannover (Germany, F.R.))

    1990-10-01

    Extracellular space and blood volume were measured using 82Br dilution and 51Cr-tagged erythrocytes in 24 tumor-free patients after total gastrectomy. Eleven of the patients suffered from early dumping. Age, blood volume, and extracellular space were significantly smaller in dumpers (P less than 0.05). The dumping score could be predicted by a multiple regression model considering blood volume per lean body mass and extracellular space (r = 0.637; P = 0.0039). Rapid (t1/2 less than 360 seconds) emptying of the gastric substitute, assessed using a 99Tc-labeled solid test meal, was significantly associated with dumping in addition to extracellular space and blood volume (r = 0.876; P = 0.0018). Both rapid emptying and a narrow extracellular space seem to contribute to the early dumping syndrome.

  12. A Novel Grey Wave Method for Predicting Total Chinese Trade Volume

    Directory of Open Access Journals (Sweden)

    Kedong Yin

    2017-12-01

    Full Text Available The total trade volume of a country is an important way of appraising its international trade situation. A prediction based on trade volume will help enterprises arrange production efficiently and promote the sustainability of the international trade. Because the total Chinese trade volume fluctuates over time, this paper proposes a Grey wave forecasting model with a Hodrick–Prescott filter (HP filter to forecast it. This novel model first parses time series into long-term trend and short-term cycle. Second, the model uses a general GM (1,1 to predict the trend term and the Grey wave forecasting model to predict the cycle term. Empirical analysis shows that the improved Grey wave prediction method provides a much more accurate forecast than the basic Grey wave prediction method, achieving better prediction results than autoregressive moving average model (ARMA.

  13. 29 CFR 779.253 - What is included in computing the total annual inflow volume.

    Science.gov (United States)

    2010-07-01

    ... FAIR LABOR STANDARDS ACT AS APPLIED TO RETAILERS OF GOODS OR SERVICES Employment to Which the Act May... taxes and other charges which the enterprise must pay for such goods. Generally, all charges will be... computing the total annual inflow volume. The goods which the establishment purchases or receives for resale...

  14. Effects of fluid communications between fluid volumes on the seismic behaviour of nuclear breeder reactor internals

    International Nuclear Information System (INIS)

    Durandet, E.; Gibert, R.J.

    1987-01-01

    The internal structures of a breeder reactor as SUPERPHENIX are mainly axisymmetrial shells separated by fluid volumes which are connected by small communications holes. These communications can destroy the axisymmetry of the problem and their effects on the inertial terms due to the fluid are important. An equivalent axisymmetrical element based on a local tridimensional solution in the vicinity of the fluid communication is defined. An axisymmetrical modelization using this type of element is built in order to calculate the horizontal seismic behaviour of the reactor internals. The effect due to three typical fluid communications are studied and compared. (orig.)

  15. Does early tetralogy of Fallot total correction give better final lung volumes?

    Science.gov (United States)

    Sadeghi, Hasan Allah; Miri, Seyed Reza; Bakhshandeh, Hooman; Mirmesdagh, Yalda; Paziraee, Nazita

    2013-06-01

    Pulmonary blood flow may affect lung development in adulthood. Early total correction of tetralogy of Fallot may affect development of final lung volumes. We evaluated the effect of age at total correction on lung volumes years after the operation. In a retrospective cohort study on patients with totally corrected tetralogy of Fallot (mean age, 13.40 years at the time of follow-up), forced vital capacity, slow vital capacity, forced expiratory volume in 1 s, and other parameters were measured 154.8 ± 46.25 months after the operation. Comparison were made of 3 groups: ≤2-, 2-8-, and >8-years old at the time of total correction surgery. Among 322 enrolled patients, the mean values of the follow-up spirometry results in ≤2-, 2-8-, >8-year-olds and the percentage of predicted values were respectively: vital capacity: 4.46 ± 0.57 L (107% ± 10.96%), 3.89 ± 0.58 L (91.10% ± 12.25%), 3.25 ± 0.48 L (82.35% ± 10.62%), p volume in 1 s: 4.22 ± 0.63 L (104.84% ± 13.64%), 3.66 ± 0.58 L (90.61% ± 12.59%), 3.02 ± 0.48 L (84.31% ± 12%), p volumes and capacities. It is better to consider total correction for all tetralogy of Fallot patients below 2-years old, or at least below 8-years old, if it is technically possible.

  16. Tranexamic Acid Reduced the Percent of Total Blood Volume Lost During Adolescent Idiopathic Scoliosis Surgery.

    Science.gov (United States)

    Jones, Kristen E; Butler, Elissa K; Barrack, Tara; Ledonio, Charles T; Forte, Mary L; Cohn, Claudia S; Polly, David W

    2017-01-01

    Multilevel posterior spine fusion is associated with significant intraoperative blood loss. Tranexamic acid is an antifibrinolytic agent that reduces intraoperative blood loss. The goal of this study was to compare the percent of total blood volume lost during posterior spinal fusion (PSF) with or without tranexamic acid in patients with adolescent idiopathic scoliosis (AIS). Thirty-six AIS patients underwent PSF in 2011-2014; the last half (n=18) received intraoperative tranexamic acid. We retrieved relevant demographic, hematologic, intraoperative and outcomes information from medical records. The primary outcome was the percent of total blood volume lost, calculated from estimates of intraoperative blood loss (numerator) and estimated total blood volume per patient (denominator, via Nadler's equations). Unadjusted outcomes were compared using standard statistical tests. Tranexamic acid and no-tranexamic acid groups were similar (all p>0.05) in mean age (16.1 vs. 15.2 years), sex (89% vs. 83% female), body mass index (22.2 vs. 20.2 kg/m2), preoperative hemoglobin (13.9 vs. 13.9 g/dl), mean spinal levels fused (10.5 vs. 9.6), osteotomies (1.6 vs. 0.9) and operative duration (6.1 hours, both). The percent of total blood volume lost (TBVL) was significantly lower in the tranexamic acid-treated vs. no-tranexamic acid group (median 8.23% vs. 14.30%, p = 0.032); percent TBVL per level fused was significantly lower with tranexamic acid than without it (1.1% vs. 1.8%, p=0.048). Estimated blood loss (milliliters) was similar across groups. Tranexamic acid significantly reduced the percentage of total blood volume lost versus no tranexamic acid in AIS patients who underwent PSF using a standardized blood loss measure.Level of Evidence: 3. Institutional Review Board status: This medical record chart review (minimal risk) study was approved by the University of Minnesota Institutional Review Board.

  17. 21 CFR 201.323 - Aluminum in large and small volume parenterals used in total parenteral nutrition.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 4 2010-04-01 2010-04-01 false Aluminum in large and small volume parenterals... for Specific Drug Products § 201.323 Aluminum in large and small volume parenterals used in total parenteral nutrition. (a) The aluminum content of large volume parenteral (LVP) drug products used in total...

  18. Effects of respiratory rate and tidal volume on gas exchange in total liquid ventilation.

    Science.gov (United States)

    Bull, Joseph L; Tredici, Stefano; Fujioka, Hideki; Komori, Eisaku; Grotberg, James B; Hirschl, Ronald B

    2009-01-01

    Using a rabbit model of total liquid ventilation (TLV), and in a corresponding theoretical model, we compared nine tidal volume-respiratory rate combinations to identify a ventilator strategy to maximize gas exchange, while avoiding choked flow, during TLV. Nine different ventilation strategies were tested in each animal (n = 12): low [LR = 2.5 breath/min (bpm)], medium (MR = 5 bpm), or high (HR = 7.5 bpm) respiratory rates were combined with a low (LV = 10 ml/kg), medium (MV = 15 ml/kg), or high (HV = 20 ml/kg) tidal volumes. Blood gases and partial pressures, perfluorocarbon gas content, and airway pressures were measured for each combination. Choked flow occurred in all high respiratory rate-high volume animals, 71% of high respiratory rate-medium volume (HRMV) animals, and 50% of medium respiratory rate-high volume (MRHV) animals but in no other combinations. Medium respiratory rate-medium volume (MRMV) resulted in the highest gas exchange of the combinations that did not induce choke. The HRMV and MRHV animals that did not choke had similar or higher gas exchange than MRMV. The theory predicted this behavior, along with spatial and temporal variations in alveolar gas partial pressures. Of the combinations that did not induce choked flow, MRMV provided the highest gas exchange. Alveolar gas transport is diffusion dominated and rapid during gas ventilation but is convection dominated and slow during TLV. Consequently, the usual alveolar gas equation is not applicable for TLV.

  19. Total energy supply-system for manned spaceship using nuclear reactor

    International Nuclear Information System (INIS)

    Narabayashi, Tadashi; Honma, Yuji; Yoshida, Yutaka; Shimazu, Yoichiro

    2007-01-01

    In order to explore the deep space, such as Mars, Jupiter, Saturn, etc in the future, a spacecraft that will be driven by nuclear power should be developed. At present, satellites or space probes have been using mainly electric source of chemical battery, fuel battery, solar battery, and RI battery. However, considering highly developed and extensive space exploration in the future, it is obvious that larger electric power is required over the long term space travel more than several years. Additionally, the solar battery used in space will be fundamentally impossible to use in planetary exploration father away form Mars because sunlight is attenuated. Therefore, larger electric power source must be installed in the space craft. In this study, we consider about co-generation system for heat and electricity using nuclear power. We think that the nuclear power is appropriate for using in deep space because of a long time operation without refueling and possibility in downsizing due to higher power density. We selected the fast reactor system of about 18 MWth compared with other type of reactors, such as PWR and high temperature gas reactor (Honma, 2006). With regard to a power generation system, we examined about efficiency of Stirling engine compared with a gas-turbine engine. Theoretical efficiency of Stirling engine is much higher than that of gas-turbine engine. Therefore, we selected Stirling engine and we have started the model test of a Stirling engine. Total power generation at International Space Station (ISS) that has been built since 1998 is about 110kWe. We estimated that about 5 times as much electricity as that of ISS is enough to explore or developed the space. In that case, 2.5MWe will be generated by the system, number of crews will be about 10 and 2MW will be used to electric propulsion. (author)

  20. Occupational radiation exposure at Commercial Nuclear Power reactors 1983. Volume 5. Annual report

    International Nuclear Information System (INIS)

    Brooks, B.G.

    1985-03-01

    This report presents an updated compilation of occupational radiation exposure at commercial nuclear power reactors for the years 1969 through 1983. The summary based on information received from the 75 light-water-cooled reactors (LWRs) and one high temperature gas-cooled reactor (HTGR). The total number of personnel monitored at LWRs in 1983 was 136,700. The number of workers that received measurable doses during 1983 and 85,600 which is about 1000 more than that found in 1982. The total collective dose at LWRs for 1983 is estimated to be 56,500 man-rems (man-cSv), which is about 4000 more man-rems (man-cSv) than that reported in 1982. This resulted in the average annual dose for each worker who received a measurable dose increasing slightly to 0.66 rems (cSv), and the average collective dose per reactor increasing by about 50 man-rems (man-cSv), and the average collective dose per reactor increasing by about 50 man-rems (man-cSv) to a value of 753 man-rems (man-cSv). The collective dose per megawatt of electricity generated by each reactor also increased slightly to an average value of 1.7 man-rems (man-cSv) per megawatt-year. Health implications of these annual occupational doses are discussed

  1. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Volume 1, Summary description

    International Nuclear Information System (INIS)

    Gertman, D.I.; Gilmore, W.E.; Galyean, W.J.; Groh, M.R.; Gentillon, C.D.; Gilbert, B.G.

    1988-02-01

    A data management system has been implemented which supports a variety of risk-related analyses and provides a repository of hardware component failure and human error probability data to the risk analyst. The Nuclear Computerized Library for Assessing Reactor Reliability, NUCLARR, is an interactive, graphically oriented system which resides on a personal computer (PC) or PC-compatible environment. An overview of the data management system, including a description of data collection, specification, data structure, and taxonomies, is presented in Volume I of this report. Programming activities, procedures for processing data, user's guide, and hard copy data manual are presented in Volumes II through V

  2. Modeling the spatial distribution of the parameters of the coolant in the reactor volume

    International Nuclear Information System (INIS)

    Nikonov, S.P.

    2011-01-01

    In this paper the approach to the question about the spatial distribution of the parameters of the coolant in-reactor volume. To describe the in-core space is used specially developed preprocessor. When the work of the preprocessor in the first place, is recreated on the basis of available information (mostly-the original drawings) with high accuracy three-dimensional description of the structures of the reactor volume and, secondly, are prepared on this basis blocks input to the nodal system code improved estimate ATHLET, allows to take into account the hydrodynamic interaction between the spatial control volumes. As an example the special case of solutions of international standard problem on the reconstruction of the transition process in the third unit of the Kalinin nuclear power plant, due to the shutdown of one of the four Main Coolant Pumps in operation at the rated capacity (first download). Model-core area consists of approximately 58 000 control volumes and spatial relationships. It shows the influence of certain structural units of the core to the distribution of the mass floe rate of its height. It is detected a strong cross-flow coolant in the area over the baffle. Moreover, we study the distribution of the coolant temperature at the assembly head of WWER-1000 reactor. It is shown that in the region of the top of the assembly head, where we have installation of thermocouples, the flow coolant for internal assemblies core is formed by only from guide channel Reactor control and protected system Control rod flow, or a mixture of the guide channel flow and flow from the area in front of top grid head assembly (the peripheral assemblies). It is shown that the magnitude of the flow guide channels affects not only the position of control rods, but also the presence of a particular type of measuring channels (Self powered neutron detector sensors or Temperature control sensors) in the cassette. (Author)

  3. Mirror Advanced Reactor Study (MARS). Final report. Volume 2. Commercial fusion synfuels plant

    International Nuclear Information System (INIS)

    Donohue, M.L.; Price, M.E.

    1984-07-01

    Volume 2 contains the following chapters: (1) synfuels; (2) physics base and parameters for TMR; (3) high-temperature two-temperature-zone blanket system for synfuel application; (4) thermochemical hydrogen processes; (5) interfacing the sulfur-iodine cycle; (6) interfacing the reactor with the thermochemical process; (7) tritium control in the blanket system; (8) the sulfur trioxide fluidized-bed composer; (9) preliminary cost estimates; and (10) fuels beyond hydrogen

  4. Metal halides vapor lasers with inner reactor and small active volume.

    Science.gov (United States)

    Shiyanov, D. V.; Sukhanov, V. B.; Evtushenko, G. S.

    2018-04-01

    Investigation of the energy characteristics of copper, manganese, lead halide vapor lasers with inner reactor and small active volume 90 cm3 was made. The optimal operating pulse repetition rates, temperatures, and buffer gas pressure for gas discharge tubes with internal and external electrodes are determined. Under identical pump conditions, such systems are not inferior in their characteristics to standard metal halide vapor lasers. It is shown that the use of a zeolite halogen generator provides lifetime laser operation.

  5. MRI estimation of total renal volume demonstrates significant association with healthy donor weight

    International Nuclear Information System (INIS)

    Cohen, Emil I.; Kelly, Sarah A.; Edye, Michael; Mitty, Harold A.; Bromberg, Jonathan S.

    2009-01-01

    Purpose: The purpose of this study was to correlate total renal volume (TRV) calculations, obtained through the voxel-count method and ellipsoid formula with various physical characteristics. Materials and methods: MRI reports and physical examination from 210 healthy kidney donors (420 kidneys), on whom renal volumes were obtained using the voxel-count method, were retrospectively reviewed. These values along with ones obtained through a more traditional method (ellipsoid formula) were correlated with subject height, body weight, body mass index (BMI), and age. Results: TRV correlated strongly with body weight (r = 0.7) and to a lesser degree with height, age, or BMI (r = 0.5, -0.2, 0.3, respectively). The left kidney volume was greater than the right, on average (p < 0.001). The ellipsoid formula method over-estimated renal volume by 17% on average which was significant (p < 0.001). Conclusions: Body weight was the physical characteristic which demonstrated the strongest correlation with renal volume in healthy subjects. Given this finding, a formula was derived for estimating the TRV for a given patient based on the his or her weight: TRV = 2.96 x weight (kg) + 113 ± 64.

  6. System of large transport containers for waste from dismantling light water and gas-cooled nuclear reactors. Volume 1

    International Nuclear Information System (INIS)

    Price, M.S.T.; Lafontaine, I.

    1985-01-01

    The purpose of this volume is to introduce the main types of nuclear reactor in the European Community (EC), select reference plants for further study, estimate the waste streams from the reference reactors, survey the transport regulations and assess existing containers

  7. Active species in a large volume N2-O2 post-discharge reactor

    International Nuclear Information System (INIS)

    Kutasi, K; Pintassilgo, C D; Loureiro, J; Coelho, P J

    2007-01-01

    A large volume post-discharge reactor placed downstream from a flowing N 2 -O 2 microwave discharge is modelled using a three-dimensional hydrodynamic model. The density distributions of the most populated active species present in the reactor-O( 3 P), O 2 (a 1 Δ g ), O 2 (b 1 Σ g + ), NO(X 2 Π), NO(A 2 Σ + ), NO(B 2 Π), NO 2 (X), O 3 , O 2 (X 3 Σ g - ) and N( 4 S)-are calculated and the main source and loss processes for each species are identified for two discharge conditions: (i) p = 2 Torr, f = 2450 MHz, and (ii) p = 8 Torr, f = 915 MHz; in the case of a N 2 -2%O 2 mixture composition and gas flow rate of 2 x 10 3 sccm. The modification of the species relative densities by changing the oxygen percentage in the initial gas mixture composition, in the 0.2%-5% range, are presented. The possible tuning of the species concentrations in the reactor by changing the size of the connecting afterglow tube between the active discharge and the large post-discharge reactor is investigated as well

  8. Feasibility of Commercially Available, Fully Automated Hepatic CT Volumetry for Assessing Both Total and Territorial Liver Volumes in Liver Transplantation

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Cheong Il; Kim, Se Hyung; Rhim, Jung Hyo; Yi, Nam Joon; Suh, Kyung Suk; Lee, Jeong Min; Han, Joon Koo; Choi, Byung Ihn [Seoul National University Hospital, Seoul (Korea, Republic of)

    2013-02-15

    To assess the feasibility of commercially-available, fully automated hepatic CT volumetry for measuring both total and territorial liver volumes by comparing with interactive manual volumetry and measured ex-vivo liver volume. For the assessment of total and territorial liver volume, portal phase CT images of 77 recipients and 107 donors who donated right hemiliver were used. Liver volume was measured using both the fully automated and interactive manual methods with Advanced Liver Analysis software. The quality of the automated segmentation was graded on a 4-point scale. Grading was performed by two radiologists in consensus. For the cases with excellent-to-good quality, the accuracy of automated volumetry was compared with interactive manual volumetry and measured ex-vivo liver volume which was converted from weight using analysis of variance test and Pearson's or Spearman correlation test. Processing time for both automated and interactive manual methods was also compared. Excellent-to-good quality of automated segmentation for total liver and right hemiliver was achieved in 57.1% (44/77) and 17.8% (19/107), respectively. For both total and right hemiliver volumes, there were no significant differences among automated, manual, and ex-vivo volumes except between automate volume and manual volume of the total liver (p = 0.011). There were good correlations between automate volume and ex-vivo liver volume ({gamma}= 0.637 for total liver and {gamma}= 0.767 for right hemiliver). Both correlation coefficients were higher than those with manual method. Fully automated volumetry required significantly less time than interactive manual method (total liver: 48.6 sec vs. 53.2 sec, right hemiliver: 182 sec vs. 244.5 sec). Fully automated hepatic CT volumetry is feasible and time-efficient for total liver volume measurement. However, its usefulness for territorial liver volumetry needs to be improved.

  9. Feasibility of Commercially Available, Fully Automated Hepatic CT Volumetry for Assessing Both Total and Territorial Liver Volumes in Liver Transplantation

    International Nuclear Information System (INIS)

    Shin, Cheong Il; Kim, Se Hyung; Rhim, Jung Hyo; Yi, Nam Joon; Suh, Kyung Suk; Lee, Jeong Min; Han, Joon Koo; Choi, Byung Ihn

    2013-01-01

    To assess the feasibility of commercially-available, fully automated hepatic CT volumetry for measuring both total and territorial liver volumes by comparing with interactive manual volumetry and measured ex-vivo liver volume. For the assessment of total and territorial liver volume, portal phase CT images of 77 recipients and 107 donors who donated right hemiliver were used. Liver volume was measured using both the fully automated and interactive manual methods with Advanced Liver Analysis software. The quality of the automated segmentation was graded on a 4-point scale. Grading was performed by two radiologists in consensus. For the cases with excellent-to-good quality, the accuracy of automated volumetry was compared with interactive manual volumetry and measured ex-vivo liver volume which was converted from weight using analysis of variance test and Pearson's or Spearman correlation test. Processing time for both automated and interactive manual methods was also compared. Excellent-to-good quality of automated segmentation for total liver and right hemiliver was achieved in 57.1% (44/77) and 17.8% (19/107), respectively. For both total and right hemiliver volumes, there were no significant differences among automated, manual, and ex-vivo volumes except between automate volume and manual volume of the total liver (p = 0.011). There were good correlations between automate volume and ex-vivo liver volume (γ= 0.637 for total liver and γ= 0.767 for right hemiliver). Both correlation coefficients were higher than those with manual method. Fully automated volumetry required significantly less time than interactive manual method (total liver: 48.6 sec vs. 53.2 sec, right hemiliver: 182 sec vs. 244.5 sec). Fully automated hepatic CT volumetry is feasible and time-efficient for total liver volume measurement. However, its usefulness for territorial liver volumetry needs to be improved.

  10. Application of the integrated analysis of safety (IAS) to sequences of Total loss of feed water in a PWR Reactor; Aplicacion del Analisis Integrado de Seguridad (ISA) a Secuencias de Perdidas Total de Agua de Alimentacion en un Reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Moreno Chamorro, P.; Gallego Diaz, C.

    2011-07-01

    The main objective of this work is to show the current status of the implementation of integrated analysis of safety (IAS) methodology and its SCAIS associated tool (system of simulation codes for IAS) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

  11. On the controllability and run-away possibility of a totally free piston, pulsed compression reactor

    NARCIS (Netherlands)

    Roestenberg, T.; Glouchenkov, Maxim Joerjevisj; glushenkov, M.J.; Kronberg, Alexandre E.; van der Meer, Theodorus H.

    2010-01-01

    The pulsed compression reactor promises to be a compact, economical and energy efficient alternative to conventional chemical reactors. While its design and operation is similar to that of a free piston internal combustion engine, it does not benefit from any controllability through the load.

  12. Automated CT-based segmentation and quantification of total intracranial volume

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar, Carlos; Wahlund, Lars-Olof; Westman, Eric [Karolinska Institute, Department of Neurobiology, Care Sciences and Society (NVS), Division of Clinical Geriatrics, Stockholm (Sweden); Edholm, Kaijsa; Cavallin, Lena; Muller, Susanne; Axelsson, Rimma [Karolinska Institute, Department of Clinical Science, Intervention and Technology, Division of Medical Imaging and Technology, Stockholm (Sweden); Karolinska University Hospital in Huddinge, Department of Radiology, Stockholm (Sweden); Simmons, Andrew [King' s College London, Institute of Psychiatry, London (United Kingdom); NIHR Biomedical Research Centre for Mental Health and Biomedical Research Unit for Dementia, London (United Kingdom); Skoog, Ingmar [Gothenburg University, Department of Psychiatry and Neurochemistry, The Sahlgrenska Academy, Gothenburg (Sweden); Larsson, Elna-Marie [Uppsala University, Department of Surgical Sciences, Radiology, Akademiska Sjukhuset, Uppsala (Sweden)

    2015-11-15

    To develop an algorithm to segment and obtain an estimate of total intracranial volume (tICV) from computed tomography (CT) images. Thirty-six CT examinations from 18 patients were included. Ten patients were examined twice the same day and eight patients twice six months apart (these patients also underwent MRI). The algorithm combines morphological operations, intensity thresholding and mixture modelling. The method was validated against manual delineation and its robustness assessed from repeated imaging examinations. Using automated MRI software, the comparability with MRI was investigated. Volumes were compared based on average relative volume differences and their magnitudes; agreement was shown by a Bland-Altman analysis graph. We observed good agreement between our algorithm and manual delineation of a trained radiologist: the Pearson's correlation coefficient was r = 0.94, tICVml[manual] = 1.05 x tICVml[automated] - 33.78 (R{sup 2} = 0.88). Bland-Altman analysis showed a bias of 31 mL and a standard deviation of 30 mL over a range of 1265 to 1526 mL. tICV measurements derived from CT using our proposed algorithm have shown to be reliable and consistent compared to manual delineation. However, it appears difficult to directly compare tICV measures between CT and MRI. (orig.)

  13. Automated CT-based segmentation and quantification of total intracranial volume

    International Nuclear Information System (INIS)

    Aguilar, Carlos; Wahlund, Lars-Olof; Westman, Eric; Edholm, Kaijsa; Cavallin, Lena; Muller, Susanne; Axelsson, Rimma; Simmons, Andrew; Skoog, Ingmar; Larsson, Elna-Marie

    2015-01-01

    To develop an algorithm to segment and obtain an estimate of total intracranial volume (tICV) from computed tomography (CT) images. Thirty-six CT examinations from 18 patients were included. Ten patients were examined twice the same day and eight patients twice six months apart (these patients also underwent MRI). The algorithm combines morphological operations, intensity thresholding and mixture modelling. The method was validated against manual delineation and its robustness assessed from repeated imaging examinations. Using automated MRI software, the comparability with MRI was investigated. Volumes were compared based on average relative volume differences and their magnitudes; agreement was shown by a Bland-Altman analysis graph. We observed good agreement between our algorithm and manual delineation of a trained radiologist: the Pearson's correlation coefficient was r = 0.94, tICVml[manual] = 1.05 x tICVml[automated] - 33.78 (R 2 = 0.88). Bland-Altman analysis showed a bias of 31 mL and a standard deviation of 30 mL over a range of 1265 to 1526 mL. tICV measurements derived from CT using our proposed algorithm have shown to be reliable and consistent compared to manual delineation. However, it appears difficult to directly compare tICV measures between CT and MRI. (orig.)

  14. Chemistry in water reactors: operating experience and new developments. 2 volumes

    International Nuclear Information System (INIS)

    1994-01-01

    These proceedings of the International conference on chemistry in water reactors (Operating experience and new developments), Volume 1, are divided into 8 sessions bearing on: (session 1) Primary coolant activity, corrosion products (5 conferences), (session 2) Dose reduction (4 conferences), (session 3) New developments (4 conferences), poster session: Primary coolant chemistry (16 posters), (session 4) Decontamination (5 conferences), poster session (2 posters), (session 5) BWR-Operating experience (3 conferences), (session 6) BWR-Modelling of operating experience (4 conferences), (session 7) BWR-Basic studies (4 conferences), (session 8) BWR-New technologies (3 conferences)

  15. Synaptic vesicle exocytosis in hippocampal synaptosomes correlates directly with total mitochondrial volume

    Science.gov (United States)

    Ivannikov, Maxim V.; Sugimori, Mutsuyuki; Llinás, Rodolfo R.

    2012-01-01

    Synaptic plasticity in many regions of the central nervous system leads to the continuous adjustment of synaptic strength, which is essential for learning and memory. In this study, we show by visualizing synaptic vesicle release in mouse hippocampal synaptosomes that presynaptic mitochondria and specifically, their capacities for ATP production are essential determinants of synaptic vesicle exocytosis and its magnitude. Total internal reflection microscopy of FM1-43 loaded hippocampal synaptosomes showed that inhibition of mitochondrial oxidative phosphorylation reduces evoked synaptic release. This reduction was accompanied by a substantial drop in synaptosomal ATP levels. However, cytosolic calcium influx was not affected. Structural characterization of stimulated hippocampal synaptosomes revealed that higher total presynaptic mitochondrial volumes were consistently associated with higher levels of exocytosis. Thus, synaptic vesicle release is linked to the presynaptic ability to regenerate ATP, which itself is a utility of mitochondrial density and activity. PMID:22772899

  16. Safe total corporal contouring with large-volume liposuction for the obese patient.

    Science.gov (United States)

    Dhami, Lakshyajit D; Agarwal, Meenakshi

    2006-01-01

    The advent of the tumescent technique in 1987 allowed for safe total corporal contouring as an ambulatory, single-session megaliposuction with the patient under regional anesthesia supplemented by local anesthetic only in selected areas. Safety and aesthetic issues define large-volume liposuction as having a 5,000-ml aspirate, mega-volume liposuction as having an 8,000-ml aspirate, and giganto-volume liposuction as having an aspirate of 12,000 ml or more. Clinically, a total volume comprising 5,000 ml of fat and wetting solution aspirated during the procedure qualifies for megaliposuction/large-volume liposuction. Between September 2000 and August 2005, 470 cases of liposuction were managed. In 296 (63%) of the 470 cases, the total volume of aspirate exceeded 5 l (range, 5,000-22,000 ml). Concurrent limited or total-block lipectomy was performed in 70 of 296 cases (23.6%). Regional anesthesia with conscious sedation was preferred, except where liposuction targeted areas above the subcostal region (the upper trunk, lateral chest, gynecomastia, breast, arms, and face), or when the patient so desired. Tumescent infiltration was achieved with hypotonic lactated Ringer's solution, adrenalin, triamcinalone, and hyalase in all cases during the last one year of the series. This approach has clinically shown less tissue edema in the postoperative period than with conventional physiologic saline used in place of the Ringer's lactate solution. The amount injected varied from 1,000 to 8,000 ml depending on the size, site, and area. Local anesthetic was included only for the terminal portion of the tumescent mixture, wherever the subcostal regions were infiltrated. The aspirate was restricted to the unstained white/yellow fat, and the amount of fat aspirated did not have any bearing on the amount of solution infiltrated. There were no major complications, and no blood transfusions were administered. The hospital stay ranged from 8 to 24 h for both liposuction and liposuction

  17. Application of the integrated analysis of safety (ISA) to sequences of Total loss of feed water in a PWR Reactor

    International Nuclear Information System (INIS)

    Moreno Chamorro, P.; Gallego Diaz, C.

    2011-01-01

    The main objective of this work is to show the current status of the implementation of integrated analysis of safety (ISA) methodology and its SCAIS associated tool (system of simulation codes for ISA) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

  18. The role of surgeon volume on patient outcome in total knee arthroplasty: a systematic review of the literature

    Directory of Open Access Journals (Sweden)

    Lau Rick L

    2012-12-01

    Full Text Available Abstract Background A number of factors have been identified as influencing total knee arthroplasty outcomes, including patient factors such as gender and medical comorbidity, technical factors such as alignment of the prosthesis, and provider factors such as hospital and surgeon procedure volumes. Recently, strategies aimed at optimizing provider factors have been proposed, including regionalization of total joint arthroplasty to higher volume centers, and adoption of volume standards. To contribute to the discussions concerning the optimization of provider factors and proposals to regionalize total knee arthroplasty practices, we undertook a systematic review to investigate the association between surgeon volume and primary total knee arthroplasty outcomes. Methods We performed a systematic review examining the association between surgeon volume and primary knee arthroplasty outcomes. To be included in the review, the study population had to include patients undergoing primary total knee arthroplasty. Studies had to report on the association between surgeon volume and primary total knee arthroplasty outcomes, including perioperative mortality and morbidity, patient-reported outcomes, or total knee arthroplasty implant survivorship. There were no restrictions placed on study design or language. Results Studies were variable in defining surgeon volume (‘low’: 5 to >70 total knee arthroplasty per year. Mortality rate, survivorship and thromboembolic events were not found to be associated with surgeon volume. We found a significant association between low surgeon volume and higher rate of infection (0.26% - 2.8% higher, procedure time (165 min versus 135 min, longer length of stay (0.4 - 2.13 days longer, transfusion rate (13% versus 4%, and worse patient reported outcomes. Conclusions Findings suggest a trend towards better outcomes for higher volume surgeons, but results must be interpreted with caution.

  19. Experimental fusion power reactor conceptual design study. Final report. Volume II

    International Nuclear Information System (INIS)

    Baker, C.C.

    1976-12-01

    This document is the final report which describes the work carried out by General Atomic Company for the Electric Power Research Institute on a conceptual design study of a fusion experimental power reactor (EPR) and an overall EPR facility. The primary objective of the two-year program was to develop a conceptual design of an EPR that operates at ignition and produces continuous net power. A conceptual design was developed for a Doublet configuration based on indications that a noncircular tokamak offers the best potential of achieving a sufficiently high effective fuel containment to provide a viable reactor concept at reasonable cost. Other objectives included the development of a planning cost estimate and schedule for the plant and the identification of critical R and D programs required to support the physics development and engineering and construction of the EPR. This volume contains the following sections: (1) reactor components, (2) auxiliary systems, (3) operations, (4) facility design, (5) program considerations, and (6) conclusions and recommendations

  20. Nuclear piping criteria for Advanced Light-Water Reactors, Volume 1--Failure mechanisms and corrective actions

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    This WRC Bulletin concentrates on the major failure mechanisms observed in nuclear power plant piping during the past three decades and on corrective actions taken to minimize or eliminate such failures. These corrective actions are applicable to both replacement piping and the next generation of light-water reactors. This WRC Bulletin was written with the objective of meeting a need for piping criteria in Advanced Light-Water Reactors, but there is application well beyond the LWR industry. This Volume, in particular, is equally applicable to current nuclear power plants, fossil-fueled power plants, and chemical plants including petrochemical. Implementation of the recommendations for mitigation of specific problems should minimize severe failures or cracking and provide substantial economic benefit. This volume uses a case history approach to high-light various failure mechanisms and the corrective actions used to resolve such failures. Particular attention is given to those mechanisms leading to severe piping failures, where severe denotes complete severance, large ''fishmouth'' failures, or long throughwall cracks releasing a minimum of 50 gpm. The major failure mechanisms causing severe failure are erosion-corrosion and vibrational fatigue. Stress corrosion cracking also has been a common problem in nuclear piping systems. In addition thermal fatigue due to mixing-tee and to thermal stratification also is discussed as is microbiologically-induced corrosion. Finally, water hammer, which represents the ultimate in internally-generated dynamic high-energy loads, is discussed

  1. Current self-reported symptoms of attention deficit/hyperactivity disorder are associated with total brain volume in healthy adults.

    Directory of Open Access Journals (Sweden)

    Martine Hoogman

    Full Text Available BACKGROUND: Reduced total brain volume is a consistent finding in children with Attention Deficit/Hyperactivity Disorder (ADHD. In order to get a better understanding of the neurobiology of ADHD, we take the first step in studying the dimensionality of current self-reported adult ADHD symptoms, by looking at its relation with total brain volume. METHODOLOGY/PRINCIPAL FINDINGS: In a sample of 652 highly educated adults, the association between total brain volume, assessed with magnetic resonance imaging, and current number of self-reported ADHD symptoms was studied. The results showed an association between these self-reported ADHD symptoms and total brain volume. Post-hoc analysis revealed that the symptom domain of inattention had the strongest association with total brain volume. In addition, the threshold for impairment coincides with the threshold for brain volume reduction. CONCLUSIONS/SIGNIFICANCE: This finding improves our understanding of the biological substrates of self-reported ADHD symptoms, and suggests total brain volume as a target intermediate phenotype for future gene-finding in ADHD.

  2. Total-dose radiation effects data for semiconductor devices. 1985 supplement. Volume 2, part A

    International Nuclear Information System (INIS)

    Martin, K.E.; Gauthier, M.K.; Coss, J.R.; Dantas, A.R.V.; Price, W.E.

    1986-05-01

    Steady-state, total-dose radiation test data, are provided in graphic format for use by electronic designers and other personnel using semiconductor devices in a radiation environment. The data were generated by JPL for various NASA space programs. This volume provides data on integrated circuits. The data are presented in graphic, tabular, and/or narrative format, depending on the complexity of the integrated circuit. Most tests were done using the JPL or Boeing electron accelerator (Dynamitron) which provides a steady-state 2.5 MeV electron beam. However, some radiation exposures were made with a cobalt-60 gamma ray source, the results of which should be regarded as only an approximate measure of the radiation damage that would be incurred by an equivalent electron dose

  3. Reference values for total blood volume and cardiac output in humans

    Energy Technology Data Exchange (ETDEWEB)

    Williams, L.R. [Indiana Univ., South Bend, IN (United States). Division of Liberal Arts and Sciences

    1994-09-01

    Much research has been devoted to measurement of total blood volume (TBV) and cardiac output (CO) in humans but not enough effort has been devoted to collection and reduction of results for the purpose of deriving typical or {open_quotes}reference{close_quotes} values. Identification of normal values for TBV and CO is needed not only for clinical evaluations but also for the development of biokinetic models for ultra-short-lived radionuclides used in nuclear medicine (Leggett and Williams 1989). The purpose of this report is to offer reference values for TBV and CO, along with estimates of the associated uncertainties that arise from intra- and inter-subject variation, errors in measurement techniques, and other sources. Reference values are derived for basal supine CO and TBV in reference adult humans, and differences associated with age, sex, body size, body position, exercise, and other circumstances are discussed.

  4. Blood volume, blood pressure and total body sodium: internal signalling and output control

    DEFF Research Database (Denmark)

    Bie, P

    2009-01-01

    Total body sodium and arterial blood pressure (ABP) are mutually dependent variables regulated by complex control systems. This review addresses the role of ABP in the normal control of sodium excretion (NaEx), and the physiological control of renin secretion. NaEx is a pivotal determinant of ABP......, and under experimental conditions, ABP is a powerful, independent controller of NaEx. Blood volume is a function of dietary salt intake; however, ABP is not, at least not in steady states. A transient increase in ABP after a step-up in sodium intake could provide a causal relationship between ABP...... and the regulation of NaEx via a hypothetical integrative control system. However, recent data show that subtle sodium loading (simulating salty meals) causes robust natriuresis without changes in ABP. Changes in ABP are not necessary for natriuresis. Normal sodium excretion is not regulated by pressure. Plasma...

  5. VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 3. Programmer's manual. Final report

    International Nuclear Information System (INIS)

    Stewart, C.W.; Koontz, A.S.; Cuta, J.M.; Montgomery, S.D.

    1983-05-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear-reactor-core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This is Volume 3, the Programmer's Manual. It explains the codes' structures and the computer interfaces

  6. Total reference air kerma can accurately predict isodose surface volumes in cervix cancer brachytherapy. A multicenter study

    DEFF Research Database (Denmark)

    Nkiwane, Karen S; Andersen, Else; Champoudry, Jerome

    2017-01-01

    PURPOSE: To demonstrate that V60 Gy, V75 Gy, and V85 Gy isodose surface volumes can be accurately estimated from total reference air kerma (TRAK) in cervix cancer MRI-guided brachytherapy (BT). METHODS AND MATERIALS: 60 Gy, 75 Gy, and 85 Gy isodose surface volumes levels were obtained from treatm...

  7. A model to estimate volume change due to radiolytic gas bubbles and thermal expansion in solution reactors

    International Nuclear Information System (INIS)

    Souto, F.J.; Heger, A.S.

    2001-01-01

    To investigate the effects of radiolytic gas bubbles and thermal expansion on the steady-state operation of solution reactors at the power level required for the production of medical isotopes, a calculational model has been developed. To validate this model, including its principal hypotheses, specific experiments at the Los Alamos National Laboratory SHEBA uranyl fluoride solution reactor were conducted. The following sections describe radiolytic gas generation in solution reactors, the equations to estimate the fuel solution volume change due to radiolytic gas bubbles and thermal expansion, the experiments conducted at SHEBA, and the comparison of experimental results and model calculations. (author)

  8. Nuclear proliferation and civilian nuclear power: report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle descriptions

    Energy Technology Data Exchange (ETDEWEB)

    1979-12-01

    The Nonproliferation Alternative Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates. Volume IX is divided into three sections: Chapter 1, Reactor Systems; Chapter 2, Fuel-Cycle Systems; and the Appendixes. Chapter 1 contains the characterizations of the following 12 reactor types: light-water reactor; heavy-water reactor; water-cooled breeder reactor; high-temperature gas-cooled reactor; gas-cooled fast reactor; liquid-metal fast breeder reactor; spectral-shift-controlled reactor; accelerator-driven reactor; molten-salt reactor; gaseous-core reactor; tokamak fusion-fisson hybrid reactor; and fast mixed-spectrum reactor. Chapter 2 contains similar information developed for fuel-cycle facilities in the following categories: mining and milling; conversion and enrichment; fuel fabrication; spent fuel reprocessing; waste handling and disposal; and transportation of nuclear materials.

  9. Nuclear proliferation and civilian nuclear power: report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle descriptions

    International Nuclear Information System (INIS)

    1979-12-01

    The Nonproliferation Alternative Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates. Volume IX is divided into three sections: Chapter 1, Reactor Systems; Chapter 2, Fuel-Cycle Systems; and the Appendixes. Chapter 1 contains the characterizations of the following 12 reactor types: light-water reactor; heavy-water reactor; water-cooled breeder reactor; high-temperature gas-cooled reactor; gas-cooled fast reactor; liquid-metal fast breeder reactor; spectral-shift-controlled reactor; accelerator-driven reactor; molten-salt reactor; gaseous-core reactor; tokamak fusion-fisson hybrid reactor; and fast mixed-spectrum reactor. Chapter 2 contains similar information developed for fuel-cycle facilities in the following categories: mining and milling; conversion and enrichment; fuel fabrication; spent fuel reprocessing; waste handling and disposal; and transportation of nuclear materials

  10. Optimal repairable spare-parts procurement policy under total business volume discount environment

    International Nuclear Information System (INIS)

    Pascual, Rodrigo; Santelices, Gabriel; Lüer-Villagra, Armin; Vera, Jorge; Cawley, Alejandro Mac

    2017-01-01

    In asset intensive fields, where components are expensive and high system availability is required, spare parts procurement is often a critical issue. To gain competitiveness and market share is common for vendors to offer Total Business Volume Discounts (TBVD). Accordingly, companies must define the procurement and stocking policy of their spare parts in order to reduce procurement costs and increase asset availability. In response to those needs, this work presents an optimization model that maximizes the availability of the equipment under a TBVD environment, subject to a budget constraint. The model uses a single-echelon structure where parts can be repaired. It determines the optimal number of repairable spare parts to be stocked, giving emphasis on asset availability, procurement costs and service levels as the main decision criteria. A heuristic procedure that achieves high quality solutions in a fast and time-consistent way was implemented to improve the time required to obtain the model solution. Results show that using an optimal procurement policy of spare parts and accounting for TBVD produces better overall results and yields a better availability performance. - Highlights: • We propose a model for procurement of repairable components in single-echelon and business volume discount environments. • We used a mathematical model to develop a competitive heuristic that provides high quality solutions in very short times. • Our model places emphasis on using system availability, procurement costs and service levels as leading decision criteria. • The model can be used as an engine for a multi-criteria Decision Support System.

  11. Experimental determination of the total isothermal reactivity feedback coefficient for the University of Arizona TRIGA research reactor

    International Nuclear Information System (INIS)

    Spriggs, Gregory D.; Nelson, George W.

    1976-01-01

    An experiment was performed to measure the total isothermal (or bath) feedback coefficient of reactivity for the University of Arizona TRIGA Research Reactor (UARR). It was found that the bath coefficient was temperature-dependent and may be represented by the expression α iso .2634 x 10 -2 + .3428 x 10 -3 T - 2.471 x 10 -5 T 2 + 3.476 x 10 -7 T 3 for the temperature range of 7 C to 43 C. (author)

  12. Longitudinal Changes in Total Brain Volume in Schizophrenia: Relation to Symptom Severity, Cognition and Antipsychotic Medication

    NARCIS (Netherlands)

    Veijola, J.; Guo, J.Y.; Moilanen, J.S.; Jaaskelainen, E.; Miettunen, J.; Kyllonen, M.; Haapea, M.; Huhtaniska, S.; Alaraisanen, A.; Maki, P.; Kiviniemi, V.; Nikkinen, J.; Starck, T.; Remes, J.J.; Tanskanen, P.; Tervonen, O.; Wink, A.M.; Kehagia, A.; Suckling, J.; Kobayashi, H.; Barnett, J.H.; Barnes, A.; Koponen, H.J.; Jones, P.B.; Isohanni, M.; Murray, G.K.

    2014-01-01

    Studies show evidence of longitudinal brain volume decreases in schizophrenia. We studied brain volume changes and their relation to symptom severity, level of function, cognition, and antipsychotic medication in participants with schizophrenia and control participants from a general population

  13. Experimental fusion power reactor conceptual design study. Final report. Volume III

    International Nuclear Information System (INIS)

    Baker, C.C.

    1976-12-01

    This document is the final report which describes the work carried out by General Atomic Company for the Electric Power Research Institute on a conceptual design study of a fusion experimental power reactor (EPR) and an overall EPR facility. The primary objective of the two-year program was to develop a conceptual design of an EPR that operates at ignition and produces continuous net power. A conceptual design was developed for a Doublet configuration based on indications that a noncircular tokamak offers the best potential of achieving a sufficiently high effective fuel containment to provide a viable reactor concept at reasonable cost. Other objectives included the development of a planning cost estimate and schedule for the plant and the identification of critical R and D programs required to support the physics development and engineering and construction of the EPR. This volume contains the following appendices: (1) tradeoff code analysis, (2) residual mode transport, (3) blanket/first wall design evaluations, (4) shielding design evaluation, (5) toroidal coil design evaluation, (6) E-coil design evaluation, (7) F-coil design evaluation, (8) plasma recycle system design evaluation, (9) primary coolant purification design evaluation, (10) power supply system design evaluation, (11) number of coolant loops, (12) power conversion system design evaluation, and (13) maintenance methods evaluation

  14. RETRAN code analysis of Tsuruga-2 plant chemical volume control system (CVCS) reactor coolant leakage incident

    International Nuclear Information System (INIS)

    Kawai, Hiroshi

    2002-01-01

    In the Chemical Volume Control System (CVCS) reactor primary coolant leakage incident, which occurred in Tsuruga-2 (4-loop PWR, 3,423 MWt, 1,160 MWe) on July 12, 1999, it took about 14 hours before the leakage isolation. The delayed leakage isolation and a large amount of leakage have become a social concern. Effective procedure modification was studied. Three betterments were proposed based on a qualitative analysis to reduce the pressure and temperature of the primary loop as fast as possible by the current plant facilities while maintaining enough subcooling of the primary loop. I analyzed the incident with RETRAN code in order to quantitatively evaluate the leakage reduction when these betterments are adopted. This paper is very new because it created a typical analysis method for PWR plant behavior during plant shutdown procedure which conventional RETRAN transient analyses rarely dealt with. Also the event time is very long. To carry out this analysis successfully, I devised new models such as an Residual Heat Removal System (RHR) model etc. and simplified parts of the conventional model. Based on the analysis results, I confirmed that leakage can be reduced by about 30% by adopting these betterments. Then the Japan Atomic Power Company (JAPC) modified the operational procedure for reactor primary coolant leakage events adopting these betterments. (author)

  15. Licensing assessment of the CANDU pressurized heavy water reactor. Volume I. Preliminary safety information document

    International Nuclear Information System (INIS)

    1977-06-01

    The PHWR design contains certain features that will require significant modifications to comply with USNRC siting and safety requirements. The most significant of these features are the reactor vessel; control systems; quality assurance program requirements; seismic design of structures, systems and components; and providing an inservice inspection program capability. None of these areas appear insolvable with current state-of-the-art engineering or with upgrading of the quality assurance program for components constructed outside of the USA. In order to be licensed in the U. S., the entire reactor assembly would have to be redesigned to comply with ASME Boiler and Pressure Vessel Code, Section III, Division 1 and Division 2. A summary matrix at the end of this volume identifies compliance of the systems and structures of the PHWR plant with the USNRC General Design Criteria. The matrix further identifies the estimated incremental cost to a 600 MWe PHWR that would be required to license the plant in the U. S. Further, the matrix identifies whether or not the incremental licensing cost is size dependent and the relative percentage of the base direct cost of a Canadian sited plant

  16. Total volume and composition of fluid intake and mortality in older women: a cohort study

    Science.gov (United States)

    Lim, Wai H; Wong, Germaine; Lewis, Joshua R; Lok, Charmaine E; Polkinghorne, Kevan R; Hodgson, Jonathan; Lim, Ee M; Prince, Richard L

    2017-01-01

    Objectives The health benefits of ‘drinking at least 8 glasses of water a day” in healthy individuals are largely unproven. We aimed to examine the relationship between total fluid and the sources of fluid consumption, risk of rapid renal decline, cardiovascular disease (CVD) mortality and all-cause mortality in elderly women. Design, setting and participants We conducted a longitudinal analysis of a population-based cohort study of 1055 women aged ≥70 years residing in Australia. Main outcome measures The associations between total daily fluid intake (defined as total volume of beverage excluding alcohol and milk) and the types of fluid (water, black tea, coffee, milk and other fluids) measured as cups per day and rapid renal decline, CVD and all-cause mortality were assessed using adjusted logistic and Cox regression analyses. Results Over a follow-up period of 10 years, 70 (6.6%) experienced rapid renal decline and 362 (34.4%) died, of which 142 (13.5%) deaths were attributed to CVD. The median (IQR) intake of total fluid was 10.4 (8.5–12.5) cups per day, with water (median (IQR) 4 (2–6) cups per day) and black tea (median (IQR) 3 (1–4) cups per day) being the most frequent type of fluid consumed. Every cup per day higher intake of black tea was associated with adjusted HRs of 0.90 (95% CI 0.81 to 0.99) and 0.92 (95% CI 0.86 to 0.98) for CVD mortality and all-cause mortality, respectively. There were no associations between black tea intake and rapid renal decline, or between the quantity or type of other fluids, including water intake, and any clinical outcomes. Conclusions Habitual higher intake of black tea may potentially improve long-term health outcomes, independent of treating traditional CVD risk factors, but validation of our study findings is essential. PMID:28341683

  17. Total quality management for addressing suspect parts at the Oak Ridge High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Hendrix, K.A.; Tulay, M.P.

    1993-01-01

    Martin Marietta Energy System (MMES) Research Reactors Division (RRD), operator of the High Flux Isotope Reactor (HFIR) recently embarked on an aggressive Program to address the issue of suspect Parts and to enhance their procurement process. Through the application of TQM process improvement, RRD has already achieved improved efficiency in specifying, procuring, and accepting replacement items for its largest research reactor. These process improvements have significantly decreased the risk of installing suspect parts in the HFIR safety systems. To date, a systematic plan has been implemented, which includes the following elements: Process assessment and procedure review; Procedural enhancements; On-site training and technology transfer; Enhanced receiving inspections; Performance supplier evaluations and source verifications integrated processes for utilizing commercial grade products in nuclear safety-related applications. This paper will describe the above elements, how a partnership between MMES and Gilbert/Commonwealth facilitated the execution of the plan, and how process enhancements were applied. We will also present measures for improved efficiency and productivity, that MMES intends to continually address with Quality Action Teams

  18. Electrical Capacitance Volume Tomography for the Packed Bed Reactor ISS Flight Experiment

    Science.gov (United States)

    Marashdeh, Qussai; Motil, Brian; Wang, Aining; Liang-Shih, Fan

    2013-01-01

    Fixed packed bed reactors are compact, require minimum power and maintenance to operate, and are highly reliable. These features make this technology a highly desirable unit operation for long duration life support systems in space. NASA is developing an ISS experiment to address this technology with particular focus on water reclamation and air revitalization. Earlier research and development efforts funded by NASA have resulted in two hydrodynamic models which require validation with appropriate instrumentation in an extended microgravity environment. To validate these models, the instantaneous distribution of the gas and liquid phases must be measured.Electrical Capacitance Volume Tomography (ECVT) is a non-invasive imaging technology recently developed for multi-phase flow applications. It is based on distributing flexible capacitance plates on the peripheral of a flow column and collecting real-time measurements of inter-electrode capacitances. Capacitance measurements here are directly related to dielectric constant distribution, a physical property that is also related to material distribution in the imaging domain. Reconstruction algorithms are employed to map volume images of dielectric distribution in the imaging domain, which is in turn related to phase distribution. ECVT is suitable for imaging interacting materials of different dielectric constants, typical in multi-phase flow systems. ECVT is being used extensively for measuring flow variables in various gas-liquid and gas-solid flow systems. Recent application of ECVT include flows in risers and exit regions of circulating fluidized beds, gas-liquid and gas-solid bubble columns, trickle beds, and slurry bubble columns. ECVT is also used to validate flow models and CFD simulations. The technology is uniquely qualified for imaging phase concentrations in packed bed reactors for the ISS flight experiments as it exhibits favorable features of compact size, low profile sensors, high imaging speed, and

  19. The use of phase sequence image sets to reconstruct the total volume occupied by a mobile lung tumor

    International Nuclear Information System (INIS)

    Gagne, Isabelle M.; Robinson, Don M.; Halperin, Ross; Roa, Wilson

    2005-01-01

    The use of phase sequence image (PSI) sets to reveal the total volume occupied by a mobile target is presented. Isocontrast composite clinical target volumes (CCTVs) may be constructed from PSI sets in order to reveal the total volume occupied by a mobile target during the course of its travel. The ability of the CCTV technique to properly account for target motion is demonstrated by comparison to contours of the true total volume occupied (TVO) for a number of experimental phantom geometries. Finally, using real patient data, the clinical utility of the CCTV technique to properly account for internal tumor motion while minimizing the volume of healthy lung tissue irradiated is assessed by comparison to the standard approach of applying safety margins. Results of the phantom study reveal that CCTV cross sections constructed at the 20% isocontrast level yield good agreement with the total cross sections (TXO) of mobile targets. These CCTVs conform well to the TVOs of the moving targets examined whereby the addition of small uniform margins ensures complete circumscription of the TVO with the inclusion of minimal amounts of surrounding external volumes. The CCTV technique is seen to be clearly superior to the common practice of the addition of safety margins to individual CTV contours in order to account for internal target motion. Margins required with the CCTV technique are eight to ten times smaller than those required with individual CTVs

  20. Two neural network based strategies for the detection of a total instantaneous blockage of a sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Martinez-Martinez, Sinuhe; Messai, Nadhir; Jeannot, Jean-Philippe; Nuzillard, Danielle

    2015-01-01

    The total instantaneous blockage (TIB) of an assembly in the core of a sodium-cooled fast reactor (SFR) is investigated. Such incident could appear as an abnormal rise in temperature on the assemblies neighbouring the blockage. Its detection relies on a dataset of temperature measurements of the assemblies making up the core of the French Phenix Nuclear Reactor. The data are provided by the French Commission of Atomic and Alternatives Energies (CEA). Here, two strategies are proposed depending on whether the sensor measurement of the suspected assembly is reliable or not. The proposed methodology implements a time-lagged feed-forward neural (TLFFN) Network in order to predict the one-step-ahead temperature of a given assembly. The incident is declared if the difference between the predicted process and the actual one exceeds a threshold. In these simulated conditions, the method is efficient to detect small gradients as expected in reality. - Highlights: • We study the total instantaneous blockage (TIB) of a sodium-cooled fast reactor. • The TIB symptom is simulated as an abrupt rise on temperature (0.1–1 °C/s). • The goal is to improve the early detection of the incident. • Two strategies laying on neural networks are proposed. • TIB is detected in 3 s for 1 °C/s and 18–21 s for 0.1 °C/s

  1. Gas cooled reactor assessment. Volume II. Final report, February 9, 1976--June 30, 1976

    International Nuclear Information System (INIS)

    1976-08-01

    This report was prepared to document the estimated power plant capital and operating costs, and the safety and environmental assessments used in support of the Gas Cooled Reactor Assessment performed by Arthur D. Little, Inc. (ADL), for the U.S. Energy Research and Development Administration. The gas-cooled reactor technologies investigated include: the High Temperature Gas Reactor Steam Cycle (HTGR-SC), the HTGR Direct Cycle (HTGR-DC), the Very High Temperature Reactor (VHTR) and the Gas Cooled Fast Reactor (GCFR). Reference technologies used for comparison include: Light Water Reactors (LWR), the Liquid Metal Fast Breeder Reactor (LMFBR), conventional coal-fired steam plants, and coal combustion for process heat

  2. Novel regenerative large-volume immobilized enzyme reactor: preparation, characterization and application.

    Science.gov (United States)

    Ruan, Guihua; Wei, Meiping; Chen, Zhengyi; Su, Rihui; Du, Fuyou; Zheng, Yanjie

    2014-09-15

    A novel large-volume immobilized enzyme reactor (IMER) on small column was prepared with organic-inorganic hybrid silica particles and applied for fast (10 min) and oriented digestion of protein. At first, a thin enzyme support layer was formed in the bottom of the small column by polymerization with α-methacrylic acid and dimethacrylate. After that, amino SiO2 particles was prepared by the sol-gel method with tetraethoxysilane and 3-aminopropyltriethoxysilane. Subsequently, the amino SiO2 particles were activated by glutaraldehyde for covalent immobilization of trypsin. Digestive capability of large-volume IMER for proteins was investigated by using bovine serum albumin (BSA), cytochrome c (Cyt-c) as model proteins. Results showed that although the sequence coverage of the BSA (20%) and Cyt-c (19%) was low, the large-volume IMER could produce peptides with stable specific sequence at 101-105, 156-160, 205-209, 212-218, 229-232, 257-263 and 473-451 of the amino sequence of BSA when digesting 1mg/mL BSA. Eight of common peptides were observed during each of the ten runs of large-volume IMER. Besides, the IMER could be easily regenerated by reactivating with GA and cross-linking with trypsin after breaking the -C=N- bond by 0.01 M HCl. The sequence coverage of BSA from regenerated IMER increased to 25% comparing the non-regenerated IMER (17%). 14 common peptides. accounting for 87.5% of first use of IMER, were produced both with IMER and regenerated IMER. When the IMER was applied for ginkgo albumin digestion, the sequence coverage of two main proteins of ginkgo, ginnacin and legumin, was 56% and 55%, respectively. (Reviewer 2) Above all, the fast and selective digestion property of the large-volume IMER indicated that the regenerative IMER could be tentatively used for the production of potential bioactive peptides and the study of oriented protein digestion. Copyright © 2014 Elsevier B.V. All rights reserved.

  3. Analysis of a total loss of pool water accident in MTR-type research reactors

    International Nuclear Information System (INIS)

    Yilmazer, A.; Yavuz, H.

    2004-01-01

    In this study, the transient in which the pool water is lost throughout one or more of the main coolant pipes which are supposed to be broken guillotine-like is investigated for the TR-2 research reactor in Istanbul. The applicability of the methods used for other similar types of research reactors is shown. Decrease of the pool water level until the top of the core, and from the top to the bottom of the core are examined as two successive phases of the accident. Finite difference scheme and integral methods are employed to solve energy equations and the results of both methods are compared. The finite difference solution uses an explicit form for the analysis of the first phase, and a moving boundary approach for the second phase. The integral method is based on the assumption that the temperatures appearing in the energy equations have the same profiles during the transient as the steady state ones. Analyses are done both for nominal and hot channel, and the results of both methods are observed to be in agreement. (orig.)

  4. Analysis of a total loss of pool water accident in MTR-type research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yilmazer, A. [Hacettepe University, Ankara (Turkey). Nuclear Engineering Department; Yavuz, H. [Istanbul Technical University (Turkey). Energy Institute

    2004-08-01

    In this study, the transient in which the pool water is lost throughout one or more of the main coolant pipes which are supposed to be broken guillotine-like is investigated for the TR-2 research reactor in Istanbul. The applicability of the methods used for other similar types of research reactors is shown. Decrease of the pool water level until the top of the core, and from the top to the bottom of the core are examined as two successive phases of the accident. Finite difference scheme and integral methods are employed to solve energy equations and the results of both methods are compared. The finite difference solution uses an explicit form for the analysis of the first phase, and a moving boundary approach for the second phase. The integral method is based on the assumption that the temperatures appearing in the energy equations have the same profiles during the transient as the steady state ones. Analyses are done both for nominal and hot channel, and the results of both methods are observed to be in agreement. (orig.)

  5. TOTAL WOOD VOLUME ESTIMATION OF EUCALYPTUS SPECIES BY IMAGES OF LANDSAT SATELLITE

    Directory of Open Access Journals (Sweden)

    Elias Fernando Berra

    2012-12-01

    Full Text Available http://dx.doi.org/10.5902/198050987566Models relating spectral answers with biophysical parameters aim estimate variables, like wood volume, without the necessity of frequent field measurements. The objective was to develop models to estimate wood volume by Landsat 5 TM images, supported by regional forest inventory data. The image was geo-referenced and converted to spectral reflectance. After, the images-index NDVI (Normalized Difference Vegetation Index and SR (Simple Ratio was generated. The reflectance values of the bands (TM1, TM2, TM3 e TM4 and of the indices (NDVI and SR was related with the wood volume. The biggest correlation with volume was with the NDVI and SR indices. The variables selection was made by Stepwise method, which returned three regression models as significant to explain the variation in volume. Finally, the best fitted model was selected (volume = -830,95 + 46,05 (SR + 107,47 (TM2, which was applied on the Landsat image where the pixels had started to represent the estimated volume in m³/ha on the Eucalyptus sp. production units. This model, significant at 95% confidence level, explains 68% of the wood volume variation.

  6. Effect of total cementitious content on shear strength of high-volume fly ash concrete beams

    International Nuclear Information System (INIS)

    Arezoumandi, Mahdi; Volz, Jeffery S.; Ortega, Carlos A.; Myers, John J.

    2013-01-01

    Highlights: ► Existing design standards conservatively predicted the capacity of the HVFAC beams. ► In general, the HVFAC beams exceeded the code predicted shear strengths. ► The cementitious content did not have effect on the shear behavior of the HVFAC beams. - Abstract: The production of portland cement – the key ingredient in concrete – generates a significant amount of carbon dioxide. However, due to its incredible versatility, availability, and relatively low cost, concrete is the most consumed manmade material on the planet. One method of reducing concrete’s contribution to greenhouse gas emissions is the use of fly ash to replace a significant amount of the cement. This paper compares two experimental studies that were conducted to investigate the shear strength of full-scale beams constructed with high-volume fly ash concrete (HVFAC) – concrete with at least 50% of the cement replaced with fly ash. The primary difference between the two studies involved the amount of cementitious material, with one mix having a relatively high total cementitious content (502 kg/m 3 ) and the other mix having a relatively low total cementitious content (337 kg/m 3 ). Both mixes utilized a 70% replacement of portland cement with a Class C fly ash. Each of these experimental programs consisted of eight beams (six without shear reinforcing and two with shear reinforcing in the form of stirrups) with three different longitudinal reinforcement ratios. The beams were tested under a simply supported four-point loading condition. The experimental shear strengths of the beams were compared with both the shear provisions of selected standards (US, Australia, Canada, Europe, and Japan) and a shear database of conventional concrete (CC) specimens. Furthermore, statistical data analyses (both parametric and nonparametric) were performed to evaluate whether or not there is any statistically significant difference between the shear strength of both mixes. Results of these

  7. Effects of tidal volume and methacholine on low-frequency total respiratory impedance in dogs.

    Science.gov (United States)

    Lutchen, K R; Jackson, A C

    1990-05-01

    The frequency dependence of respiratory impedance (Zrs) from 0.125 to 4 Hz (Hantos et al., J. Appl. Physiol. 60: 123-132, 1986) may reflect inhomogeneous parallel time constants or the inherent viscoelastic properties of the respiratory tissues. However, studies on the lung alone or chest wall alone indicate that their impedance features are also dependent on the tidal volumes (VT) of the forced oscillations. The goals of this study were 1) to identify how total Zrs at lower frequencies measured with random noise (RN) compared with that measure with larger VT, 2) to identify how Zrs measured with RN is affected by bronchoconstriction, and 3) to identify the impact of using linear models for analyzing such data. We measured Zrs in six healthy dogs by use of a RN technique from 0.125 to 4 Hz or with a ventilator from 0.125 to 0.75 Hz with VT from 50 to 250 ml. Then methacholine was administered and the RN was repeated. Two linear models were fit to each separate set of data. Both models assume uniform airways leading to viscoelastic tissues. For healthy dogs, the respiratory resistance (Rrs) decreased with frequency, with most of the decrease occurring from 0.125 to 0.375 Hz. Significant VT dependence of Rrs was seen only at these lower frequencies, with Rrs higher as VT decreased. The respiratory compliance (Crs) was dependent on VT in a similar fashion at all frequencies, with Crs decreasing as VT decreased. Both linear models fit the data well at all VT, but the viscoelastic parameters of each model were very sensitive to VT. After methacholine, the minimum Rrs increased as did the total drop with frequency. Nevertheless the same models fit the data well, and both the airways and tissue parameters were altered after methacholine. We conclude that inferences based only on low-frequency Zrs data are problematic because of the effects of VT on such data (and subsequent linear modeling of it) and the apparent inability of such data to differentiate parallel

  8. Equal Pay for Equal Work: Medicare Procedure Volume and Reimbursement for Male and Female Surgeons Performing Total Knee and Total Hip Arthroplasty.

    Science.gov (United States)

    Holliday, Emma B; Brady, Christina; Pipkin, William C; Somerson, Jeremy S

    2018-02-21

    The observed sex gap in physician salary has been the topic of much recent debate in the United States, but it has not been well-described among orthopaedic surgeons. The objective of this study was to evaluate for sex differences in Medicare claim volume and reimbursement among orthopaedic surgeons. The Medicare Provider Utilization and Payment Public Use File was used to compare claim volume and reimbursement between female and male orthopaedic surgeons in 2013. Data were extracted for each billing code per orthopaedic surgeon in the year 2013 for total claims, surgical claims, total knee arthroplasty (TKA) claims, and total hip arthroplasty (THA) claims. A total of 20,546 orthopaedic surgeons who treated traditional Medicare patients were included in the initial analysis. Claim volume and reimbursement received were approximately twofold higher for all claims and more than threefold higher for surgical claims for male surgeons when compared with female surgeons (p 10 TKAs and THAs, respectively, in 2013 for Medicare patients and were included in the subset analyses. Although male surgeons performed a higher mean number of TKAs than female surgeons (mean and standard deviation, 37 ± 33 compared with 26 ± 17, respectively, p men and women for TKA or THA ($1,135 ± $228 compared with $1,137 ± $184 for TKA, respectively, p = 0.380; $1,049 ± $226 compared with $1,043 ± $266 for THA, respectively, p = 0.310). Female surgeons had a lower number of total claims and reimbursements compared with male surgeons. However, among surgeons who performed >10 THAs and TKAs, there were no sex differences in the mean reimbursement payment per surgeon. The number of women in orthopaedics is rising, and there is much interest in how their productivity and compensation compare with their male counterparts.

  9. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1997-02-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: reactor pressure vessel embrittlement and thermal annealing; reactor vessel lower head integrity; and evaluation and projection of steam generator tube condition and integrity. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  10. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity

    International Nuclear Information System (INIS)

    Monteleone, S.

    1997-02-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: reactor pressure vessel embrittlement and thermal annealing; reactor vessel lower head integrity; and evaluation and projection of steam generator tube condition and integrity. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  11. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  12. Longitudinal changes in total brain volume in schizophrenia: relation to symptom severity, cognition and antipsychotic medication.

    Directory of Open Access Journals (Sweden)

    Juha Veijola

    Full Text Available Studies show evidence of longitudinal brain volume decreases in schizophrenia. We studied brain volume changes and their relation to symptom severity, level of function, cognition, and antipsychotic medication in participants with schizophrenia and control participants from a general population based birth cohort sample in a relatively long follow-up period of almost a decade. All members of the Northern Finland Birth Cohort 1966 with any psychotic disorder and a random sample not having psychosis were invited for a MRI brain scan, and clinical and cognitive assessment during 1999-2001 at the age of 33-35 years. A follow-up was conducted 9 years later during 2008-2010. Brain scans at both time points were obtained from 33 participants with schizophrenia and 71 control participants. Regression models were used to examine whether brain volume changes predicted clinical and cognitive changes over time, and whether antipsychotic medication predicted brain volume changes. The mean annual whole brain volume reduction was 0.69% in schizophrenia, and 0.49% in controls (p = 0.003, adjusted for gender, educational level, alcohol use and weight gain. The brain volume reduction in schizophrenia patients was found especially in the temporal lobe and periventricular area. Symptom severity, functioning level, and decline in cognition were not associated with brain volume reduction in schizophrenia. The amount of antipsychotic medication (dose years of equivalent to 100 mg daily chlorpromazine over the follow-up period predicted brain volume loss (p = 0.003 adjusted for symptom level, alcohol use and weight gain. In this population based sample, brain volume reduction continues in schizophrenia patients after the onset of illness, and antipsychotic medications may contribute to these reductions.

  13. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  14. Scale-4 analysis of pressurized water reactor critical configurations: Volume 5, North Anna Unit 1 Cycle 5

    International Nuclear Information System (INIS)

    Bowman, S.M.; Suto, T.

    1996-10-01

    ANSI/ANS 8.1 requires that calculational methods for away-from- reactor (AFR) criticality safety analyses be validated against experiment. This report summarizes part of the ongoing effort to benchmark AFR criticality analysis methods using selected critical configurations from commercial PWRs. Codes and data in the SCALE-4 code system were used. This volume documents the SCALE system analysis of one reactor critical configuration for North Anna Unit 1 Cycle 5. The KENO V.a criticality calculations for the North Anna 1 Cycle 5 beginning-of-cycle model yielded a value for k eff of 1. 0040±0.0005

  15. Office for Analysis and Evaluation of Operational Data 1996 annual report. Volume 10, Number 1: Reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    This annual report of the US Nuclear Regulatory Commission`s Office for Analysis and Evaluation of Operational Data (AEOD) describes activities conducted during 1996. The report is published in three parts. NUREG-1272, Vol. 10, No. 1, covers power reactors and presents an overview of the operating experience of the nuclear power industry from the NRC perspective, including comments about trends of some key performance measures. The report also includes the principal findings and issues identified in AEOD studies over the past year and summarizes information from such sources as licensee event reports and reports to the NRC`s Operations Center. NUREG-1272, Vol. 10, No. 2, covers nuclear materials and presents a review of the events and concerns during 1996 associated with the use of licensed material in nonreactor applications, such as personnel overexposures and medical misadministrations. Both reports also contain a discussion of the Incident Investigation Team program and summarize both the Incident Investigation Team and Augmented Inspection Team reports. Each volume contains a list of the AEOD reports issued from CY 1980 through 1996. NUREG-1272, Vol. 10, No. 3, covers technical training and presents the activities of the Technical Training Center in support of the NRC`s mission in 1996.

  16. Reactor Emergency Action Level Monitor: Volume 2, REALM user's reference guide: Final report

    International Nuclear Information System (INIS)

    Touchton, R.A.

    1988-09-01

    A User Manual for the Reactor Emergency Action Level Monitor (REALM) expert system prototype is provided in this volume. REALM has been designed to provide expert assistance in the identification of a nuclear power plant emergency situation and the determination of its severity. REALM has been developed as an expert system which can provide sensor interpretation and situation assessment in a real-time processing environment. In its state of development at project completion, these capabilities are used in an off-line (i.e., stand-alone, desktop) fashion to provide emergency preparedness assistance in the areas of emergency classification training and emergency exercise scenario generation. REALM also serves a prototype and stepping-stone for the possible connection to the plant for on-line use. In order to distinguish the off-line system (now complete) from the on-line system (now moving from a research prototype to an installed system), the term ''REALM'' is used to indicate the on-line version, with users in the control room, technical support center, and the emergency operations facility, The off-line version is referred to as ''uREALM.''

  17. Office for Analysis and Evaluation of Operational Data 1996 annual report. Volume 10, Number 1: Reactors

    International Nuclear Information System (INIS)

    1997-12-01

    This annual report of the US Nuclear Regulatory Commission's Office for Analysis and Evaluation of Operational Data (AEOD) describes activities conducted during 1996. The report is published in three parts. NUREG-1272, Vol. 10, No. 1, covers power reactors and presents an overview of the operating experience of the nuclear power industry from the NRC perspective, including comments about trends of some key performance measures. The report also includes the principal findings and issues identified in AEOD studies over the past year and summarizes information from such sources as licensee event reports and reports to the NRC's Operations Center. NUREG-1272, Vol. 10, No. 2, covers nuclear materials and presents a review of the events and concerns during 1996 associated with the use of licensed material in nonreactor applications, such as personnel overexposures and medical misadministrations. Both reports also contain a discussion of the Incident Investigation Team program and summarize both the Incident Investigation Team and Augmented Inspection Team reports. Each volume contains a list of the AEOD reports issued from CY 1980 through 1996. NUREG-1272, Vol. 10, No. 3, covers technical training and presents the activities of the Technical Training Center in support of the NRC's mission in 1996

  18. Functional Response of Tumor Vasculature to PaCO2: Determination of Total and Microvascular Blood Volume by MRI

    Directory of Open Access Journals (Sweden)

    Scott D. Packard

    2003-07-01

    Full Text Available In order to identify differences in functional activity, we compared the reactivity of glioma vasculature and the native cerebral vasculature to both dilate and constrict in response to altered PaCO2. Gliomas were generated by unilateral implantation of U87MGdEGFR human glioma tumor cells into the striatum of adult female athymic rats. Relative changes in total and microvascular cerebral blood volume were determined by steady state contrast agent-enhanced magnetic resonance imaging for transitions from normocarbia to hypercarbia and hypocarbia. Although hypercarbia induced a significant increase in both total and microvascular blood volume in normal brain and glioma, reactivity of glioma vasculature was significantly blunted in comparison to normal striatum; glioma total CBV increased by 0.6±0.1%/mm Hg CO2 whereas normal striatum increased by 1.5±0.2%/mm Hg CO2, (P < .0001, group t-test. Reactivity of microvascular blood volume was also significantly blunted. In contrast, hypocarbia decreased both total and microvascular blood volumes more in glioma than in normal striatum. These results indicate that cerebral blood vessels derived by tumor-directed angiogenesis do retain reactivity to CO2. Furthermore, reduced reactivity of tumor vessels to a single physiological perturbation, such as hypercarbia, should not be construed as a generalized reduction of functional activity of the tumor vascular bed.

  19. Draft environmental impact statement siting, construction, and operation of New Production Reactor capacity. Volume 4, Appendices D-R

    Energy Technology Data Exchange (ETDEWEB)

    None

    1991-04-01

    This Environmental Impact Statement (EIS) assesses the potential environmental impacts, both on a broad programmatic level and on a project-specific level, concerning a proposed action to provide new tritium production capacity to meet the nation`s nuclear defense requirements well into the 21st century. A capacity equivalent to that of about a 3,000-megawatt (thermal) heavy-water reactor was assumed as a reference basis for analysis in this EIS; this is the approximate capacity of the existing production reactors at DOE`s Savannah River Site near Aiken, South Carolina. The EIS programmatic alternatives address Departmental decisions to be made on whether to build new production facilities, whether to build one or more complexes, what size production capacity to provide, and when to provide this capacity. Project-specific impacts for siting, constructing, and operating new production reactor capacity are assessed for three alternative sites: the Hanford Site near Richland, Washington; the Idaho National Engineering Laboratory near Idaho Falls, Idaho; and the Savannah River Site. For each site, the impacts of three reactor technologies (and supporting facilities) are assessed: a heavy-water reactor, a light-water reactor, and a modular high-temperature gas-cooled reactor. Impacts of the no-action alternative also are assessed. The EIS evaluates impacts related to air quality; noise levels; surface water, groundwater, and wetlands; land use; recreation; visual environment; biotic resources; historical, archaeological, and cultural resources; socioeconomics; transportation; waste management; and human health and safety. The EIS describes in detail the potential radioactive releases from new production reactors and support facilities and assesses the potential doses to workers and the general public. This volume contains 15 appendices.

  20. Safety Research Experiment Facility Project. Conceptual design report. Volume VII. Reactor cooling

    International Nuclear Information System (INIS)

    1975-12-01

    The Reactor Cooling System (RCS) will provide the required cooling during test operations of the Safety Research Experiment Facility (SAREF) reactor. The RCS transfers the reactor energy generated in the core to a closed-loop water storage system located completely inside the reactor containment building. After the reactor core has cooled to a safe level, the stored heat is rejected through intermediate heat exchangers to a common forced-draft evaporative cooling tower. The RCS is comprised of three independent cooling loops of which any two can remove sufficient heat from the core to prevent structural damage to the system components

  1. Compilation of reports of the Advisory Committee on Reactor Safeguards, 1957-1984. Volume 5. Generic Subjects H-R

    International Nuclear Information System (INIS)

    1985-04-01

    This six-volume compilation contains over 1000 reports prepared by the Advisory Committee on Reactor Safeguards from September 1957 through December 1984. The reports are divided into two groups: Part 1: ACRS Reports on Project Reviews, and Part 2: ACRS Reports on Generic Subjects. Part 1 contains ACRS reports alphabetized by project name and within project name by chronological order. Part 2 categorizes the reports by the most appropriate generic subject area and within subject area by chronological order. This volume presents generic subjects arranged alphabetically from H to R

  2. Compilation of reports of the Advisory Committee on Reactor Safeguards, 1957-1984. Volume 4. Generic Subjects A-G

    International Nuclear Information System (INIS)

    1985-04-01

    This six-volume compilation contains over 1000 reports prepared by the Advisory Committee on Reactor Safeguards from September 1957 through December 1984. The reports are divided into two groups: Part 1: ACRS Reports on Project Reviews, and Part 2: ACRS Reports on Generic Subjects. Part 1 contains ACRS reports alphabetized by subject name and within project name by chronological order. Part 2 categorizes the reports by the most appropriate generic subject area and within subject area by chronological order. This volume contains generic reports arranged alphabetically from A to G

  3. Compilation of reports of the Advisory Committee on Reactor Safeguards, 1957-1984. Volume 1. Project Reviews A-F

    International Nuclear Information System (INIS)

    1985-04-01

    This six-volume compilation contains over 1000 reports prepared by the Advisory Committee on Reactor Safeguards from September 1957 through December 1984. The reports are divided into two groups: Part 1: ACRS Reports on Project Reviews, and Part 2: ACRS Reports on Generic Subjects. Part 1 contains ACRS reports alphabetized by project name and within project name by chronological order. Part 2 categorizes the reports by the most appropriate generic subject area and within subject area by chronological order. This volume contains project reviews arranged alphabetically from A to F

  4. Compilation of reports of the Advisory Committee on Reactor Safeguards, 1957-1984. Volume 3. Project Reviews Q-Z

    International Nuclear Information System (INIS)

    1985-04-01

    This six-volume compilation contains over 1000 reports prepared by the Advisory Committee on Reactor Safeguards from September 1957 through December 1984. The reports are divided into two groups: Part 1: ACRS Resports on Project Reviews, and Part 2: ACRS Reports on Generic Subjects. Part 1 contains ACRS reports alphabetized by project name and within project name by chronological order. Part 2 categorizes the reports by the most appropriate generic subject area and within subject area by chronological order. This volume includes reports arranged alpbabetically from Q to Z

  5. Twenty-third water reactor safety information meeting: Volume 1, plenary session, high burnup fuel behavior, thermal hydraulic research. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1996-03-01

    This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 1, present topics on High Burnup Fuel Behavior, Thermal Hydraulic Research, and Plenary Session topics. Individual papers have been cataloged separately.

  6. Twenty-third water reactor safety information meeting: Volume 1, plenary session, high burnup fuel behavior, thermal hydraulic research. Proceedings

    International Nuclear Information System (INIS)

    Monteleone, S.

    1996-03-01

    This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 1, present topics on High Burnup Fuel Behavior, Thermal Hydraulic Research, and Plenary Session topics. Individual papers have been cataloged separately

  7. Plasma properties in a large-volume, cylindrical and asymmetric radio-frequency capacitively coupled industrial-prototype reactor

    International Nuclear Information System (INIS)

    Lazović, Saša; Puač, Nevena; Spasić, Kosta; Malović, Gordana; Petrović, Zoran Lj; Cvelbar, Uroš; Mozetič, Miran; Radetić, Maja

    2013-01-01

    We have developed a large-volume low-pressure cylindrical plasma reactor with a size that matches industrial reactors for treatment of textiles. It was shown that it efficiently produces plasmas with only a small increase in power as compared with a similar reactor with 50 times smaller volume. Plasma generated at 13.56 MHz was stable from transition to streamers and capable of long-term continuous operation. An industrial-scale asymmetric cylindrical reactor of simple design and construction enabled good control over a wide range of active plasma species and ion concentrations. Detailed characterization of the discharge was performed using derivative, Langmuir and catalytic probes which enabled determination of the optimal sets of plasma parameters necessary for successful industry implementation and process control. Since neutral atomic oxygen plays a major role in many of the material processing applications, its spatial profile was measured using nickel catalytic probe over a wide range of plasma parameters. The spatial profiles show diffusion profiles with particle production close to the powered electrode and significant wall losses due to surface recombination. Oxygen atom densities range from 10 19 m −3 near the powered electrode to 10 17 m −3 near the wall. The concentrations of ions at the same time are changing from 10 16 to the 10 15 m −3 at the grounded chamber wall. (paper)

  8. Compilation of reports of the Advisory Committee on Reactor Safeguards, 1957-1984. Volume 2. Project Reviews G-P

    International Nuclear Information System (INIS)

    1985-04-01

    This six-volume compilation contains over 1000 reports prepared by the Advisory Committee on Reactor Safeguards from September 1957 through December 1984. The reports are divided into two groups: Part 1: ACRS Reports on Project Reviews, and Part 2: ACRS Reports on Generic Subjects. Part 1 contains ACRS reports alphabetized by project name and within project name by chronological order. Part 2 categorizes the reports by the most appropriate generic subject area and within subject area by chronological order

  9. Final Generic Environmental Impact Statement. Handling and storage of spent light water power reactor fuel. Volume 2. Appendices

    International Nuclear Information System (INIS)

    1979-08-01

    This volume contains the following appendices: LWR fuel cycle, handling and storage of spent fuel, termination case considerations (use of coal-fired power plants to replace nuclear plants), increasing fuel storage capacity, spent fuel transshipment, spent fuel generation and storage data, characteristics of nuclear fuel, away-from-reactor storage concept, spent fuel storage requirements for higher projected nuclear generating capacity, and physical protection requirements and hypothetical sabotage events in a spent fuel storage facility

  10. Archaea and Bacteria Acclimate to High Total Ammonia in a Methanogenic Reactor Treating Swine Waste

    Directory of Open Access Journals (Sweden)

    Sofia Esquivel-Elizondo

    2016-01-01

    Full Text Available Inhibition by ammonium at concentrations above 1000 mgN/L is known to harm the methanogenesis phase of anaerobic digestion. We anaerobically digested swine waste and achieved steady state COD-removal efficiency of around 52% with no fatty-acid or H2 accumulation. As the anaerobic microbial community adapted to the gradual increase of total ammonia-N (NH3-N from 890±295 to 2040±30 mg/L, the Bacterial and Archaeal communities became less diverse. Phylotypes most closely related to hydrogenotrophic Methanoculleus (36.4% and Methanobrevibacter (11.6%, along with acetoclastic Methanosaeta (29.3%, became the most abundant Archaeal sequences during acclimation. This was accompanied by a sharp increase in the relative abundances of phylotypes most closely related to acetogens and fatty-acid producers (Clostridium, Coprococcus, and Sphaerochaeta and syntrophic fatty-acid Bacteria (Syntrophomonas, Clostridium, Clostridiaceae species, and Cloacamonaceae species that have metabolic capabilities for butyrate and propionate fermentation, as well as for reverse acetogenesis. Our results provide evidence countering a prevailing theory that acetoclastic methanogens are selectively inhibited when the total ammonia-N concentration is greater than ~1000 mgN/L. Instead, acetoclastic and hydrogenotrophic methanogens coexisted in the presence of total ammonia-N of ~2000 mgN/L by establishing syntrophic relationships with fatty-acid fermenters, as well as homoacetogens able to carry out forward and reverse acetogenesis.

  11. Total petroleum hydrocarbon degradation by hybrid electrobiochemical reactor in oilfield produced water

    International Nuclear Information System (INIS)

    Mousa, Ibrahim E.

    2016-01-01

    The crude oil drilling and extraction operations are aimed to maximize the production may be counterbalanced by the huge production of contaminated produced water (PW). PW is conventionally treated through different physical, chemical, and biological technologies. The efficiency of suggested hybrid electrobiochemical (EBC) methods for the simultaneous removal of total petroleum hydrocarbon (TPH) and sulfate from PW generated by petroleum industry is studied. Also, the factors that affect the stability of PW quality are investigated. The results indicated that the effect of biological treatment is very important to keep control of the electrochemical by-products and more TPH removal in the EBC system. The maximum TPH and sulfate removal efficiency was achieved 75% and 25.3%, respectively when the detention time was about 5.1 min and the energy consumption was 32.6 mA/cm 2 . However, a slight increasing in total bacterial count was observed when the EBC compact unit worked at a flow rate of average 20 L/h. Pseudo steady state was achieved after 30 min of current application in the solution. Also, the results of the study indicate that when the current intensity was increased above optimum level, no significant results occurred due to the release of gases. - Highlights: • The hybrid electrolytic biological cell was used for degradation of oilfield produced water. • Decomposition of Total Petroleum Hydrocarbon with or without the biofilter. • High saline water with the high chloride and sulfate ions content treatment. • The removal of electrochemical by-products is a phase change technique that requires the maintenance the biofilm on the filter media, which is sensitive and a complex operation. • Biofilter is efficient for the degradation of PW bye products, the critical drawback to their utility in full-scale operations is high TDS water content and detention time of treatment.

  12. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

    After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

  13. Safety Research Experiment Facility Project. Conceptual design report. Volume IV. Reactor containment

    International Nuclear Information System (INIS)

    1975-12-01

    The principal purpose of the SAREF Reactor Containment Building (RCB) is to prevent the uncontrolled release of radioactive materials to the atmosphere as a result of accidental occurrences inside the containment. The RCB houses numerous reactor systems and components including the Prestressed Concrete Reactor Vessel (PCRV). The design of the RCB is of reinforced concrete (steel-lined). The containment building is embedded nearly 100 feet in lava rock. It has therefore been necessary to independently formulate an appropriate and conservative design approach

  14. Risk factors for radiation pneumonitis after stereotactic radiation therapy for lung tumours: clinical usefulness of the planning target volume to total lung volume ratio.

    Science.gov (United States)

    Ueyama, Tomoko; Arimura, Takeshi; Takumi, Koji; Nakamura, Fumihiko; Higashi, Ryutaro; Ito, Soichiro; Fukukura, Yoshihiko; Umanodan, Tomokazu; Nakajo, Masanori; Koriyama, Chihaya; Yoshiura, Takashi

    2018-06-01

    To identify risk factors for symptomatic radiation pneumonitis (RP) after stereotactic radiation therapy (SRT) for lung tumours. We retrospectively evaluated 68 lung tumours in 63 patients treated with SRT between 2011 and 2015. RP was graded according to the National Cancer Institute-Common Terminology Criteria for Adverse Events version 4.0. SRT was delivered at 7.0-12.0 Gy per each fraction, once daily, to a total of 48-64 Gy (median, 50 Gy). Univariate analysis was performed to assess patient- and treatment-related factors, including age, sex, smoking index (SI), pulmonary function, tumour location, serum Krebs von den Lungen-6 value (KL-6), dose-volume metrics (V5, V10, V20, V30, V40 and VS5), homogeneity index of the planning target volume (PTV), PTV dose, mean lung dose (MLD), contralateral MLD and V2, PTV volume, lung volume and the PTV/lung volume ratio (PTV/Lung). Performance of PTV/Lung in predicting symptomatic RP was also analysed using receiver operating characteristic (ROC) analysis. The median follow-up period was 21 months. 10 of 63 patients (15.9%) developed symptomatic RP after SRT. On univariate analysis, V10, V20, PTV volume and PTV/Lung were significantly associated with occurrence of RP  ≥Grade 2. ROC curves indicated that symptomatic RP could be predicted using PTV/Lung [area under curve (AUC): 0.88, confidence interval (CI: 0.78-0.95), cut-off value: 1.09, sensitivity: 90.0% and specificity: 72.4%]. PTV/Lung is a good predictor of symptomatic RP after SRT. Advances in knowledge: The cases with high PTV/Lung should be carefully monitored with caution for the occurrence of RP after SRT.

  15. Radiation effects and tritium technology for fusion reactors. Volume I. Proceedings of the international conference, Gatlinburg, Tennessee, October 1--3, 1975

    Energy Technology Data Exchange (ETDEWEB)

    Watson, J.S.; Wiffen, F.W.; Bishop, J.L.; Breeden, B.K. (eds.)

    1976-03-01

    Separate abstracts were prepared for the 29 included papers in Vol. I. The topics covered in this volume include swelling and microstructures in thermonuclear reactor materials. Some papers on modeling and damage analysis are included. (MOW)

  16. SU-E-J-249: Correlation of Mean Lung Ventilation Value with Ratio of Total Lung Volumes

    International Nuclear Information System (INIS)

    Yu, N; Qu, H; Xia, P

    2014-01-01

    Purpose: Lung ventilation function measured from 4D-CT and from breathing correlated CT images is a novel concept to incorporate the lung physiologic function into treatment planning of radiotherapy. The calculated ventilation functions may vary from different breathing patterns, affecting evaluation of the treatment plans. The purpose of this study is to correlate the mean lung ventilation value with the ratio of the total lung volumes obtained from the relevant CTs. Methods: A ventilation map was calculated from the variations of voxel-to-voxel CT densities from two breathing phases from either 4D-CT or breathing correlated CTs. An open source image registration tool of Plastimatch was used to deform the inhale phase images to the exhale phase images. To calculate the ventilation map inside lung, the whole lung was delineated and the tissue outside the lung was masked out. With a software tool developed in house, the 3D ventilation map was then converted in the DICOM format associated with the planning CT images. The ventilation map was analyzed on a clinical workstation. To correlate ventilation map thus calculated with lung volume change, the total lung volume change was compared the mean ventilation from our method. Results: Twenty two patients who underwent stereotactic body irradiation for lung cancer was selected for this retrospective study. For this group of patients, the ratio of lung volumes for the inhale (Vin ) and exhale phase (Vex ) was shown to be linearly related to the mean of the local ventilation (Vent), Vin/Vex=1.+0.49*Vent (R2=0.93, p<0.01). Conclusion: The total lung volume change is highly correlated with the mean of local ventilation. The mean of local ventilation may be useful to assess the patient's lung capacity

  17. Volume-dependent hemodynamic effects of blood collection in canine donors - evaluation of 13% and 15% of total blood volume depletion

    Directory of Open Access Journals (Sweden)

    RUI R.F. FERREIRA

    2015-03-01

    Full Text Available Background: There is no consensus regarding the blood volume that could be safely donated by dogs, ranging from 11 to 25% of its total blood volume (TBV. No previous studies evaluated sedated donors.Aim: To evaluate the hemodynamic effects of blood collection from sedated and non-sedated dogs and to understand if such effects were volume-dependent.Materials and Methods: Fifty three donations of 13% of TBV and 20 donations of 15% TBV were performed in dogs sedated with diazepam and ketamine. Additionally, a total of 30 collections of 13% TBV and 20 collections of 15% TBV were performed in non-sedated dogs. Non-invasive arterial blood pressures and pulse rates were registered before and 15 min after donation. Results: Post-donation pulse rates increased significantly in both sedated groups, with higher differences in the 15% TBV collections. Systolic arterial pressures decreased significantly in these groups, while diastolic pressures increased significantly in 13% TBV donations. Non-sedated groups revealed a slight, but significant, SBP decrease. No clinical signs related to donations were registered.Conclusion: These results suggest that the collection of 15% TBV in sedated donors induces hemodynamic variations that may compromise the harmlessness of the procedure, while it seems to be a safe procedure in non-sedated dogs.

  18. Combining a gas turbine modular helium reactor and an accelerator and for near total destruction of weapons grade plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Baxter, A.M.; Lane, R.K.; Sherman, R. [General Atomics, San Diego, CA (United States)

    1995-10-01

    Fissioning surplus weapons-grade plutonium (WG-Pu) in a reactor is an effective means of rendering this stockpile non-weapons useable. In addition the enormous energy content of the plutonium is released by the fission process and can be captured to produce valuable electric power. While no fission option has been identified that can accomplish the destruction of more than about 70% of the WG-Pu without repeated reprocessing and recycling, which presents additional opportunities for diversion, the gas turbine modular helium-cooled reactor (GT-MHR), using an annular graphite core and graphite inner and outer reflectors combines the maximum plutonium destruction and highest electrical production efficiency and economics in an inherently safe system. Accelerator driven sub-critical assemblies have also been proposed for WG-Pu destruction. These systems offer almost complete WG-Pu destruction, but achieve this goal by using circulating aqueous or molten salt solutions of the fuel, with potential safety implications. By combining the GT-MHR with an accelerator-driven sub-critical MHR assembly, the best features of both systems can be merged to achieve the near total destruction of WG-Pu in an inherently safe, diversion-proof system in which the discharged fuel elements are suitable for long term high level waste storage without the need for further processing. More than 90% total plutonium destruction, and more than 99.9% Pu-239 destruction, could be achieved. The modular concept minimizes the size of each unit so that both the GT-MHR and the accelerator would be straightforward extensions of current technology.

  19. Licensed operating reactors. Status summary report, data as of August 31, 1983. Volume 8, No. 9

    International Nuclear Information System (INIS)

    1984-10-01

    This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  20. Operating history report for the Peach Bottom HTGR. Volume I. Reactor operating history

    International Nuclear Information System (INIS)

    Scheffel, W.J.; Baldwin, N.L.; Tomlin, R.W.

    1976-01-01

    The operating history for the Peach Bottom-1 Reactor is presented for the years 1966 through 1975. Information concerning general chemistry data, general physics data, location of sensing elements in the primary helium circuit, and postirradiation examination and testing of reactor components is presented

  1. Total Stem and Merchantable Volume Equations of Norway Spruce (Picea abies (L.) Karst.) Growing on Former Farmland in Sweden

    OpenAIRE

    Johansson, Tord

    2014-01-01

    An equation was constructed to estimate the stem volume of Norway spruce (Picea abies (L.) Karst.) in 145 stands growing on former farmland in Sweden (Latitude 56-63 degrees N). The mean total age was 40 +/- 13 (range 17-91) years, the mean diameter at breast height (ob) was 15 +/- 4 (range 5-27) cm and the mean density was 1621 +/- 902 (range 100-7600) stems ha(-1). The equation which fits the data best used the diameter at breast height and total stem height as predictive variables. Merchan...

  2. Increased epicardial fat volume quantified by 64-multidetector computed tomography is associated with coronary atherosclerosis and totally occlusive lesions

    International Nuclear Information System (INIS)

    Ueno, Koji; Anzai, Toshihisa; Jinzaki, Masahiro

    2009-01-01

    The relationship between the epicardial fat volume measured by 64-slice multidetector computed tomography (MDCT) and the extension and severity of coronary atherosclerosis was investigated. Both MDCT and conventional coronary angiography (CAG) were performed in 71 consecutive patients who presented with effort angina. The volume of epicardial adipose tissue (EAT) was measured by MDCT. The severity of coronary atherosclerosis was assessed by evaluating the extension of coronary plaques in 790 segments using MDCT data, and the percentage diameter stenosis in 995 segments using CAG data. The estimated volume of EAT indexed by body surface area was defined as VEAT. Increased VEAT was associated with advanced age, male sex, degree of metabolic alterations, a history of acute coronary syndrome (ACS) and the presence of total occlusions, and showed positive correlation with the stenosis score r=0.28, P=0.02) and the atheromatosis score (r=0.67, P 3 /m 2 ) to be the strongest independent determinant of the presence of total occlusions odds ratio 4.64. P=0.02). VEAT correlates with the degree of metabolic alterations and coronary atheromatosis. Excessive accumulation of EAT might contribute to the development of ACS and coronary total occlusions. (author)

  3. Safety Research Experiment Facility Project. Conceptual design report. Volume V. Reactor vessel and closure

    International Nuclear Information System (INIS)

    1975-12-01

    The Prestressed Concrete Reactor Vessel (PCRV) will serve as the primary pressure retaining structure for the Safety Research Experiment Facility (SAREF) reactor. The reactor core, control rod drive room, primary heat exchangers, and gas circulators will be located in cavities within the PCRV. The orientation of these cavities, except for the control rod drive room, will be similar to the high-temperature gas-cooled reactor (HTGR) designs that are currently proposed or under design. Due to the nature of this type of structure, all biological and radiological shielding requirements are incorporated into the basic vessel design. At the midcore plane there are three radially oriented slots that will extend from the outside surface of the PCRV to the reactor core liner. These slots will accommodate each of the fuel motion monitoring systems which will be part of the observation apparatus used with the loop experiments

  4. Advanced light water reactor utility requirements document: Volume 1--ALWR policy and summary of top-tier requirements

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    The U.S. utilities are leading an industry wide effort to establish the technical foundation for the design of the Advanced Light Water Reactor (ALWR). This effort, the ALWR Program, is being managed for the U.S. electric utility industry by the Electric Power Research Institute (EPRI) and includes participation and sponsorship of several international utility companies and close cooperation with the U.S. Department of Energy (DOE). The cornerstone of the ALWR Program is a set of utility design requirements which are contained in the ALWR Requirements Document. The purpose of the Requirement Document is to present a clear, complete statement of utility desires for their next generation of nuclear plants. The Requirements Document covers the entire plant up to the grid interface. It therefore is the basis for an integrated plant design, i.e., nuclear steam supply system and balance of plant, and it emphasizes those areas which are most important to the objective of achieving an ALWR which is excellent with respect to safety, performance, constructibility, and economics. The document applies to both Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The Requirements Document is organized in three volumes. Volume 1 summarizes AlWR Program policy statements and top-tier requirements. The top-tier design requirements are categorized by major functions, including safety and investment protection, performance, and design process and constructibility. There is also a set of general design requirements, such as simplification and proven technology, which apply broadly to the ALWR design, and a set of economic goals for the ALWR program. The top-tier design requirements are described further in Volume 1 and are formally invoked as requirements in Volumes 2 and 3

  5. Association of Automatically Quantified Total Blood Volume after Aneurysmal Subarachnoid Hemorrhage with Delayed Cerebral Ischemia

    NARCIS (Netherlands)

    Zijlstra, I. A.; Gathier, C. S.; Boers, A. M.; Marquering, H. A.; Slooter, A. J.; Velthuis, B. K.; Coert, B. A.; Verbaan, D.; van den Berg, R.; Rinkel, G. J.; Majoie, C. B.

    2016-01-01

    The total amount of extravasated blood after aneurysmal subarachnoid hemorrhage, assessed with semiquantitative methods such as the modified Fisher and Hijdra scales, is known to be a predictor of delayed cerebral ischemia. However, prediction rates of delayed cerebral ischemia are moderate, which

  6. Paid to Perform: Aligning Total Military Compensation With Talent Management. Volume 8

    Science.gov (United States)

    2015-06-01

    study upon which a wider compensation model can be built, we propose a system that thoughtfully integrates rede - signed basic pays and pensions...Taken together, these basic pay and pension rede - sign components will create a far more cost-effective total compensation system than the one

  7. Licensing assessment of the Candu Pressurized Heavy Water Reactor. Preliminary safety information document. Volume II

    International Nuclear Information System (INIS)

    1977-06-01

    ERDA has requested United Engineers and Constructors (UE and C) to evaluate the design of the Canadian natural uranium fueled, heavy water moderated (CANDU) nuclear reactor power plant to assess its conformance with the licensing criteria and guidelines of the U.S. Nuclear Regulatory Commission (USNRC) for light water reactors. This assessment was used to identify cost significant items of nonconformance and to provide a basis for developing a detailed cost estimate for a 1140 MWe, 3-loop Pressurized Heavy Water Reactor (PHWR) located at the Middletown, USA Site

  8. Licensed operating reactors: status summary report, data as of May 31, 1983. Volume 7, No. 6

    International Nuclear Information System (INIS)

    1983-06-01

    The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  9. Institutional applications of solar total-energy systems. Draft final report. Volume 2. Appendixes

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-07-01

    The appendices present the analytical basis for the analysis of solar total energy (STE) systems. A regional-climate model and a building-load requirements model are developed, along with fuel-price scenarios. Life-cycle costs are compared for conventional-utility, total energy, and STE systems. Thermal STE system design trade-offs are performed and thermal STE system performance is determined. The sensitivity of STE competitiveness to fuel prices is examined. The selection of the photovoltaic array is briefly discussed. The institutional-sector decision processes are analyzed. Hypothetical regional back-up rates and electrical-energy costs are calculated. The algorithms and equations used in operating the market model are given, and a general methodology is developed for projecting the size of the market for STE systems and applied to each of 8 institutional subsectors. (LEW)

  10. Allometric relations of total volumes of prolactin cells and corticotropic cells to body length in the annual cyprinodont Cynolebias whitei: effects of environmental salinity, stress and ageing

    NARCIS (Netherlands)

    Ruijter, J. M.; Wendelaar Bonga, S. E.

    1987-01-01

    An analysis of the allometric relations of the total volumes occupied by prolactin (PRL) and corticotropic (ACTH) cells (PRL volume and ACTH volume, respectively) to body length and a study of the immunocytochemical staining intensity of PRL and ACTH cells were used to determine the differences in

  11. Advanced reactor safety research quarterly report, October-December 1982. Volume 24

    Energy Technology Data Exchange (ETDEWEB)

    None

    1984-04-01

    This report describes progress in a number of activities dealing with current safety issues relevant to both light water reactors (LWRs) and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

  12. Licensed operating reactors. Status summary report, data as of February 28, 1986. Volume 10, No. 3

    International Nuclear Information System (INIS)

    1986-04-01

    This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  13. Licensed operating reactors status summary report, data as of August 31, 1985. Volume 9, No. 9

    International Nuclear Information System (INIS)

    1985-10-01

    This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  14. Licensed operating reactors. Status summary report, data as of May 31, 1986. Volume 10, Number 6

    International Nuclear Information System (INIS)

    1986-08-01

    This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  15. Licensed operating reactors. Status summary report: data as of July 31, 1985. Volume 9, No. 8

    International Nuclear Information System (INIS)

    1985-09-01

    This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Head-quarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capabilitiy, reactor years of experience and non-power reactors in the United States

  16. Licensed operating reactors. Status summary report data as of January 31, 1986. Volume 10, No. 2

    International Nuclear Information System (INIS)

    1986-03-01

    This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  17. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  18. Maximized liquid radwaste volume reduction through a total integrated process: A new technology success story

    International Nuclear Information System (INIS)

    Rae, G.A.

    1996-01-01

    A fundamental nuclear industry goal is the minimization of the generation of radioactive waste. This goal has been dramatically reinforced over the past few years due to the spiraling increased costs of both commercial and DOE disposal. To assist in meeting these goals and reducing the industry's costs, NUKEM initiated a new technology program to maximize the reduction of liquid radwaste through the use of a systematic approach or TIPS (Total Integrated Process System). This concept evaluates the total life cycle of various technologies in a combination that results in the final waste form being minimized to the pure solids content of the waste stream. Additionally, it allows for a final waste form that maximizes the utilization of the waste package and is conditioned to be readily acceptable to additional processing to meet new waste form requirements at future disposal sites, should interim storage of the waste be required. The TIPS, although first introduced at commercial facilities, has broad applications for DOE's liquid waste streams

  19. Total Stem and Merchantable Volume Equations of Norway Spruce (Picea abies (L. Karst. Growing on Former Farmland in Sweden

    Directory of Open Access Journals (Sweden)

    Tord Johansson

    2014-08-01

    Full Text Available An equation was constructed to estimate the stem volume of Norway spruce (Picea abies (L. Karst. in 145 stands growing on former farmland in Sweden (Latitude 56–63° N. The mean total age was 40 ± 13 (range 17–91 years, the mean diameter at breast height (ob was 15 ± 4 (range 5–27 cm and the mean density was 1621 ± 902 (range 100–7600 stems ha−1. The equation which fits the data best used the diameter at breast height and total stem height as predictive variables. Merchantable volume equations for the estimation of commercial volume for any top diameter and bole length were developed. Soil types in the stands were sediments (coarse sand, fine sand and silt and heavy, medium and light clay, tills (sandy, fine sandy and silty and peat. The standing volume was calculated; the mean was 253 ± 103 (range 26–507 m3 ha−1 with a MAI (mean annual increment of 6.9±3.5 (range 1.3–16.7 m3 ha−1 year−1. There were statistically significant differences between MAI and coarse sand, sand and silt, light clay, peat and silty till soils. Spruce stands growing on silty tills had the lowest MAI (4.94 ± 2.27 m3 ha−1 year−1 and light clay, fine sand and silt and peat the highest (7.62 ± 4.24, 7.46 ± 3.33 and 8.67 ± 2.83 m3 ha−1 year−1.

  20. Proceedings of the US Nuclear Regulatory Commission twentieth water reactor safety information meeting; Volume 2, Severe accident research, Thermal hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Weiss, A.J. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1993-03-01

    This three-volume report contains papers presented at the Twentieth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 21--23, 1992. The papers describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included 10 different papers presented by researchersfrom CEC, China, Finland, France, Germany, Japan, Spain and Taiwan. Selected papers have been processed separately for inclusion in the Energy Science and Technology Database.

  1. Optimization of total arc degree for stereotactic radiotherapy by using integral biologically effective dose and irradiated volume

    International Nuclear Information System (INIS)

    Lim, Do Hoon; Kim, Dae Yong; Lee, Myung Za; Chun, Ha Chung

    2001-01-01

    To find the optimal values of total arc degree to protect the normal brain tissue from high dose radiation in stereotactic radiotherapy planning. With Xknife-3 planning system and 4 MV linear accelerator, the authors planned under various values of parameters. One isocenter, 12, 20, 30, 40, 50, and 60 mm of collimator diameters, 100 deg, 200 deg, 300 deg, 400 deg, 500 deg, 600 deg, of total arc degrees, and 30 deg or 45 deg of arc intervals were used. After the completion of planning, the plans were compared each other using V 50 (the volume of normal brain that is delivered high dose radiation) and integral biologically effective dose. At 30 deg of arc interval, the values of V 50 had the decreased pattern with the increase of total arc degree in any collimator diameter. At 45 deg arc interval, up to 400 deg of total arc degree, the values of V 50 decreased with the increase of total arc degree, but at 500 deg and 600 deg of total arc degrees, the values increased. At 30 deg of arc interval, integral biologically effective dose showed the decreased pattern with the increase of total arc degree in any collimator diameter. At 45 deg arc interval with less than 40 mm collimator diameter, the integral biologically effective dose decreased with the increase of total arc degree, but with 50 and 60 mm of collimator diameters, up to 400 deg of total arc degree, integral biologically effective dose decreased with the increase of total arc degree, but at 500 deg and 600 deg of total arc degrees, the values increased. In the stereotactic radiotherapy planning for brain lesions, planning with 400 deg of total arc degree is optimal. Especially, when the larger collimator more than 50 mm diameter should be used, the uses of 500 deg and 600 deg of total arc degrees make the increase of V 50 and integral biologically effective dose, Therefore stereotactic radiotherapy planning using 400 deg of total arc degree can increase the therapeutic ratio and produce the effective outcome

  2. Viability Assessment of a Repository at Yucca Mountain. Volume 3: Total System Performance Assessment

    Energy Technology Data Exchange (ETDEWEB)

    None

    1998-12-01

    This volume reports the development of TSPA for the VA. This first section defines the general process involved in developing any TSPA, it describes the overall TSPA process as implemented by programs in the US and elsewhere in the world, and discusses the acceptability of TSPA as a process or tool for analyzing a nuclear waste repository system. Section 2 discusses the more specific use of the TSPA process for the TSPA-VA for Yucca Mountain, including goals, approach, and methods. It also includes a very brief synopsis of TSPA-VA results. Section 3 briefly discusses each of the component models that comprise the TSPA-VA. Each TSPA component model represents a discrete set of processes. The TSPA-VA components are: unsaturated zone flow, thermal hydrology, near- field geochemical environment, waste package degradation, waste form alteration and mobilization, unsaturated zone transport, saturated zone flow and transport, and biosphere. For each of these components, this section introduces the conceptualization of each individual process, describes the data sources, and discusses model parameter development and computer methods used to simulate each component. Section 4 explains the mechanics of how the individual TSPA components were combined into a ''base case'' and then provides the ''expected value'' results of a deterministic base case analysis. Section 4 also contains a description of the probabilistic analyses and results that help determine the relative importance of the various TSPA components and the data used to describe the components. Section 5 addresses sensitivity studies run for each of the TSPA components to understand how uncertainty in various parameters within a component change the TSPA results. Section 6 presents the findings of the sensitivity studies run on the various components in Section 5, and prioritizes the findings of the entire set of uncertainty and sensitivity studies of the components relative

  3. RETRAN code analysis of Tsuruga-2 plant chemical volume control system (CVCS) reactor coolant leakage incident

    International Nuclear Information System (INIS)

    Kawai, H.

    2001-01-01

    JAPC purchased RETRAN, a program for transient thermal hydraulic analysis of complex fluid flow system, from the U.S. Electric Power Research Institute in 1992. Since then, JAPC has been utilizing RETRAN to evaluate safety margins of actual plant operation, in coping with troubles (investigating trouble causes and establishing countermeasures), and supporting reactor operation (reviewing operational procedures etc.). In this paper, a result of plant analysis performed on a CVCS reactor primary coolant leakage incident which occurred at JAPC's Tsuruga-2 plant (4-loop PWR, 3423 MWt, 1160 MW) on July 12 of 1999 and, based on the result, we made a plan to modify our operational procedure for reactor primary coolant leakage events in order to make earlier plant shutdown and this reduced primary coolant leakage. (author)

  4. Physics of plutonium recycling: volume V. Plutonium recycling in fast reactors

    International Nuclear Information System (INIS)

    1996-01-01

    As part of a programme proposed by the OECD/NEA Working Party on Physics of Plutonium Recycling (WPPR) to evaluate different scenarios for the use of plutonium, fast reactor physics benchmarks were developed. In this report, the multi-recycle performance of the metal-fuelled benchmark is evaluated. Benchmark results assess the reactor performance and toxicity behaviour in a closed nuclear fuel cycle for a parametric variation of the conversion ratio between 0.5 and 1.0. Results indicate that a fast burner reactor closed fuel cycle can be utilised to significantly reduce the radiotoxicity originating in the LWR cycle which would otherwise be destined for burial. (Author). tabs., figs., refs

  5. Licensed operating reactors. Status summary report as of February 29, 1984. Volume 8, No. 3

    International Nuclear Information System (INIS)

    1984-04-01

    The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Resource Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement (IE), from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  6. The Oak Ridge Research Reactor: safety analysis: Volume 2, supplement 2

    International Nuclear Information System (INIS)

    Hurt, S.S.

    1986-11-01

    The Oak Ridge Research Reactor Safety Analysis was last updated via ORNL-4169, Vol. 2, Supplement 1, in May of 1978. Since that date, several changes have been effected through the change-memo system described below. While these changes have involved the cooling system, the electrical system, and the reactor instrumentation and controls, they have not, for the most part, presented new or unreviewed safety questions. However, some of the changes have been based on questions or recommendations stemming from safety reviews or from reactor events at other sites. This paper discusses those changes which were judged to be safety related and which include revisions to the syphon-break system and changes related to seismic considerations which were very recently completed. The maximum hypothetical accident postulated in the original safety analysis requires dynamic containment and filtered flow for compliance with 10CFR100 limits at the site boundary

  7. Special topics reports for the reference tandem mirror fusion breeder. Volume 2. Reactor safety assessment

    International Nuclear Information System (INIS)

    Maya, I.; Hoot, C.G.; Wong, C.P.C.; Schultz, K.R.; Garner, J.K.; Bradbury, S.J.; Steele, W.G.; Berwald, D.H.

    1984-09-01

    The safety features of the reference fission suppressed fusion breeder reactor are presented. These include redundancy and overcapacity in primary coolant system components to minimize failure probability, an improved valve location logic to provide for failed component isolation, and double-walled coolant piping and steel guard vessel protection to further limit the extent of any leak. In addition to the primary coolant and decay heat removal system, reactor safety systems also include an independent shield cooling system, the module safety/fuel transfer coolant system, an auxiliary first wall cooling system, a psssive dump tank cooling system based on the use of heat pipes, and several lithium fire suppression systems. Safety system specifications are justified based on the results of thermal analysis, event tree construction, consequence calculations, and risk analysis. The result is a reactor design concept with an acceptably low probability of a major radioactivity release. Dose consequences of maximum credible accidents appear to be below 10CFR100 regulatory limits

  8. Boiling water reactor turbine trip (TT) benchmark. Volume II: Summary Results of Exercise 1

    International Nuclear Information System (INIS)

    Akdeniz, Bedirhan; Ivanov, Kostadin N.; Olson, Andy M.

    2005-06-01

    The OECD Nuclear Energy Agency (NEA) completed under US Nuclear Regulatory Commission (NRC) sponsorship a PWR main steam line break (MSLB) benchmark against coupled system three-dimensional (3-D) neutron kinetics and thermal-hydraulic codes. Another OECD/NRC coupled-code benchmark was recently completed for a BWR turbine trip (TT) transient and is the object of the present report. Turbine trip transients in a BWR are pressurisation events in which the coupling between core space-dependent neutronic phenomena and system dynamics plays an important role. The data made available from actual experiments carried out at the Peach Bottom 2 plant make the present benchmark particularly valuable. While defining and coordinating the BWR TT benchmark, a systematic approach and level methodology not only allowed for a consistent and comprehensive validation process, but also contributed to the study of key parameters of pressurisation transients. The benchmark consists of three separate exercises, two initial states and five transient scenarios. The BWR TT Benchmark will be published in four volumes as NEA reports. CD-ROMs will also be prepared and will include the four reports and the transient boundary conditions, decay heat values as a function of time, cross-section libraries and supplementary tables and graphs not published in the paper version. BWR TT Benchmark - Volume I: Final Specifications was issued in 2001 [NEA/NSC/DOC(2001)]. The benchmark team [Pennsylvania State University (PSU) in co-operation with Exelon Nuclear and the NEA] has been responsible for coordinating benchmark activities, answering participant questions and assisting them in developing their models, as well as analysing submitted solutions and providing reports summarising the results for each phase. The benchmark team has also been involved in the technical aspects of the benchmark, including sensitivity studies for the different exercises. Volume II summarises the results for Exercise 1 of the

  9. Total-system performance assessment for Yucca Mountain - SNL second iteration (TSPA-1993); Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, M.L.; Gauthier, J.H.; Barnard, R.W.; Barr, G.E.; Dockery, H.A.; Dunn, E.; Eaton, R.R.; Guerin, D.C.; Lu, N.; Martinez, M.J. [and others

    1994-04-01

    Sandia National Laboratories has completed the second iteration of the periodic total-system performance assessments (TSPA-93) for the Yucca Mountain Site Characterization Project (YMP). These analyses estimate the future behavior of a potential repository for high-level nuclear waste at the Yucca Mountain, Nevada, site under consideration by the Department of Energy. TSPA-93 builds upon previous efforts by emphasizing YMP concerns relating to site characterization, design, and regulatory compliance. Scenarios describing expected conditions (aqueous and gaseous transport of contaminants) and low-probability events (human-intrusion drilling and volcanic intrusion) are modeled. The hydrologic processes modeled include estimates of the perturbations to ambient conditions caused by heating of the repository resulting from radioactive decay of the waste. Hydrologic parameters and parameter probability distributions have been derived from available site data. Possible future climate changes are modeled by considering two separate groundwater infiltration conditions: {open_quotes}wet{close_quotes} with a mean flux of 10 mm/yr, and {open_quotes}dry{close_quotes} with a mean flux of 0.5 mm/yr. Two alternative waste-package designs and two alternative repository areal thermal power densities are investigated. One waste package is a thin-wall container emplaced in a vertical borehole, and the second is a container designed with corrosion-resistant and corrosion-allowance walls emplaced horizontally in the drift. Thermal power loadings of 57 kW/acre (the loading specified in the original repository conceptual design) and 114 kW/acre (a loading chosen to investigate effects of a {open_quotes}hot repository{close_quotes}) are considered. TSPA-93 incorporates significant new detailed process modeling, including two- and three-dimensional modeling of thermal effects, groundwater flow in the saturated-zone aquifers, and gas flow in the unsaturated zone.

  10. Total-system performance assessment for Yucca Mountain - SNL second iteration (TSPA-1993); Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, M.L.; Barnard, R.W.; Barr, G.E.; Dockery, H.A.; Dunn, E.; Eaton, R.R.; Martinez, M.J. [Sandia National Labs., Albuquerque, NM (United States); Gauthier, J.H.; Guerin, D.C.; Lu, N. [and others

    1994-04-01

    Sandia National Laboratories has completed the second iteration of the periodic total-system performance assessments (TSPA-93) for the Yucca Mountain Site Characterization Project (YMP). These analyses estimate the future behavior of a potential repository for high-level nuclear waste at the Yucca Mountain, Nevada, site under consideration by the Department of Energy. TSPA-93 builds upon previous efforts by emphasizing YMP concerns relating to site characterization, design, and regulatory compliance. Scenarios describing expected conditions (aqueous and gaseous transport of contaminants) and low-probability events (human-intrusion drilling and volcanic intrusion) are modeled. The hydrologic processes modeled include estimates of the perturbations to ambient conditions caused by heating of the repository resulting from radioactive decay of the waste. Hydrologic parameters and parameter probability distributions have been derived from available site data. Possible future climate changes are modeled by considering two separate groundwater infiltration conditions: {open_quotes}wet{close_quotes} with a mean flux of 10 mm/yr, and {open_quotes}dry{close_quotes} with a mean flux of 0.5 mm/yr. Two alternative waste-package designs and two alternative repository areal thermal power densities are investigated. One waste package is a thin-wall container emplaced in a vertical borehole, and the second is a container designed with corrosion-resistant and corrosion-allowance walls emplaced horizontally in the drift. Thermal power loadings of 57 kW/acre (the loading specified in the original repository conceptual design) and 114 kW/acre (a loading chosen to investigate effects of a {open_quotes}hot repository{close_quotes}) are considered. TSPA-93 incorporates significant new detailed process modeling, including two- and three-dimensional modeling of thermal effects, groundwater flow in the saturated-zone aquifers, and gas flow in the unsaturated zone.

  11. Licensed operating reactors. Status summary report: data as of September 30, 1985. Volume 9, No. 10

    International Nuclear Information System (INIS)

    1985-11-01

    This report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Resource Management from the Headquarters staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, IE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non-power reactors in the US

  12. Environmentally assisted cracking in light water reactors. Semiannual report, April 1994--September 1994, Volume 19

    International Nuclear Information System (INIS)

    Chopra, O.K.; Chung, H.M.; Gavenda, D.J.

    1995-09-01

    This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors from April to September 1994. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in piping and reactor pressure vessels, (b) EAC of austenitic stainless steels (SSs) and Alloy 600, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests have been conducted on A106-Gr B and A533-Gr B steels in oxygenated water to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack growth data were obtained on fracture-mechanics specimens of SSs and Alloy 600 to investigate EAC in simulated boiling water reactor (BWR) and pressurized water reactor environments at 289 degrees C. The data were compared with predictions from crack growth correlations developed at ANL for SSs in water and from rates in air from Section XI of the ASME Code. Microchemical changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials

  13. Environmentally assisted cracking in Light Water Reactors: Semiannual report, April 1993--September 1993. Volume 17

    International Nuclear Information System (INIS)

    Chopra, O.K.; Chung, H.M.; Karlsen, T.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E.; Sanecki, J.E.; Shack, W.J.; Soppet, W.K.

    1994-06-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) during the six months from April 1993 to September 1993. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels; (b) EAC of cast stainless steels (SSs); and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions in simulated boiling-water reactor (BWR) water at 289 degree C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section 11 of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy

  14. Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

    International Nuclear Information System (INIS)

    1992-07-01

    Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility

  15. Advanced Instrumentation and Control Methods for Small and Medium Reactors with IRIS Demonstration. Final Report. Volume 1

    International Nuclear Information System (INIS)

    Hines, J. Wesley; Upadhyaya, Belle R.; Doster, J. Michael; Edwards, Robert M.; Lewis, Kenneth D.; Turinsky, Paul; Coble, Jamie

    2011-01-01

    on meeting two of the eight needs outlined in the recently published 'Technology Roadmap on Instrumentation, Control, and Human-Machine Interface (ICHMI) to Support DOE Advanced Nuclear Energy Programs' which was created 'to provide a systematic path forward for the integration of new ICHMI technologies in both near-term and future nuclear power plants and the reinvigoration of the U.S. nuclear ICHMI community and capabilities.' The research consortium is led by The University of Tennessee (UT) and is focused on three interrelated topics: Topic 1 (simulator development and measurement sensitivity analysis) is led by Dr. Mike Doster with Dr. Paul Turinsky of North Carolina State University (NCSU). Topic 2 (multivariate autonomous control of modular reactors) is led by Dr. Belle Upadhyaya of the University of Tennessee (UT) and Dr. Robert Edwards of Penn State University (PSU). Topic 3 (monitoring, diagnostics, and prognostics system development) is led by Dr. Wes Hines of UT. Additionally, South Carolina State University (SCSU, Dr. Ken Lewis) participated in this research through summer interns, visiting faculty, and on-campus research projects identified throughout the grant period. Lastly, Westinghouse Science and Technology Center (Dr. Mario Carelli) was a no-cost collaborator and provided design information related to the IRIS demonstration platform and defining needs that may be common to other SMR designs. The results of this research are reported in a six-volume Final Report (including the Executive Summary, Volume 1). Volumes 2 through 6 of the report describe in detail the research and development under the topical areas. This volume serves to introduce the overall NERI-C project and to summarize the key results. Section 2 provides a summary of the significant contributions of this project. A list of all the publications under this project is also given in Section 2. Section 3 provides a brief summary of each of the five volumes (2-6) of the report. The

  16. Environmentally assisted cracking in light water reactors. Semiannual report, October 1993--March 1994. Volume 18

    International Nuclear Information System (INIS)

    Chung, H.M.; Chopra, O.K.; Erck, R.A.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E.; Sanecki, J.E.; Shack, W.J.; Soppet, W.K.

    1995-03-01

    This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1993 to March 1994. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns in operating plants and as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels, (b) EAC of wrought and cast austenitic stainless steels (SSs), and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS after accumulation of relatively high fluence. Fatigue tests have been conducted on A302-Gr B low-alloy steel to verify whether the current predictions of modest decreases of fatigue life in simulated pressurized water reactor water are valid for high-sulfur heats that show environmentally enhanced fatigue crack growth rates. Additional crack growth data were obtained on fracture-mechanics specimens of austenitic SSs to investigate threshold stress intensity factors for EAC in high-purity oxygenated water at 289 degrees C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating boiling water reactors were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements, which are not specified in the ASTM specifications, may contribute to IASCC of solution-annealed materials

  17. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  18. Analysis of a total flow blockage of a Fuel Assembly in a typical MTR Research Reactor by RELAP5/MOD3.3

    International Nuclear Information System (INIS)

    Adorni, M.; Salah, A.B.; Di Maro, B.; Pierro, F.; D'Auria, F.; Hamidouche, T.

    2004-01-01

    The lack of full understanding of complex mechanisms connected with the interaction between thermal-hydraulics and neutronics still challenge the design and the operation of nuclear reactors by the adoption of conservative safety limits. The recent availability of powerful computer and computational techniques together with the continuing increase in operational experience imposes the revisiting of those areas and eventually the identification of design/safety requirements that can be relaxed [1]. Currently, the enlarged commercial exploitation of nuclear Research Reactors (RR) has increased the consideration to their corresponding safety issues. Almost all of the safety analyses have so far been performed using conservative computational tools [2]. Nowadays, the application of Best-Estimate (BE) methods constitutes a real necessity in order to increase their commercial productivity. In this framework, an attempt is made to apply the BE technique to perform a safety evaluation under research reactors operational conditions. In fact, this technique has been largely verified and validated for power reactors using coupled system thermal-hydraulic and three-dimensional neutron kinetics [1]. For this purpose, as typical representative of research reactors, the IAEA 10 MW MTR Research Reactors problem [3] is considered. The system thermal-hydraulic RELAP5 [4] code was developed to simulate transient scenarios in Power reactors such PWR, BWR, VVER, etc. However, only limited work was performed to access the applicability of the code to Research Reactors operating conditions (low pressure, mass flow rates, power, etc) [5]. Previous works performed in this field are reported in [5], [6] and [7]. In this framework, total and partial blockage of a single Fuel Assembly cooling channel are investigated. As a first attempt the calculations are performed by applying the BE thermal-hydraulic system code RELAP5 alone using its point kinetic model to derive the instantaneous core

  19. Licensed operating reactors. Status summary report, data as of September 30, 1984. Volume 8, No. 10

    International Nuclear Information System (INIS)

    1984-11-01

    This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States. The percentage computations, Items 20 through 24 in Section 2, the vendor capacity factors, and actual vs. potential energy production are computed using actual data for the period of consideration. The percentages listed in power generation are computed as an arithmetic average. The factors for the life-span of each unit (the Cumulative column) are reported by the utility. Utility power production data is checked for consistency with previously submitted statistics

  20. Analysis of the total system life cycle cost for the Civilian Radioactive Waste Management Program: Volume 2, Supporting information

    International Nuclear Information System (INIS)

    1987-06-01

    This report provides cost estimates for the fifth evaluation of the adequacy of the fee and is consistent with the program strategy and plans. The total-system cost for the reference cases in the improved-performance system is estimated at $32.1 to $38.2 billion (expressed in constant 1986 collars) over the entire life of the system, or $1.5 to $1.6 billion more than that of the authorized system (i.e., the system without an MRS facility). The current estimate of the total-system cost for the reference cases in the improved-performance system is $3.8 to $5.4 billion higher than the estimate for the same system in the 1986 TSLCC analysis. In the case with the maximum increase, nearly all of the higher cost is due to a $5.2-billion increase in the costs of development and evaluation (D and E); all other system costs are essentially unchanged. The cost difference between the improved-performance system and the authorized system is smaller than the difference estimated in last year's TSLCC analysis. Volume 2 presents the detailed results for the 1987 analysis of the total-system life cycle cost (TSLCC). It consists of four sections: Section A presents the yearly flows of waste between waste-management facilities for the 12 aggregate logistics cases that were studied; Section B presents the annual total-system costs for each of the 30 TSLCC cases by major cost category; Section C presents the annual costs for the disposal of 16,000 canisters of defense high-level waste (DHLW) by major cost category for each of the 30 TSLCC cases; and Section D presents a summary of the cost-allocation factors that were calculated to determine the defense waste share of the total-system costs

  1. Light-water reactors: preliminary safety and environmental information document. Volume I

    International Nuclear Information System (INIS)

    1980-01-01

    Information is presented concerning the reference PWR reactor system; once-through, low-enrichment uranium-235 fuel, 30 MWD per kilogram (PWR LEU(5)-OT); once-through, low-enrichment, high-burnup uranium fuel (PWR LEU(5)-Mod OT); self-generated plutonium spiked recycle (PWR LEU(5)-Pu-Spiked Recycle); denatured uranium-233/thorium cycle (PWR DU(3)-Th Recycle DU(3)); and plutonium/thorium cycle

  2. Review of the proposed Strategic National Plan for Civilian Nuclear Reactor Development: Volume 1

    International Nuclear Information System (INIS)

    1986-10-01

    On August 9, 1985, the Secretary of Energy requested that the Chairman of the Energy Research Advisory Board establish an ad-hoc Panel to review a draft ''Strategic National Plan for Civilian Nuclear Reactor Development.'' The resulting report, approved by the Board, contains suggestions for improving the draft plan and also contains major recommendations for alleviating the several institutional barriers that appear to preclude the construction of any new nuclear power plants in this country

  3. Reactor-building-basement radionuclide and source distribution studies. Volume 3

    International Nuclear Information System (INIS)

    Cox, T.E.; Horan, J.T.; Worku, G.

    1983-06-01

    The Three Mile Island Unit 2 (TMI-2) Reactor Building basement has been sampled several times since August 1979. This report compiles the analytical results and sample history for the liquid and solid samples obtained to date. In addition, basement radiation levels were also obtained using thermoluminescent dosimeters (TLDs). The data obtained will provide information to support ongoing mass balance and source term studies and will aid in characterizing the 282-ft elevation for decontamination planning and dose reduction

  4. Environmentally assisted cracking in Light Water Reactors: Semiannual report, October 1994--March 1995. Volume 20

    International Nuclear Information System (INIS)

    Chung, H.M.; Chopra, O.K.; Gavenda, D.J.; Hins, A.G.; Kassner, T.F.; Ruther, W.E.; Shack, W.J.; Soppet, W.K.

    1996-01-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) from October 1994 to March 1995. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, (b) EAC of Alloy 600 and 690, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water with several dissolvedoxygen (DO) concentrations to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Tensile properties and microstructures of several heats of Alloy 600 and 690 were characterized for correlation with EAC of the alloys in simulated LWR environments. Effects of DO and electrochemical potential on susceptibility to intergranular cracking of high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath irradiated in boiling water reactors were determined in slow-strain-rate-tensile tests at 289 degrees C. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials

  5. Licensed operating reactors status summary report, data as of December 31, 1985. Volume 10, No. 1

    International Nuclear Information System (INIS)

    1986-01-01

    The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Resource Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the Errata page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  6. Total physical activity volume, physical activity intensity, and metabolic syndrome: 1999-2004 National Health and Nutrition Examination Survey.

    Science.gov (United States)

    Churilla, James R; Fitzhugh, Eugene C

    2012-02-01

    This study examined the association of total physical activity volume (TPAV) and physical activity (PA) from three domains [leisure-time physical activity (LTPA), domestic, transportation] with metabolic syndrome. We also investigated the relationship between LTPA intensity and metabolic syndrome risk. Sample included adults who participated in the 1999-2004 National Health and Nutrition Examination Survey. Physical activity measures were created for TPAV, LTPA, domestic PA, and transportational PA. For each, a six-level measure based upon no PA (level 1) and quintiles (levels 2-6) of metabolic equivalents (MET)·min·wk(-1) was created. A three-level variable associated with the current Department of Health and Human Services (DHHS) PA recommendation was also created. SAS and SUDAAN were used for the statistical analysis. Adults reporting the greatest volume of TPAV and LTPA were found to be 36% [odds ratio (OR) 0.64; 95% confidence interval (CI) 0.49-0.83] and 42% (OR 0.58; 95% CI 0.43-0.77), respectively, less likely to have metabolic syndrome. Domestic and transportational PA provided no specific level of protection from metabolic syndrome. Those reporting a TPAV that met the DHHS PA recommendation were found to be 33% (OR 0.67; 95%; CI 0.55-0.83) less likely to have metabolic syndrome compared to their sedentary counterparts. Adults reporting engaging in only vigorous-intensity LTPA were found to be 37% (OR 0.63; 95 CI 0.42-0.96) to 56% (OR 0.44; 95% CI 0.29-0.67) less likely to have metabolic syndrome. Volume, intensity, and domain of PA may all play important roles in reducing the prevalence and risk of metabolic syndrome.

  7. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 3, Sessions 12-16

    Energy Technology Data Exchange (ETDEWEB)

    Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)

    1995-09-01

    This document, Volume 3, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, ad the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected abstracts have been indexed separately for inclusion in the Energy Science and Technology Database.

  8. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 2, Sessions 6-11

    Energy Technology Data Exchange (ETDEWEB)

    Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)

    1995-09-01

    This document, Volume 2, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  9. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

    International Nuclear Information System (INIS)

    Block, R.C.; Feiner, F.

    1995-09-01

    This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  10. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

    Energy Technology Data Exchange (ETDEWEB)

    Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)

    1995-09-01

    This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  11. The component content of active particles in a plasma-chemical reactor based on volume barrier discharge

    Science.gov (United States)

    Soloshenko, I. A.; Tsiolko, V. V.; Pogulay, S. S.; Terent'yeva, A. G.; Bazhenov, V. Yu; Shchedrin, A. I.; Ryabtsev, A. V.; Kuzmichev, A. I.

    2007-02-01

    In this paper the results of theoretical and experimental studies of the component content of active particles formed in a plasma-chemical reactor composed of a multiple-cell generator of active particles, based on volume barrier discharge, and a working chamber are presented. For calculation of the content of uncharged plasma components an approach is proposed which is based on averaging of the power introduced over the entire volume. Advantages of such an approach lie in an absence of fitting parameters, such as the dimensions of microdischarges, their surface density and rate of breakdown. The calculation and the experiment were accomplished with the use of dry air (20% relative humidity) as the plasma generating medium. Concentrations of O3, HNO3, HNO2, N2 O5 and NO3 were measured experimentally in the discharge volume and working chamber for the residence time of particles on a discharge of 0.3 s and more and discharge specific power of 1.5 W cm-3. It has been determined that the best agreement between the calculation and the experiment occurs at calculated gas medium temperatures in the discharge plasma of about 400-425 K, which correspond to the experimentally measured rotational temperature of nitrogen. In most cases the calculated concentrations of O3, HNO3, HNO2, N2O5 and NO3 for the barrier discharge and the working chamber are in fairly good agreement with the respective measured values.

  12. The component content of active particles in a plasma-chemical reactor based on volume barrier discharge

    International Nuclear Information System (INIS)

    Soloshenko, I A; Tsiolko, V V; Pogulay, S S; Terent'yeva, A G; Bazhenov, V Yu; Shchedrin, A I; Ryabtsev, A V; Kuzmichev, A I

    2007-01-01

    In this paper the results of theoretical and experimental studies of the component content of active particles formed in a plasma-chemical reactor composed of a multiple-cell generator of active particles, based on volume barrier discharge, and a working chamber are presented. For calculation of the content of uncharged plasma components an approach is proposed which is based on averaging of the power introduced over the entire volume. Advantages of such an approach lie in an absence of fitting parameters, such as the dimensions of microdischarges, their surface density and rate of breakdown. The calculation and the experiment were accomplished with the use of dry air (20% relative humidity) as the plasma generating medium. Concentrations of O 3 , HNO 3 , HNO 2 , N 2 O 5 and NO 3 were measured experimentally in the discharge volume and working chamber for the residence time of particles on a discharge of 0.3 s and more and discharge specific power of 1.5 W cm -3 . It has been determined that the best agreement between the calculation and the experiment occurs at calculated gas medium temperatures in the discharge plasma of about 400-425 K, which correspond to the experimentally measured rotational temperature of nitrogen. In most cases the calculated concentrations of O 3 , HNO 3 , HNO 2 , N 2 O 5 and NO 3 for the barrier discharge and the working chamber are in fairly good agreement with the respective measured values

  13. Proceedings of the US Nuclear Regulatory Commission fifteenth water reactor safety information meeting: Volume 6, Decontamination and decommissioning, accident management, TMI-2

    International Nuclear Information System (INIS)

    Weiss, A.J.

    1988-02-01

    This six-volume report contains 140 papers out of the 164 that were presented at the Fifteenth Water Reactor Safety Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 26-29, 1987. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. This report, Volume 6, discusses decontamination and decommissioning, accident management, and the Three Mile Island-2 reactor accident. Thirteen reports have been cataloged separately

  14. Total energy cycle assessment of electric and conventional vehicles: an energy and environmental analysis. Volume 1: technical report

    Energy Technology Data Exchange (ETDEWEB)

    Cuenca, R.; Formento, J.; Gaines, L.; Marr, B.; Santini, D.; Wang, M. [Argonne National Lab., IL (United States); Adelman, S.; Kline, D.; Mark, J.; Ohi, J.; Rau, N. [National Renewable Energy Lab., Golden, CO (United States); Freeman, S.; Humphreys, K.; Placet, M. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-01-01

    This report compares the energy use, oil use and emissions of electric vehicles (EVs) with those of conventional, gasoline-powered vehicles (CVs) over the total life cycle of the vehicles. The various stages included in the vehicles` life cycles include vehicle manufacture, fuel production, and vehicle operation. Disposal is not included. An inventory of the air emissions associated with each stage of the life cycle is estimated. Water pollutants and solid wastes are reported for individual processes, but no comprehensive inventory is developed. Volume I contains the major results, a discussion of the conceptual framework of the study, and summaries of the vehicle, utility, fuel production, and manufacturing analyses. It also contains summaries of comments provided by external peer reviewers and brief responses to these comments.

  15. Stereotactic ultrasound for target volume definition in a patient with prostate cancer and bilateral total hip replacement.

    Science.gov (United States)

    Boda-Heggemann, Judit; Haneder, Stefan; Ehmann, Michael; Sihono, Dwi Seno Kuncoro; Wertz, Hansjörg; Mai, Sabine; Kegel, Stefan; Heitmann, Sigrun; von Swietochowski, Sandra; Lohr, Frank; Wenz, Frederik

    2015-01-01

    Target-volume definition for prostate cancer in patients with bilateral metal total hip replacements (THRs) is a challenge because of metal artifacts in the planning computed tomography (CT) scans. Magnetic resonance imaging (MRI) can be used for matching and prostate delineation; however, at a spatial and temporal distance from the planning CT, identical rectal and vesical filling is difficult to achieve. In addition, MRI may also be impaired by metal artifacts, even resulting in spatial image distortion. Here, we present a method to define prostate target volumes based on ultrasound images acquired during CT simulation and online-matched to the CT data set directly at the planning CT. A 78-year-old patient with cT2cNxM0 prostate cancer with bilateral metal THRs was referred to external beam radiation therapy. T2-weighted MRI was performed on the day of the planning CT with preparation according to a protocol for reproducible bladder and rectal filling. The planning CT was obtained with the immediate acquisition of a 3-dimensional ultrasound data set with a dedicated stereotactic ultrasound system for online intermodality image matching referenced to the isocenter by ceiling-mounted infrared cameras. MRI (offline) and ultrasound images (online) were thus both matched to the CT images for planning. Daily image guided radiation therapy (IGRT) was performed with transabdominal ultrasound and compared with cone beam CT. Because of variations in bladder and rectal filling and metal-induced image distortion in MRI, soft-tissue-based matching of the MRI to CT was not sufficient for unequivocal prostate target definition. Ultrasound-based images could be matched, and prostate, seminal vesicles, and target volumes were reliably defined. Daily IGRT could be successfully completed with transabdominal ultrasound with good accordance between cone beam CT and ultrasound. For prostate cancer patients with bilateral THRs causing artifacts in planning CTs, ultrasound referenced to

  16. Licensed operating reactors. Status summary report data as of March 31, 1986. Volume 10, No. 4

    International Nuclear Information System (INIS)

    1986-05-01

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Resource Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the ERRATA page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States. The percentage computations, Items 20 through 24 in Section 2, the vendor capacity factors on page 1-7, and actual vs potential energy production on page 1-2 are computed using actual data for the period of consideration. The percentages listed in power generation on page 1-2 are computed as a arithmetic average. The factors for the life-span of each unit (the ''Cumulative'' column) are reported by the utility and are not entirely recomputed by NRC. Utility power production data is checked for consistency with previously submitted statistics. It is hoped this status report proves informative and helpful to all agencies and individuals interested in analyzing trends in the nuclear industry which might have safety implications, or in maintaining an awareness of the US energy situation as a whole

  17. Licensed operating reactors. Status summary report, data as of October 31, 1985. Volume 9, No. 11

    International Nuclear Information System (INIS)

    1985-12-01

    The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Resource Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the errata page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States. The percentage computations, Items 20 through 24 in Section 2, the vendor capacity factors on page 1-7, and actual vs potential energy production on page 1-2 are computed using actual data for the period of consideration. The percentages listed in power generation on page 1-2 are computed as an arithmetic average. The factors for the life-span of each unit (the ''Cumulative'' column) are reported by the utility and are not entirely re-computed by NRC. Utility power production data is checked for consistency with previously submitted statistics

  18. Licensed operating reactors: Status summary report, data as of December 31, 1995. Volume 20

    International Nuclear Information System (INIS)

    1996-06-01

    The US Nuclear Regulatory Commission's monthly summary of licensed nuclear power reactor data is based primarily on the operating data report submitted by licensees for each unit. This report is divided into two sections: the first contains summary highlights and the second contains data on each individual unit in commercial operation. Section 1 availability factors, capacity factors, and forced outage rates are simple arithmetic averages. Section 2 items in the cumulative column are generally as reported by the licensees and notes to the use of weighted averages and starting dates other than commercial operation are provided

  19. Mirror Advanced Reactor Study (MARS). Final report. Volume 1-B. Commercial fusion electric plant

    International Nuclear Information System (INIS)

    Donohue, M.L.; Price, M.E.

    1984-07-01

    Volume 1-B contains the following chapters: (1) blanket and reflector; (2) central cell shield; (3) central cell structure; (4) heat transport and energy conversion; (5) tritium systems; (6) cryogenics; (7) maintenance; (8) safety; (9) radioactivity, activation, and waste disposal; (10) instrumentation and control; (11) balance of plant; (12) plant startup and operation; (13) plant availability; (14) plant construction; and (15) economic analysis

  20. VIPRE-01: a thermal-hydraulic code for reactor cores. Volume 2: user's manual (Revision 2)

    International Nuclear Information System (INIS)

    Cuta, J.M.; Koontz, A.S.; Stewart, C.W.; Montgomery, S.D.; Nomura, K.K.

    1985-07-01

    Revisions to the VIPRE code documents for Volume 2 are presented. These revisions conform to the changes made to VIPRE-01, CYCLE-00 to produce the new version of the code denoted by VIPRE-01, CYCLE-01. The first pages of the revisions specify where the replacement pages are to be inserted and which pages of the original documents should be retained

  1. Mirror Advanced Reactor Study (MARS). Final report. Volume 1-A. Commercial fusion electric plant

    International Nuclear Information System (INIS)

    Donohue, M.L.; Price, M.E.

    1984-07-01

    Volume 1-A contains the following chapters: (1) plasma engineering, (2) magnets, (3) ecr heating systems, (4) anchor ion-cyclotron resonance heating system, (5) sloshing ion neutral beam, (6) end cell structure, (7) end plasma technology, (8) fueling, (9) startup ion cyclotron resonant heating systems, and (10) end cell radiation analysis

  2. Mirror Advanced Reactor Study (MARS). Final report. Volume 1-B. Commercial fusion electric plant

    Energy Technology Data Exchange (ETDEWEB)

    Donohue, M.L.; Price, M.E. (eds.)

    1984-07-01

    Volume 1-B contains the following chapters: (1) blanket and reflector; (2) central cell shield; (3) central cell structure; (4) heat transport and energy conversion; (5) tritium systems; (6) cryogenics; (7) maintenance; (8) safety; (9) radioactivity, activation, and waste disposal; (10) instrumentation and control; (11) balance of plant; (12) plant startup and operation; (13) plant availability; (14) plant construction; and (15) economic analysis.

  3. Mirror Advanced Reactor Study (MARS). Final report. Volume 1-A. Commercial fusion electric plant

    Energy Technology Data Exchange (ETDEWEB)

    Donohue, M.L.; Price, M.E. (eds.)

    1984-07-01

    Volume 1-A contains the following chapters: (1) plasma engineering, (2) magnets, (3) ecr heating systems, (4) anchor ion-cyclotron resonance heating system, (5) sloshing ion neutral beam, (6) end cell structure, (7) end plasma technology, (8) fueling, (9) startup ion cyclotron resonant heating systems, and (10) end cell radiation analysis. (MOW)

  4. Single-slice epicardial fat area measurement. Do we need to measure the total epicardial fat volume?

    International Nuclear Information System (INIS)

    Oyama, Noriko; Goto, Daisuke; Ito, Yoichi M.

    2011-01-01

    The aim of this study was to assess a method for measuring epicardial fat volume (EFV) by means of a single-slice area measurement. We investigated the relation between a single-slice fat area measurement and total EFV. A series of 72 consecutive patients (ages 65±11 years; 36 men) who had undergone cardiac computed tomography (CT) on a 64-slice multidetector scanner with prospective electrocardiographic triggering were retrospectively reviewed. Pixels in the pericardium with a density range from -230 to -30 Hounsfield units were considered fat, giving the per-slice epicardial fat area (EFA). The EFV was estimated by the summation of EFAs multiplied by the slice thickness. We investigated the relation between total EFV and each EFA. EFAs measured at several anatomical landmarks - right pulmonary artery, origins of the left main coronary artery, right coronary artery, coronary sinus - all correlated with the EFV (r=0.77-0.92). The EFA at the LMCA level was highly reproducible and showed an excellent correlation with the EFV (r=0.92). The EFA is significantly correlated with the EFV. The EFA is a simple, quick method for representing the time-consuming EFV, which has been used as a predictive indicator of cardiovascular diseases. (author)

  5. Age and total and free prostate-specific antigen levels for predicting prostate volume in patients with benign prostatic hyperplasia.

    Science.gov (United States)

    Coban, Soner; Doluoglu, Omer Gokhan; Keles, Ibrahim; Demirci, Hakan; Turkoglu, Ali Riza; Guzelsoy, Muhammet; Karalar, Mustafa; Demirbas, Murat

    2016-06-01

    To investigate the predictive values of free prostate-specific antigen (fPSA), total PSA (tPSA) and age on the prostate volume. The data of 2148 patients with lower urinary tract symptoms were analyzed retrospectively. The patients who had transrectal ultrasonography guided 10 core biopsies owing to the findings obtained on digital rectal examination and presence of high PSA levels (PSA = 2.5-10 ng/dl), and proven to have BPH histopathologically were included in the study. Age, tPSA, fPSA and the prostate volumes (PV) of the patients were noted. One thousand patients that fulfilled the inclusion criteria were included in the study. The PV of the patients were significantly correlated with age, tPSA and fPSA (p < 0.001 and r = 0.307, p < 0.001 and r = 0.382, p < 0.001 and r = 0.296, respectively). On linear regression model, fPSA was found as a stronger predictive for PV (AUC = 0.75, p < 0.001) when compared to age (AUC = 0.64, p < 0.001), and tPSA (AUC = 0.69, p = 0.013). Although tPSA is an important prognostic factor for predicting PV, the predictive value of fPSA is higher. PV can easily be predicted by using age, and serum tPSA and fPSA levels.

  6. Association Between Provider Volume and Comorbidity on Hospital Utilization and Outcomes of Total Hip Arthroplasty Among National Health Insurance Enrollees

    Directory of Open Access Journals (Sweden)

    Chung-Shih Huang

    2011-06-01

    Conclusions: This study revealed that the volume of THAs performed by individual surgeons was a more important determinant of hospital utilization than hospital volume. Perioperative adverse events were associated with patients' age and comorbidity.

  7. Replacement Nuclear Research Reactor. Supplement to Draft Environmental Impact Statement. Volume 3

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-01-01

    The Draft Environmental Impact Statement for a replacement research reactor at Lucas Heights, was available for public examination and comment for some three months during 1998. A Supplement to the Draft Environmental Impact Statement (Draft EIS) has been completed and was lodged with Environment Australia on 18 January 1999. The Supplement is an important step in the overall environmental assessment process. It reviews submissions received and provides the proponent`s response to issues raised in the public review period. General issues extracted from submissions and addressed in the Supplement include concern over liability issues, Chernobyl type accidents, the ozone layer and health issues. Further studies, relating to issues raised in the public submission process, were undertaken for the Supplementary EIS. These studies confirm, in ANSTO`s view, the findings of the Draft EIS and hence the findings of the Final EIS are unchanged from the Draft EIS

  8. Licensed operating reactors. Status summary report data as of 12-31-94: Volume 19

    International Nuclear Information System (INIS)

    1995-04-01

    The Nuclear Regulatory Commission's annual summary of licensed nuclear power reactor data is based primarily on the report of operating data submitted by licensees for each unit for the month of December because that report contains data for the month of December, the year to date (in this case calendar year 1994) and cumulative data, usually from the date of commercial operation. The data is not independently verified, but various computer checks are made. The report is divided into two sections. The first contains summary highlights and the second contains data on each individual unit in commercial operation. Section 1 capacity and availability factors are simple arithmetic averages. Section 2 items in the cumulative column are generally as reported by the licensee and notes as to the use of weighted averages and starting dates other than commercial operation are provided

  9. Licensed operating reactors. Status summary report data as of 12-31-94: Volume 19

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-04-01

    The Nuclear Regulatory Commission`s annual summary of licensed nuclear power reactor data is based primarily on the report of operating data submitted by licensees for each unit for the month of December because that report contains data for the month of December, the year to date (in this case calendar year 1994) and cumulative data, usually from the date of commercial operation. The data is not independently verified, but various computer checks are made. The report is divided into two sections. The first contains summary highlights and the second contains data on each individual unit in commercial operation. Section 1 capacity and availability factors are simple arithmetic averages. Section 2 items in the cumulative column are generally as reported by the licensee and notes as to the use of weighted averages and starting dates other than commercial operation are provided.

  10. Replacement Nuclear Research Reactor. Supplement to Draft Environmental Impact Statement. Volume 3

    International Nuclear Information System (INIS)

    1999-01-01

    The Draft Environmental Impact Statement for a replacement research reactor at Lucas Heights, was available for public examination and comment for some three months during 1998. A Supplement to the Draft Environmental Impact Statement (Draft EIS) has been completed and was lodged with Environment Australia on 18 January 1999. The Supplement is an important step in the overall environmental assessment process. It reviews submissions received and provides the proponent's response to issues raised in the public review period. General issues extracted from submissions and addressed in the Supplement include concern over liability issues, Chernobyl type accidents, the ozone layer and health issues. Further studies, relating to issues raised in the public submission process, were undertaken for the Supplementary EIS. These studies confirm, in ANSTO's view, the findings of the Draft EIS and hence the findings of the Final EIS are unchanged from the Draft EIS

  11. Modification to ORIGEN2 for generating N Reactor source terms. Volume 1

    International Nuclear Information System (INIS)

    Schwarz, R.A.

    1997-04-01

    This report discusses work that has been done to upgrade the ORIGEN2 code cross sections to be compatible with the WIMS computer code data. Because of the changes in the ORIGEN2 calculations. Details on changes made to the ORIGEN2 computer code and the Radnuc code will be discussed along with additional work that should be done in the future to upgrade both ORIGEN2 and Radnuc. A detailed historical description of how source terms have been generated for N Reactor fuel stored in the K Basins has been generated. The neutron source discussed in this description was generated by the WIMS computer code (Gubbins et al. 1982) because of known shortcomings in the ORIGEN2 (Croff 1980) cross sections. Another document includes a discussion of the ORIGEN2 cross sections

  12. Fort Hood Solar Total Energy Project. Volume II. Preliminary design. Part 1. System criteria and design description. Final report

    Energy Technology Data Exchange (ETDEWEB)

    None,

    1979-01-01

    This volume documents the preliminary design developed for the Solar Total Energy System to be installed at Fort Hood, Texas. Current system, subsystem, and component designs are described and additional studies which support selection among significant design alternatives are presented. Overall system requirements which form the system design basis are presented. These include program objectives; performance and output load requirements; industrial, statutory, and regulatory standards; and site interface requirements. Material in this section will continue to be issued separately in the Systems Requirements Document and maintained current through revision throughout future phases of the project. Overall system design and detailed subsystem design descriptions are provided. Consideration of operation and maintenance is reflected in discussion of each subsystem design as well as in an integrated overall discussion. Included are the solar collector subsystem; the thermal storage subsystem, the power conversion sybsystem (including electrical generation and distribution); the heating/cooling and domestic hot water subsystems; overall instrumentation and control; and the STES building and physical plant. The design of several subsystems has progressed beyond the preliminary stage; descriptions for such subsystems are therefore provided in more detail than others to provide complete documentation of the work performed. In some cases, preliminary design parameters require specific verificaton in the definitive design phase and are identified in the text. Subsystem descriptions will continue to be issued and revised separately to maintain accuracy during future phases of the project. (WHK)

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

    International Nuclear Information System (INIS)

    Guerra, Bruno Teixeira

    2011-01-01

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

  14. Reactor analysis support package (RASP). Volume 7. PWR set-point methodology. Final report

    International Nuclear Information System (INIS)

    Temple, S.M.; Robbins, T.R.

    1986-09-01

    This report provides an overview of the basis and methodology requirements for determining Pressurized Water Reactor (PWR) technical specifications related setpoints and focuses on development of the methodology for a reload core. Additionally, the report documents the implementation and typical methods of analysis used by PWR vendors during the 1970's to develop Protection System Trip Limits (or Limiting Safety System Settings) and Limiting Conditions for Operation. The descriptions of the typical setpoint methodologies are provided for Nuclear Steam Supply Systems as designed and supplied by Babcock and Wilcox, Combustion Engineering, and Westinghouse. The description of the methods of analysis includes the discussion of the computer codes used in the setpoint methodology. Next, the report addresses the treatment of calculational and measurement uncertainties based on the extent to which such information was available for each of the three types of PWR. Finally, the major features of the setpoint methodologies are compared, and the principal effects of each particular methodology on plant operation are summarized for each of the three types of PWR

  15. Regulation characteristics of oxide generation and formaldehyde removal by using volume DBD reactor

    Science.gov (United States)

    Bingyan, CHEN; Xiangxiang, GAO; Ke, CHEN; Changyu, LIU; Qinshu, LI; Wei, SU; Yongfeng, JIANG; Xiang, HE; Changping, ZHU; Juntao, FEI

    2018-02-01

    Discharge plasmas in air can be accompanied by ultraviolet (UV) radiation and electron impact, which can produce large numbers of reactive species such as hydroxyl radical (OH·), oxygen radical (O·), ozone (O3), and nitrogen oxides (NO x ), etc. The composition and dosage of reactive species usually play an important role in the case of volatile organic compounds (VOCs) treatment with the discharge plasmas. In this paper, we propose a volume discharge setup used to purify formaldehyde in air, which is configured by a plate-to-plate dielectric barrier discharge (DBD) channel and excited by an AC high voltage source. The results show that the relative spectral-intensity from DBD cell without formaldehyde is stronger than the case with formaldehyde. The energy efficiency ratios (EERs) of both oxides yield and formaldehyde removal can be regulated by the gas flow velocity in DBD channel, and the most desirable processing effect is the gas flow velocity within the range from 2.50 to 3.33 m s-1. Moreover, the EERs of both the generated dosages of oxides (O3 and NO2) and the amount of removed formaldehyde can also be regulated by both of the applied voltage and power density loaded on the DBD cell. Additionally, the EERs of both oxides generation and formaldehyde removal present as a function of normal distribution with increasing the applied power density, and the peak of the function is appeared in the range from 273.5 to 400.0 W l-1. This work clearly demonstrates the regulation characteristic of both the formaldehyde removal and oxides yield by using volume DBD, and it is helpful in the applications of VOCs removal by using discharge plasma.

  16. Comparison between total lung capacity and residual volume values obtained by pletysmography and single breath methods with methane

    Directory of Open Access Journals (Sweden)

    Ricardo Marques Dias

    2006-11-01

    Full Text Available We analyzed pulmonary function tests of twenty asthmatic patients from Gaffrée e Guinle University Hospital, classified according to Brazilian Guidelines for Asthma (2002, similar to GINA, into mild persistent or moderate (9 or severe (11 asthma. We obtained parameters from spirometry, plethysmograph(PL and single breath technique for diffusion capacity (SB, with methane. Total lung capacity and residual volume were called TLCPL and RVPL when measured by pletysmography and TLCSB and RVSB when determined by single breath test. There were 13 women and 7 men with mean age of 47.6 years. The pulmonary dysfunction degree to FEV1/FVC was 58.8% with CI95=53.9 to 63.6. The mean values in litres for TLCPL (5.94 and RVPL (2.55 were significantly higher than for TLCSB (4.73 and RVSB (1.66. Multiple regression equations were determined for TLCPL e RVPL using only single breath values, TLCSB or RVSB, and spirographic parameters, with significant regression coefficients. However, the inclusion of spirometric parameters, except for FVC, did not improve the predicted capacity for the equations. Considering only the TLCSB, r2=0.79, the equation is: TLCPL=(TLCSB*1.025+1.088, with EPE=0.64. The regression for RVPL, r2=0.23, is: RVPL=(RVSB*0.9268+1.012. The results obtained after bronchodilation with 400 mcg of salbutamol did not improve the regression. We concluded that the SB technique did not obtain the same results as pletysmography for TLC and RV, but for TLC this difference can be predicted. Resumo: Foram analisados exames de função pulmonar de 20 asmáticos, em acompanhamento no HU Gaffrée Guinle, classificados, segundo o Consenso Brasileiro (2002, em asma leve persistente ou moderada (9 e grave (11. Foram obtidos os valores dos parâmetros da espirografia, da pletismografia e da técnica de respiração única, com metano, para a medida da difusão pulmonar (DLco. Assim, a capacidade pulmonar total e o volume residual, quando

  17. Proposed nuclear weapons nonproliferation policy concerning foreign research reactor spent nuclear fuel: Appendix B, foreign research reactor spent nuclear fuel characteristics and transportation casks. Volume 2

    International Nuclear Information System (INIS)

    1995-03-01

    This is Appendix B of a draft Environmental Impact Statement (EIS) on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel. It discusses relevant characterization and other information of foreign research reactor spent nuclear fuel that could be managed under the proposed action. It also discusses regulations for the transport of radioactive materials and the design of spent fuel casks

  18. Nuclear proliferation and civilian nuclear power. Report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle description

    International Nuclear Information System (INIS)

    1980-06-01

    The Nonproliferation Alterntive Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates

  19. Nuclear proliferation and civilian nuclear power. Report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle description

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    The Nonproliferation Alterntive Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates.

  20. Occupational radiation exposure at commercial nuclear power reactors and other facilities 1996: Twenty-ninth annual report. Volume 18

    International Nuclear Information System (INIS)

    Thomas, M.L.

    1998-02-01

    This report summarizes the occupational exposure data that are maintained in the US Nuclear Regulatory Commission's (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 1996 annual reports submitted by six of the seven categories of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. Since there are no geologic repositories for high level waste currently licensed, only six categories will be considered in this report. Annual reports for 1996 were received from a total of 300 NRC licensees, of which 109 were operators of nuclear power reactors in commercial operation. Compilations of the reports submitted by the 300 licensees indicated that 138,310 individuals were monitored, 75,139 of whom received a measurable dose. The collective dose incurred by these individuals was 21,755 person-cSv (person-rem) 2 which represents a 13% decrease from the 1995 value. The number of workers receiving a measurable dose also decreased, resulting in the average measurable dose of 0.29 cSv (rem) for 1996. The average measurable dose is defined to be the total collective dose (TEDE) divided by the number of workers receiving a measurable dose. These figures have been adjusted to account for transient reactor workers. Analyses of transient worker data indicate that 22,348 individuals completed work assignments at two or more licensees during the monitoring year. The dose distributions are adjusted each year to account for the duplicate reporting of transient workers by multiple licensees. In 1996, the average measurable dose calculated from reported was 0.24 cSv (rem). The corrected dose distribution resulted in an average measurable dose of 0.29 cSv (rem)

  1. Occupational radiation exposure at commercial nuclear power reactors and other facilities 1996: Twenty-ninth annual report. Volume 18

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, M.L. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Regulatory Applications; Hagemeyer, D. [Science Applications International Corp., Oak Ridge, TN (United States)

    1998-02-01

    This report summarizes the occupational exposure data that are maintained in the US Nuclear Regulatory Commission`s (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 1996 annual reports submitted by six of the seven categories of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. Since there are no geologic repositories for high level waste currently licensed, only six categories will be considered in this report. Annual reports for 1996 were received from a total of 300 NRC licensees, of which 109 were operators of nuclear power reactors in commercial operation. Compilations of the reports submitted by the 300 licensees indicated that 138,310 individuals were monitored, 75,139 of whom received a measurable dose. The collective dose incurred by these individuals was 21,755 person-cSv (person-rem){sup 2} which represents a 13% decrease from the 1995 value. The number of workers receiving a measurable dose also decreased, resulting in the average measurable dose of 0.29 cSv (rem) for 1996. The average measurable dose is defined to be the total collective dose (TEDE) divided by the number of workers receiving a measurable dose. These figures have been adjusted to account for transient reactor workers. Analyses of transient worker data indicate that 22,348 individuals completed work assignments at two or more licensees during the monitoring year. The dose distributions are adjusted each year to account for the duplicate reporting of transient workers by multiple licensees. In 1996, the average measurable dose calculated from reported was 0.24 cSv (rem). The corrected dose distribution resulted in an average measurable dose of 0.29 cSv (rem).

  2. Microcomputer based program for predicting heat transfer under reactor accident conditions. Volume I

    International Nuclear Information System (INIS)

    Cheng, S.C.; Groeneveld, D.C.; Leung, L.K.H.; Wong, Y.L.; Nguyen, C.

    1987-07-01

    A microcomputer based program called Heat Transfer Prediction Software (HTPS) has been developed. It calculates the heat transfer for the tube and bundle geometries for steady state and transient conditions. This program is capable of providing the best estimated of the hot pin temperatures during slow transients for 37- and 28-element CANDU type fuel bundles. The program is designed for an IBM-PC AT/XT (or IBM-PC compatible computer) equipped with a Math Co-processor. The following input parameters are required: pressure, mass flux, hydraulic diameter, and quality. For the steady state case, the critical heat flux (CHF), the critical heat flux temperature, the minimum film boiling temperature, and the minimum film boiling heat flux are the primary outputs. With either the surface heat flux or wall temperature specified, the program determines the heat transfer regime and calculates the surface heat flux, wall temperatures and heat transfer coefficient. For the slow transient case, the pressure, mass flux, quality, and volumetric heat generation rate are the time dependent input parameters required to calculate the hot pin sheath temperatures and surface heat fluxes. A simple routine for generating properties has been developed for light water to support the above program. It contains correlations that have been verified for pressures ranging from 0.6kPa to 30 MPa, and temperatures up to 1100 degrees Celcius. The thermodynamic and transport properties that can be generated from this routine are: density, specific volume, enthalpy, specific heat capacity, conductivity, viscosity, surface tension and Prandtl number for saturated liquid, saturated vapour, subcooled liquid for superheated vapour. A software for predicting flow regime has also been developed. It determines the flow pattern at specific flow conditions, and provides a correction factor for calculating the CHF during partially stratified horizontal flow. The technical bases for the program and its

  3. Microcomputer based program for predicting heat transfer under reactor accident conditions. Volume II

    International Nuclear Information System (INIS)

    Cheng, S.C.; Groeneveld, D.C.; Leung, L.K.H.; Wong, Y.L.; Nguyen, C.

    1987-07-01

    A microcomputer based program called Heat Transfer Prediction Software (HTPS) has been developed. It calculates the heat transfer for tube and bundle geometries for steady state and transient conditions. This program is capable of providing the best estimated of the hot pin temperatures during slow transients for 37- and 28-element CANDU type fuel bundles. The program is designed for an IBM-PC AT/XT (or IBM-PC compatible computer) equipped with a Math Co-processor. The following input parameters are required: pressure, mass flux, hydraulic diameter, and quality. For the steady state case, the critical heat flux (CHF), the critical heat flux temperature, the minimum film boiling temperature, and the minimum film boiling heat flux are the primary outputs. With either the surface heat flux or wall temperature specified, the program determines the heat transfer regime and calculates the surface heat flux, wall temperature and heat transfer coefficient. For the slow transient case, the pressure, mass flux, quality, and volumetric heat generation rate are the time dependent input parameters are required to calculate the hot pin sheath temperatures and surface heat fluxes. A simple routine for generating properties has been developed for light water to support the above program. It contains correlations that have been verified for pressures ranging from 0.6kPa to 30 MPa, and temperatures up to 1100 degrees Celcius. The thermodynamic and transport properties that can be generated from this routine are: density, specific volume, enthalpy, specific heat capacity, conductivity, viscosity, surface tension and Prandtle number for saturated liquid, saturated vapour, subcooled liquid of superheated vapour. A software for predicting flow regime has also been developed. It determines the flow pattern at specific flow conditions, and provides a correction factor for calculating the CHF during partially stratified horizontal flow. The technical bases for the program and its structure

  4. Office for Analysis and Evaluation of Operational Data 1992 annual report: Power reactors. Volume 7, No. 1

    Energy Technology Data Exchange (ETDEWEB)

    None

    1993-07-01

    The annual report of the US Nuclear Regulatory Commission`s Office for Analysis and Evaluation of Operational Data (AEOD) is devoted to the activities performed during 1992. The report is published in two separate parts. NUREG-1272, Vol. 7, No. 1, covers power reactors and presents an overview of the operating experience of the nuclear power industry from the NRC perspective, including comments about the trends of some key performance, measures. The report also includes the principal findings and issues identified in AEOD studies over the past year, and summarizes information from such sources as licensee event report% diagnostic evaluations, and reports to the NRC`s Operations Center. The reports contain a discussion of the Incident Investigation Team program and summarize the Incident Investigation Team and Augmented Inspection Team reports for that group of licensees. NUREG-1272, Vol. 7, No. 2, covers nonreactors and presents a review of the events and concerns during 1992 associated with the use of licensed material in nonreactor applications, such as personnel overexposures and medical misadministrations. Each volume contains a list of the AEOD reports issued for 1984--1992.

  5. Structural imaging of the brain reveals decreased total brain and total gray matter volumes in obese but not in lean women with polycystic ovary syndrome compared to body mass index-matched counterparts.

    Science.gov (United States)

    Ozgen Saydam, Basak; Has, Arzu Ceylan; Bozdag, Gurkan; Oguz, Kader Karli; Yildiz, Bulent Okan

    2017-07-01

    To detect differences in global brain volumes and identify relations between brain volume and appetite-related hormones in women with polycystic ovary syndrome (PCOS) compared to body mass index-matched controls. Forty subjects participated in this study. Cranial magnetic resonance imaging and measurements of fasting ghrelin, leptin and glucagon-like peptide 1 (GLP-1), as well as GLP-1 levels during mixed-meal tolerance test (MTT), were performed. Total brain volume and total gray matter volume (GMV) were decreased in obese PCOS compared to obese controls (p lean PCOS and controls did not show a significant difference. Secondary analyses of regional brain volumes showed decreases in GMV of the caudate nucleus, ventral diencephalon and hippocampus in obese PCOS compared to obese controls (p lean patients with PCOS had lower GMV in the amygdala than lean controls (p PCOS, suggests volumetric reductions in global brain areas in obese women with PCOS. Functional studies with larger sample size are needed to determine physiopathological roles of these changes and potential effects of long-term medical management on brain structure of PCOS.

  6. Association of translocator protein total distribution volume with duration of untreated major depressive disorder: a cross-sectional study.

    Science.gov (United States)

    Setiawan, Elaine; Attwells, Sophia; Wilson, Alan A; Mizrahi, Romina; Rusjan, Pablo M; Miler, Laura; Xu, Cynthia; Sharma, Sarita; Kish, Stephen; Houle, Sylvain; Meyer, Jeffrey H

    2018-04-01

    People with major depressive disorder frequently exhibit increasing persistence of major depressive episodes. However, evidence for neuroprogression (ie, increasing brain pathology with longer duration of illness) is scarce. Microglial activation, which is an important component of neuroinflammation, is implicated in neuroprogression. We examined the relationship of translocator protein (TSPO) total distribution volume (V T ), a marker of microglial activation, with duration of untreated major depressive disorder, and with total illness duration and antidepressant exposure. In this cross-sectional study, we recruited participants aged 18-75 years from the Toronto area and the Centre for Addiction and Mental Health (Toronto, ON, Canada). Participants either had major depressive episodes secondary to major depressive disorder or were healthy, as confirmed with a structured clinical interview and consultation with a study psychiatrist. To be enrolled, participants with major depressive episodes had to score a minimum of 17 on the 17-item Hamilton Depression Rating Scale, and had to be medication free or taking a stable dose of medication for at least 4 weeks before PET scanning. Eligible participants were non-smokers; had no history of or concurrent alcohol or substance dependence, neurological illness, autoimmune disorder, or severe medical problems; and were free from acute medical illnesses for the previous 2 weeks before PET scanning. Participants were excluded if they had used brain stimulation treatments within the 6 months before scanning, had used anti-inflammatory drugs lasting at least 1 week within the past month, were taking hormone replacement therapy, had psychotic symptoms, had bipolar disorder (type I or II) or borderline antisocial personality disorder, or were pregnant or breastfeeding. We scanned three primary grey-matter regions of interest (prefrontal cortex, anterior cingulate cortex, and insula) and 12 additional regions and subregions using 18

  7. Draft environmental impact statement for the siting, construction, and operation of New Production Reactor capacity. Volume 2, Sections 1-6

    Energy Technology Data Exchange (ETDEWEB)

    1991-04-01

    This (EIS) assesses the potential environmental impacts, both on a broad programmatic level and on a project-specific level, concerning a proposed action to provide new tritium production capacity to meet the nation`s nuclear defense requirements well into the 21st century. A capacity equivalent to that of about a 3,000-megawatt (thermal) heavy-water reactor was assumed as a reference basis for analysis in this EIS; this is the approximate capacity of the existing production reactors at DOE`s Savannah River Site. The EIS programmatic alternatives address Departmental decisions to be made on whether to build new production facilities, whether to build one or more complexes, what size production capacity to provide, and when to provide this capacity. Project-specific impacts for siting, constructing, and operating new production reactor capacity are assessed for three alternative sites: the Hanford Site near Richland, Washington; the Idaho National Engineering Laboratory near Idaho Falls, Idaho; and the Savannah River Site. For each site, the impacts of three reactor technologies (and supporting facilities) are assessed: a heavy-water reactor, a light-water reactor, and a modular high-temperature gas-cooled reactor. Impacts of the no-action alternative also are assessed. The EIS evaluates impacts related to air quality; noise levels; surface water, groundwater, and wetlands; land use; recreation; visual environment; biotic resources; historical, archaeological, and cultural resources; socioeconomics; transportation; waste management; and human health and safety. The EIS describes in detail the potential radioactive releases from new production reactors and support facilities and assesses the potential doses to workers and the general public. This volume contains the analysis of programmatic alternatives, project alternatives, affected environment of alternative sites, environmental consequences, and environmental regulations and permit requirements.

  8. Draft environmental impact statement for the siting, construction, and operation of New Production Reactor capacity. Volume 3, Sections 7-12, Appendices A-C

    Energy Technology Data Exchange (ETDEWEB)

    1991-04-01

    This Environmental Impact Statement (EIS) assesses the potential environmental impacts, both on a broad programmatic level and on a project-specific level, concerning a proposed action to provide new tritium production capacity to meet the nation`s nuclear defense requirements well into the 21st century. A capacity equivalent to that of about a 3,000-megawatt (thermal) heavy-water reactor was assumed as a reference basis for analysis in this EIS; this is the approximate capacity of the existing production reactors at DOE`s Savannah River Site near Aiken, South Carolina. The EIS programmatic alternatives address Departmental decisions to be made on whether to build new production facilities, whether to build one or more complexes, what size production capacity to provide, and when to provide this capacity. Project-specific impacts for siting, constructing, and operating new production reactor capacity are assessed for three alternative sites: the Hanford Site near Richland, Washington; the Idaho National Engineering Laboratory near Idaho Falls, Idaho; and the Savannah River Site. For each site, the impacts of three reactor technologies (and supporting facilities) are assessed: a heavy-water reactor, a light-water reactor, and a modular high-temperature gas-cooled reactor. Impacts of the no-action alternative also are assessed. The EIS evaluates impacts related to air quality; noise levels; surface water, groundwater, and wetlands; land use; recreation; visual environment; biotic resources; historical, archaeological, and cultural resources; socioeconomics; transportation; waste management; and human health and safety. The EIS describes in detail the potential radioactive releases from new production reactors and support facilities and assesses the potential doses to workers and the general public. This volume contains references; a list of preparers and recipients; acronyms, abbreviations, and units of measure; a glossary; an index and three appendices.

  9. ASAMPSA2 best-practices guidelines for L2 PSA development and applications. Volume 3 - Extension to Gen IV reactors

    International Nuclear Information System (INIS)

    Bassi, C.; Bonneville, H.; Brinkman, H.; Burgazzi, L.; Polidoro, F.; Vincon, L.; Jouve, S.

    2010-01-01

    The main objective assigned to the Work Package 4 (WP4) of the 'ASAMPSA2' project (EC 7. FPRD) consist in the verification of the potential compliance of L2PSA guidelines based on PWR/BWR reactors (which are specific tasks of WP2 and WP3) with Generation IV representative concepts. Therefore, in order to exhibit potential discrepancies between LWRs and new reactor types, the following work was based on the up-to-date designs of: - The European Fast Reactor (EFR) which will be considered as prototypical of a pool-type Sodium-cooled Fast Reactor (SFR); - The ELSY design for the Lead-cooled Fast Reactor (LFR) technology; - The ANTARES project which could be representative of a Very-High Temperature Reactor (VHTR); - The CEA 2400 MWth Gas-cooled Fast Reactor (GFR). (authors)

  10. Compilation of reports of the Advisory Committee on Reactor Safeguards, 1957-1984. Volume 6. Generic Subjects R-Z, Appendices

    International Nuclear Information System (INIS)

    1985-04-01

    This six-volume compilation contains over 1000 reports prepared by the Advisory Committee on Reactor Safeguards from September 1957 through December 1984. The reports are divided into two groups: Part 1: ACRS Reports on Project Reviews, and Part 2: ACRS Reports on Generic Subjects. Part 1 contains ACRS reports alphabetized by project name and within project name by chronological order. Part 2 categorizes the reports by the most appropriate generic subject area and within subject area by chronological order. This volume contains the generic subjects with an alphabetical listing from R to Z

  11. Twenty-third water reactor safety information meeting. Volume 3, structural and seismic engineering, primary systems integrity, equipment operability and aging, ECCS strainer blockage research and regulatory issues

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1996-03-01

    This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 3, presents topics in Structural & Seismic Engineering, Primary Systems Integrity, Equipment Operability and Aging, and ECCS Strainer Blockage Research & Regulatory Issues. Individual papers have been cataloged separately.

  12. Twenty-third water reactor safety information meeting. Volume 3, structural and seismic engineering, primary systems integrity, equipment operability and aging, ECCS strainer blockage research and regulatory issues

    International Nuclear Information System (INIS)

    Monteleone, S.

    1996-03-01

    This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 3, presents topics in Structural ampersand Seismic Engineering, Primary Systems Integrity, Equipment Operability and Aging, and ECCS Strainer Blockage Research ampersand Regulatory Issues. Individual papers have been cataloged separately

  13. Diversity of total and functional microbiome of anammox reactors fed with complex and synthetic nitrogen-rich wastewaters

    DEFF Research Database (Denmark)

    Gülay, Arda; Pellicer i Nàcher, Carles; Mutlu, Ayten Gizem

    diversity than the bioreactors treating synthetic wastewaters inferred from observed OTUs0.03, Chao1, Shannon index and Phylogenetic distance calculations. Differences in total microbial diversity agreed with the ecological theory concerning the positive correlation between substrate complexity...

  14. Increased Severe Trauma Patient Volume is Associated With Survival Benefit and Reduced Total Health Care Costs: A Retrospective Observational Study Using a Japanese Nationwide Administrative Database.

    Science.gov (United States)

    Endo, Akira; Shiraishi, Atsushi; Fushimi, Kiyohide; Murata, Kiyoshi; Otomo, Yasuhiro

    2017-06-07

    The aim of this study was to evaluate the associations of severe trauma patient volume with survival benefit and health care costs. The effect of trauma patient volume on survival benefit is inconclusive, and reports on its effects on health care costs are scarce. We conducted a retrospective observational study, including trauma patients who were transferred to government-approved tertiary emergency hospitals, or hospitals with an intensive care unit that provided an equivalent quality of care, using a Japanese nationwide administrative database. We categorized hospitals according to their annual severe trauma patient volumes [1 to 50 (reference), 51 to 100, 101 to 150, 151 to 200, and ≥201]. We evaluated the associations of volume categories with in-hospital survival and total cost per admission using a mixed-effects model adjusting for patient severity and hospital characteristics. A total of 116,329 patients from 559 hospitals were analyzed. Significantly increased in-hospital survival rates were observed in the second, third, fourth, and highest volume categories compared with the reference category [94.2% in the highest volume category vs 88.8% in the reference category, adjusted odds ratio (95% confidence interval, 95% CI) = 1.75 (1.49-2.07)]. Furthermore, significantly lower costs (in US dollars) were observed in the second and fourth categories [mean (standard deviation) for fourth vs reference = $17,800 ($17,378) vs $20,540 ($32,412), adjusted difference (95% CI) = -$2559 (-$3896 to -$1221)]. Hospitals with high volumes of severe trauma patients were significantly associated with a survival benefit and lower total cost per admission.

  15. Report of the Survey on the Design Review of New Reactor Applications. Volume 1 - Instrumentation and Control

    International Nuclear Information System (INIS)

    Downey, Steven

    2014-06-01

    At the tenth meeting of the CNRA Working Group on the Regulation of New Reactors (WGRNR) in March 2013, the members agreed to present the responses to the Second Phase, or Design Phase, of the Licensing Process Survey as a multi-volume text. As such, each report will focus on one of the eleven general technical categories covered in the survey. The general technical categories were selected to conform to the topics covered in the International Atomic Energy Agency (IAEA) Safety Guide GS-G-4.1. This report, which is the first volume, provides a discussion of the survey responses related to Instrumentation and Control (I and C). The Instrumentation and Control category includes the twelve following technical topics: Reactor trip system, actuation systems for Engineered Safety Features (ESF), safe shutdown system, safety-related display instrumentation, information and interlock systems important to safety, controls systems, main control room, supplementary control room, diverse I and C systems, data communication systems, software reliability and cyber-security. For each technical topic, the member countries described the information provided by the applicant, the scope and level of detail of the technical review, the technical basis for granting regulatory authorisation, the skill sets required and the Level of effort needed to perform the review. Based on a comparison of the information provided in response to the survey, the following observations were made: - Among the regulatory organisations that responded to the survey, there are similarities in the design information provided by an applicant. In most countries, the design information provided by an applicant includes, but is not limited to, a description of the I and C system design and functions, a description of the verification and validation programmes, and provisions for analysis, testing, and inspection of various I and C systems. - In addition to the regulations, it is a common practice for countries

  16. Individual plant examination program: Perspectives on reactor safety and plant performance. Parts 2--5: Final report; Volume 2

    International Nuclear Information System (INIS)

    1997-12-01

    This report provides perspectives gained by reviewing 75 Individual Plant Examination (IPE) submittals pertaining to 108 nuclear power plant units. IPEs are probabilistic analyses that estimate the core damage frequency (CDF) and containment performance for accidents initiated by internal events. The US Nuclear Regulatory Commission (NRC) reviewed the IPE submittals with the objective of gaining perspectives in three major areas: (1) improvements made to individual plants as a result of their IPEs and the collective results of the IPE program, (2) plant-specific design and operational features and modeling assumptions that significantly affect the estimates of CDF and containment performance, and (3) strengths and weaknesses of the models and methods used in the IPEs. These perspectives are gained by assessing the core damage and containment performance results, including overall CDF, accident sequences, dominant contributions to component failure and human error, and containment failure modes. Methods, data, boundary conditions, and assumptions used in the IPEs are considered in understanding the differences and similarities observed among the various types of plants. This report is divided into three volumes containing six parts. Part 1 is a summary report of the key perspectives gained in each of the areas identified above, with a discussion of the NRC's overall conclusions and observations. Part 2 discusses key perspectives regarding the impact of the IPE Program on reactor safety. Part 3 discusses perspectives gained from the IPE results regarding CDF, containment performance, and human actions. Part 4 discusses perspectives regarding the IPE models and methods. Part 5 discusses additional IPE perspectives. Part 6 contains Appendices A, B and C which provide the references of the information from the IPEs, updated PRA results, and public comments on draft NUREG-1560 respectively

  17. Individual plant examination program: Perspectives on reactor safety and plant performance. Part 1: Final summary report; Volume 1

    International Nuclear Information System (INIS)

    1997-12-01

    This report provides perspectives gained by reviewing 75 Individual Plant Examination (IPE) submittals pertaining to 108 nuclear power plant units. IPEs are probabilistic analyses that estimate the core damage frequency (CDF) and containment performance for accidents initiated by internal events. The US Nuclear Regulatory Commission (NRC) reviewed the IPE submittals with the objective of gaining perspectives in three major areas: (1) improvements made to individual plants as a result of their IPEs and the collective results of the IPE program, (2) plant-specific design and operational features and modeling assumptions that significantly affect the estimates of CDF and containment performance, and (3) strengths and weaknesses of the models and methods used in the IPEs. These perspectives are gained by assessing the core damage and containment performance results, including overall CDF, accident sequences, dominant contributions to component failure and human error, and containment failure modes. Methods, data, boundary conditions, and assumptions used in the IPEs are considered in understanding the differences and similarities observed among the various types of plants. This report is divided into three volumes containing six parts. Part 1 is a summary report of the key perspectives gained in each of the areas identified above, with a discussion of the NRC's overall conclusions and observations. Part 2 discusses key perspectives regarding the impact of the IPE Program on reactor safety. Part 3 discusses perspectives gained from the IPE results regarding CDF, containment performance, and human actions. Part 4 discusses perspectives regarding the IPE models and methods. Part 5 discusses additional IPE perspectives. Part 6 contains Appendices A, B and C which provide the references of the information from the IPEs, updated PRA results, and public comments on draft NUREG-1560 respectively

  18. Individual plant examination program: Perspectives on reactor safety and plant performance. Parts 2--5: Final report; Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    This report provides perspectives gained by reviewing 75 Individual Plant Examination (IPE) submittals pertaining to 108 nuclear power plant units. IPEs are probabilistic analyses that estimate the core damage frequency (CDF) and containment performance for accidents initiated by internal events. The US Nuclear Regulatory Commission (NRC) reviewed the IPE submittals with the objective of gaining perspectives in three major areas: (1) improvements made to individual plants as a result of their IPEs and the collective results of the IPE program, (2) plant-specific design and operational features and modeling assumptions that significantly affect the estimates of CDF and containment performance, and (3) strengths and weaknesses of the models and methods used in the IPEs. These perspectives are gained by assessing the core damage and containment performance results, including overall CDF, accident sequences, dominant contributions to component failure and human error, and containment failure modes. Methods, data, boundary conditions, and assumptions used in the IPEs are considered in understanding the differences and similarities observed among the various types of plants. This report is divided into three volumes containing six parts. Part 1 is a summary report of the key perspectives gained in each of the areas identified above, with a discussion of the NRC`s overall conclusions and observations. Part 2 discusses key perspectives regarding the impact of the IPE Program on reactor safety. Part 3 discusses perspectives gained from the IPE results regarding CDF, containment performance, and human actions. Part 4 discusses perspectives regarding the IPE models and methods. Part 5 discusses additional IPE perspectives. Part 6 contains Appendices A, B and C which provide the references of the information from the IPEs, updated PRA results, and public comments on draft NUREG-1560 respectively.

  19. Draft Environmental Impact Statement on a proposed nuclear weapons nonproliferation policy concerning foreign research reactor spent nuclear fuel. Volume 1

    International Nuclear Information System (INIS)

    1995-03-01

    The United States Department of Energy and United States Department of State are jointly proposing to adopt a policy to manage spent nuclear fuel from foreign research reactors. Only spent nuclear fuel containing uranium enriched in the United States would be covered by the proposed policy. The purpose of the proposed policy is to promote U.S. nuclear weapons nonproliferation policy objectives, specifically by seeking to reduce highly-enriched uranium from civilian commerce. Environmental effects and policy considerations of three Management Alternative approaches for implementation of the proposed policy are assessed. The three Management Alternatives analyzed are: (1) acceptance and management of the spent nuclear fuel by the Department of Energy in the United States, (2) management of the spent nuclear fuel at one or more foreign facilities (under conditions that satisfy United States nuclear weapons nonproliferation policy objectives), and (3) a combination of components of Management Alternatives 1 and 2 (Hybrid Alternative). A No Action Alternative is also analyzed. For each Management Alternative, there are a number of alternatives for its implementation. For Management Alternative 1, this document addresses the environmental effects of various implementation alternatives such as varied policy durations, management of various quantities of spent nuclear fuel, and differing financing arrangements. Environmental impacts at various potential ports of entry, along truck and rail transportation routes, at candidate management sites, and for alternate storage technologies are also examined. For Management Alternative 2, this document addresses two subalternatives: (1) assisting foreign nations with storage; and (2) assisting foreign nations with reprocessing of the spent nuclear fuel. Management Alternative 3 analyzes a hybrid alternative. This document is Vol. 1 of 2 plus summary volume

  20. Assessment of United States industry structural codes and standards for application to advanced nuclear power reactors: Appendices. Volume 2

    International Nuclear Information System (INIS)

    Adams, T.M.; Stevenson, J.D.

    1995-10-01

    Throughout its history, the USNRC has remained committed to the use of industry consensus standards for the design, construction, and licensing of commercial nuclear power facilities. The existing industry standards are based on the current class of light water reactors and as such may not adequately address design and construction features of the next generation of Advanced Light Water Reactors and other types of Advanced Reactors. As part of their on-going commitment to industry standards, the USNRC commissioned this study to evaluate US industry structural standards for application to Advanced Light Water Reactors and Advanced Reactors. The initial review effort included (1) the review and study of the relevant reactor design basis documentation for eight Advanced Light Water Reactors and Advanced Reactor Designs, (2) the review of the USNRCs design requirements for advanced reactors, (3) the review of the latest revisions of the relevant industry consensus structural standards, and (4) the identification of the need for changes to these standards. The results of these studies were used to develop recommended changes to industry consensus structural standards which will be used in the construction of Advanced Light Water Reactors and Advanced Reactors. Over seventy sets of proposed standard changes were recommended and the need for the development of four new structural standards was identified. In addition to the recommended standard changes, several other sets of information and data were extracted for use by USNRC in other on-going programs. This information included (1) detailed observations on the response of structures and distribution system supports to the recent Northridge, California (1994) and Kobe, Japan (1995) earthquakes, (2) comparison of versions of certain standards cited in the standard review plan to the most current versions, and (3) comparison of the seismic and wind design basis for all the subject reactor designs

  1. Determination of the total blood volume of the rat using chromium 51 (1962); Determination du volume sanguin total chez le rat a l'aide du chrome 51 (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, Y; Rinaldi, R [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1962-07-01

    In various experiments it is important that the rat's blood volume should be known and it is essential that it could be worked out by simply weighing the animal. Therefore authors decided to work out systematically with chromium 51 blood volume for rats with different weights. Results have shown that blood volume for 100 g is 5.55 ml and that it does not seen to vary with the weight of animals. (authors) [French] La connaissance de la masse sanguine chez le rat est indispensable dans de nombreuses experiences et il est essentiel de pouvoir l'evaluer apres une simple pesee de l'animal. C'est dans ce but que les auteurs ont entrepris la determination systematique de la masse sanguine a l'aide du chrome 51 chez des rats de poids differents. Les resultats obtenus ont montre que la masse sanguine rapportee a 100 grammes d'animal est de 5,53 millilitres, et qu'elle ne parait pas varier avec le poids de l'animal. (auteurs)

  2. High-volume infiltration analgesia in total knee arthroplasty: a randomized, double-blind, placebo-controlled trial

    DEFF Research Database (Denmark)

    Andersen, L.O.; Husted, H.; Otte, K.S.

    2008-01-01

    with a detailed description of the infiltration technique. METHODS: In a randomized, double-blind, placebo-controlled trial in 12 patients undergoing bilateral knee arthroplasty, saline or high-volume (170 ml) ropivacaine (0.2%) with epinephrine was infiltrated around each knee, with repeated doses administered...

  3. Nuclear Reactor RA Safety Report, Vol. 4, Reactor

    International Nuclear Information System (INIS)

    1986-11-01

    RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices [sr

  4. Computational methodology for the Oak Ridge Research Reactor (ORR) and Bulk Shielding Reactor (BSR): cross-section and validation. Volume 1

    International Nuclear Information System (INIS)

    Miller, L.F.; Williams, M.L.

    1986-03-01

    A neutronics library suitable for low-enrichment uranium (LEU) and high-enrichment uranium (HEU) fueled cores for both the Oak Ridge Research Reactor (ORR) and the Bulk Shielding Reactor (BSR) is documented herein. The library is obtained from version V of the Evaluated Nuclear Data File (ENDF/B-V) and contains 223 nuclides weighted over a variety of region-dependent neutron spectra. Self-shielding and zone-weighting effects are incorporated with 227-group calculations for several reactor-core configurations. Libraries are archived for both transport and diffusion theory seven-group calculations. Complete listings of processing details are included so that libraries with different specifications can be easily obtained. Results from validation calculations indicate that the neutronics libraries obtained from this effort are suitable for neutronics computations for the ORR and BSR. 12 refs., 5 figs., 15 tabs

  5. The effect of post-wash total progressive motile sperm count and semen volume on pregnancy outcomes in intrauterine insemination cycles: a retrospective study.

    Science.gov (United States)

    Ok, Elvan Koyun; Doğan, Omer Erbil; Okyay, Recep Emre; Gülekli, Bülent

    2013-01-01

    The purpose of this study was to determine the impact of post-wash total progressive motile sperm count (TPMSC) and semen volume on pregnancy outcomes in intrauterine insemination (IUI) cycles. The retrospective study included a total of 156 cycles (141 couples) and was performed in our center over a 24-month period. The semen parameters were recorded for each man and each insemination. The semen samples were re-evaluated after the preparation process. Post-wash TPMSC values were divided into four groups; Group 1: 10×10(6). Post-wash inseminated semen volume was divided into three groups; Group 1: 0.3 mL; Group 2: 0.4 mL; Group 3: 0.5 mL. The effect of post-wash total progressive motile sperm and semen volume on pregnancy outcomes was evaluated. The pregnancy rates per cycle and per couple were 27.56% and 30.49%, respectively. There was not a significant relationship between the inseminated semen volume and pregnancy rate (p>0.05). However, a significant linear-by-linear association was documented between the TPMSC and pregnancy rate (p=0.042). Our findings suggest that the post-wash inseminated semen volume should be between 0.3-0.5 mL. An average post-wash total motile sperm count of 10×10(6) may be a useful threshold value for IUI success, but more studies are needed to determine a cut-off value for TPMSC.

  6. Research reactors

    International Nuclear Information System (INIS)

    Merchie, Francois

    2015-10-01

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

  7. Safety report concerning the reactor Pegase - volume 1 - Description of the installation - volume 2 - Safety of the installations; Rapport de surete du reacteur pegase - tome 1 - Description des installations - tome 2 - Surete des installations

    Energy Technology Data Exchange (ETDEWEB)

    Lacour, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Legoin, P. [S.E.M. Hispano-Suiza, 92 - Colombes (France)

    1964-07-01

    In the first volume: This report is a description of the reactor Pegase, given with a view to examine the safety of the installations. The Cadarache site at which they are situated is briefly described, in particular because of the consequences on the techniques employed for building Pegase. A description is also given of the original aspects of the reactor. The independent loops which are designed for full-scale testing of fuel elements used in natural uranium-gas-graphite reactor systems are described in this report, together with their operational and control equipment. In the second volume: In the present report are examined the accidents which could cause damage to the Pegase reactor installation. Among possible causes of accidents considered are the seismicity of the region, an excessive power excursion of the reactor and a fracture in the sealing of an independent loop. Although all possible precautions have been taken to offset the effects of such accidents, their ultimate consequences are considered here. The importance is stressed of the security action and regulations which, added to the precautions taken for the construction, ensure the safety of the installations. (authors) [French] Dans le volume 1: Ce rapport est une description du reacteur Pegase, afin d'examiner la surete des installations. Le site de CADARACHE ou elles sont situees, a ete sommairement decrit, en particulier, a cause des consequences sur les techniques mises en oeuvre pour la realisation de Pegase. Nous nous sommes egalement attache a decrire les aspects originaux du reacteur. Les boucles autonomes destinees a tester en vraie grandeur des elements combustibles de la filiere uranium naturel graphite-gaz, ainsi que leurs dispositifs de controle et d'exploitation, figurent egalement dans ce rapport. Dans le volume 2: Dans le present rapport, nous examinons des accidents pouvant endommager des installations du reacteur Pegase. Les origines d'accidents examines

  8. Safety report concerning the reactor Pegase - volume 1 - Description of the installation - volume 2 - Safety of the installations; Rapport de surete du reacteur pegase - tome 1 - Description des installations - tome 2 - Surete des installations

    Energy Technology Data Exchange (ETDEWEB)

    Lacour, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Legoin, P [S.E.M. Hispano-Suiza, 92 - Colombes (France)

    1964-07-01

    In the first volume: This report is a description of the reactor Pegase, given with a view to examine the safety of the installations. The Cadarache site at which they are situated is briefly described, in particular because of the consequences on the techniques employed for building Pegase. A description is also given of the original aspects of the reactor. The independent loops which are designed for full-scale testing of fuel elements used in natural uranium-gas-graphite reactor systems are described in this report, together with their operational and control equipment. In the second volume: In the present report are examined the accidents which could cause damage to the Pegase reactor installation. Among possible causes of accidents considered are the seismicity of the region, an excessive power excursion of the reactor and a fracture in the sealing of an independent loop. Although all possible precautions have been taken to offset the effects of such accidents, their ultimate consequences are considered here. The importance is stressed of the security action and regulations which, added to the precautions taken for the construction, ensure the safety of the installations. (authors) [French] Dans le volume 1: Ce rapport est une description du reacteur Pegase, afin d'examiner la surete des installations. Le site de CADARACHE ou elles sont situees, a ete sommairement decrit, en particulier, a cause des consequences sur les techniques mises en oeuvre pour la realisation de Pegase. Nous nous sommes egalement attache a decrire les aspects originaux du reacteur. Les boucles autonomes destinees a tester en vraie grandeur des elements combustibles de la filiere uranium naturel graphite-gaz, ainsi que leurs dispositifs de controle et d'exploitation, figurent egalement dans ce rapport. Dans le volume 2: Dans le present rapport, nous examinons des accidents pouvant endommager des installations du reacteur Pegase. Les origines d'accidents examines comprennent la seismicite

  9. SCDAP/RELAP5/MOD 3.1 code manual: MATPRO, A library of materials properties for Light-Water-Reactor accident analysis. Volume 4

    International Nuclear Information System (INIS)

    Hagrman, D.T.

    1995-06-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light -- water-reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume, Volume IV, describes the material properties correlations and computer subroutines (MATPRO) used by SCDAP/RELAP5. formulation of the materials properties are generally semi-empirical in nature. The materials property subroutines contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, cadmium, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, fill gas mixtures, carbon steel, and tungsten. This document also contains descriptions of the reaction and solution rate models needed to analyze a reactor accident

  10. Nuclear reactor

    International Nuclear Information System (INIS)

    Garabedian, G.

    1988-01-01

    A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

  11. IMMUNOMETABOLIC RESPONSES AFTER SHORT AND MODERATE REST INTERVALS TO STRENGTH EXERCISE WITH AND WITHOUT SIMILAR TOTAL VOLUME.

    Directory of Open Access Journals (Sweden)

    Ricardo Agostinete

    2016-10-01

    Full Text Available This study investigated the influence of short and moderate intervals of recovery with and without equated volume during an acute bout exhaustive strength exercise on metabolic, hormonal and inflammatory responses in healthy adults. Eight physically active men (23.5 ±3.1 performed three randomized sequences: Short (70% of 1RM with 30 seconds of rest; Moderate (70% of 1RM with 90 seconds of rest; and Volume-Equated Short (70% of 1 RM with 30 seconds of rest between sets with a repetition volume equal to that performed in Moderate. All sequences of exercises were performed until movement failure in the squat, bench press and T-bar row exercises, respectively. Glucose, lactate, testosterone, IL-6, IL-10, IL-1ra and MCP-1 levels were assessed at rest, immediate post-exercise, and 1 hour post. There was a main effect of time for testosterone (p<0.001. The post hoc indicated differences between post-exercise and rest and post-1 hour and post-exercise (p<0.001. Lactate increased post-exercise when compared to pre and post-1 hour (p<0.001 and maintained higher post-1 hour in relation to rest. IL-6 was greater post-exercise than rest (p= 0.045 and post-1 hour and rest (p= 0.020. IL-10 was greater post-exercise (p= 0.007 and post-1 hour (p=0.002 than rest. IL-1ra increased post-exercise in relation to rest (p=0.003 and MCP-1 was greater post-exercise than rest (p<0.001 and post-1 hour (p=0.043. There were no significant differences between conditions or interaction. Thus, both short and moderate intervals of recovery induced greater metabolic, hormonal and inflammatory responses after acute bout of exhaustive strength exercise in healthy adult.

  12. Weapons-grade plutonium dispositioning. Volume 3: A new reactor concept without uranium or thorium for burning weapons-grade plutonium

    International Nuclear Information System (INIS)

    Ryskamp, J.M.; Schnitzler, B.G.; Fletcher, C.D.

    1993-06-01

    The National Academy of Sciences (NAS) requested that the Idaho National Engineering Laboratory (INEL) examine concepts that focus only on the destruction of 50,000 kg of weapons-grade plutonium. A concept has been developed by the INEL for a low-temperature, low-pressure, low-power density, low-coolant-flow-rate light water reactor that destroys plutonium quickly without using uranium or thorium. This concept is very safe and could be designed, constructed, and operated in a reasonable time frame. This concept does not produce electricity. Not considering other missions frees the design from the paradigms and constraints used by proponents of other dispositioning concepts. The plutonium destruction design goal is most easily achievable with a large, moderate power reactor that operates at a significantly lower thermal power density than is appropriate for reactors with multiple design goals. This volume presents the assumptions and requirements, a reactor concept overview, and a list of recommendations. The appendices contain detailed discussions on plutonium dispositioning, self-protection, fuel types, neutronics, thermal hydraulics, off-site radiation releases, and economics

  13. Energy Research Advisory Board, Civilian Nuclear Power Panel: Subpanel 1 report, Light water reactor utilization and improvement: Volume 2

    International Nuclear Information System (INIS)

    1986-10-01

    The Secretary of Energy requested that the Office of Nuclear Energy prepare a strategic national plan that outlines the Department's role in the future development of civilian nuclear power and that the Energy Research Advisory Board establish an ad hoc panel to review and comment on this plan. The Energy Research Advisory Board formed a panel for this review and three subpanels were formed. One subpanel was formed to address the institutional issues surrounding nuclear power, one on research and development for advanced nuclear power plants and a third subpanel on light water reactor utilization and improvement. The subpanel on light water reactors held two meetings at which representatives of the DOE, the NRC, EPRI, industry and academic groups made presentations. This is the report of the subpanel on light water reactor utilization and improvement. This report presents the subpanel's assessment of initiatives which the Department of Energy should undertake in the national interest, to develop and support light water reactor technologies

  14. Evaluation of apoptosis and micronucleation induced by reactor neutron beams with two different cadmium ratios in total and quiescent cell populations within solid tumors

    International Nuclear Information System (INIS)

    Masunaga, Shin-ichiro; Ono, Koji; Sakurai, Yoshinori; Takagaki, Masao; Kobayashi, Tooru; Kinashi, Yuko; Suzuki, Minoru

    2001-01-01

    Purpose: Response of quiescent (Q) and total tumor cells in solid tumors to reactor neutron beam irradiation with two different cadmium (Cd) ratios was examined in terms of micronucleus (MN) frequency and apoptosis frequency, using four different tumor cell lines. Methods and Materials: C57BL mice bearing EL4 tumors, C3H/He mice bearing SCC VII or FM3A tumors, and Balb/c mice bearing EMT6/KU tumors received 5-bromo-2'-deoxyuridine (BrdU) continuously for 5 days via implanted mini-osmotic pumps to label all proliferating (P) cells. Thirty min after i.p. injection of sodium borocaptate- 10 B (BSH), or 3 h after oral administration of p-boronophenylalanine- 10 B (BPA), the tumors were irradiated with neutron beams. The tumors without 10 B-compound administration were irradiated with neutron beams or γ-rays. This neutron beam irradiation was performed using neutrons with two different Cd ratios. The tumors were then excised, minced, and trypsinized. The tumor cell suspensions thus obtained were incubated with cytochalasin-B (a cytokinesis blocker), and the MN frequency in cells without BrdU labeling (=Q cells) was determined using immunofluorescence staining for BrdU. Meanwhile, for apoptosis assay, 6 h after irradiation, tumor cell suspensions obtained in the same manner were fixed, and the apoptosis frequency in Q cells was also determined with immunofluorescence staining for BrdU. The MN and apoptosis frequencies in total (P+Q) tumor cells were determined from the tumors that were not pretreated with BrdU. Results: Without 10 B-compounds, the sensitivity difference between total and Q cells was reduced by neutron beam irradiation. Under our particular neutron beam irradiation condition, relative biological effectiveness (RBE) of neutrons was larger in Q cells than in total cells, and the RBE values were larger for low Cd-ratio than high Cd-ratio neutrons. With 10 B-compounds, both frequencies were increased for each cell population, especially for total cells. BPA

  15. Volume totalizers analysis of pipelines operated by TRANSPETRO National Operational Control Center; Analise de totalizadores de volume em oleodutos operados pelo Centro Nacional de Controle e Operacao da TRANSPETRO

    Energy Technology Data Exchange (ETDEWEB)

    Aramaki, Thiago Lessa; Montalvao, Antonio Filipe Falcao [Petrobras Transporte S.A. (TRANSPETRO), Rio de Janeiro, RJ (Brazil); Marques, Thais Carrijo [Universidade do Estado do Rio de Janeiro (UERJ), RJ (Brazil)

    2012-07-01

    This paper aims to present the results and methodology in the analysis of differences in volume totals used in systems such as batch tracking and leak detection of pipelines operated by the National Center for Operational Control (CNCO) at TRANSPETRO. In order to optimize this type of analysis, software was developed to acquisition and processing of historical data using the methodology developed. The methodology developed takes into account the particularities encountered in systems operated by TRANSPETRO, more specifically, by CNCO. (author)

  16. Deltoid muscle volume affects clinical outcome of reverse total shoulder arthroplasty in patients with cuff tear arthropathy or irreparable cuff tears.

    Directory of Open Access Journals (Sweden)

    Jong Pil Yoon

    Full Text Available We aimed to estimate the interrelation between preoperative deltoid muscle status by measuring the 3-dimensional deltoid muscle volume and postoperative functional outcomes after reverse total shoulder arthroplasty(RTSA. Thirty-five patients who underwent RTSA participated in this study. All patients underwent preoperative magnetic resonance imaging(MRI as well as pre- and postoperative radiography and various functional outcome evaluations at least 1 year. The primary outcome parameter was set as age- and sex-matched Constant scores. The 3-dimensional deltoid muscle model was generated using a medical image processing software and in-house code, and the deltoid muscle volume was calculated automatically. Various clinical and radiographic factors comprising the deltoid muscle volume adjusted for body mass index(BMI were analyzed, and their interrelation with the outcome parameters was appraised using a multivariate analysis. As a result, all practical consequences considerably improved following surgery(all p<0.01. Overall, 20 and 15 indicated a higher and a lower practical consequence than the average, respectively, which was assessed by the matched Constant scores. The deltoid muscle volume adjusted for BMI(p = 0.009, absence of a subscapularis complete tear (p = 0.040, and greater change in acromion-deltoid tuberosity distance(p = 0.013 were associated with higher matched Constant scores. Multivariate analysis indicated that the deltoid muscle volume was the single independent prognostic factor for practical consequences(p = 0.011. In conclusion, the preoperative deltoid muscle volume significantly affected the functional outcome following RTSA in patients with cuff tear arthropathy or irreparable cuff tears. Therefore, more attention should be paid to patients with severe atrophied deltoid muscle who are at a high risk for poor practical consequences subsequent to RTSA.

  17. Deltoid muscle volume affects clinical outcome of reverse total shoulder arthroplasty in patients with cuff tear arthropathy or irreparable cuff tears.

    Science.gov (United States)

    Yoon, Jong Pil; Seo, Anna; Kim, Jeong Jun; Lee, Chang-Hwa; Baek, Seung-Hun; Kim, Shin Yoon; Jeong, Eun Taek; Oh, Kyung-Soo; Chung, Seok Won

    2017-01-01

    We aimed to estimate the interrelation between preoperative deltoid muscle status by measuring the 3-dimensional deltoid muscle volume and postoperative functional outcomes after reverse total shoulder arthroplasty(RTSA). Thirty-five patients who underwent RTSA participated in this study. All patients underwent preoperative magnetic resonance imaging(MRI) as well as pre- and postoperative radiography and various functional outcome evaluations at least 1 year. The primary outcome parameter was set as age- and sex-matched Constant scores. The 3-dimensional deltoid muscle model was generated using a medical image processing software and in-house code, and the deltoid muscle volume was calculated automatically. Various clinical and radiographic factors comprising the deltoid muscle volume adjusted for body mass index(BMI) were analyzed, and their interrelation with the outcome parameters was appraised using a multivariate analysis. As a result, all practical consequences considerably improved following surgery(all pmuscle volume adjusted for BMI(p = 0.009), absence of a subscapularis complete tear (p = 0.040), and greater change in acromion-deltoid tuberosity distance(p = 0.013) were associated with higher matched Constant scores. Multivariate analysis indicated that the deltoid muscle volume was the single independent prognostic factor for practical consequences(p = 0.011). In conclusion, the preoperative deltoid muscle volume significantly affected the functional outcome following RTSA in patients with cuff tear arthropathy or irreparable cuff tears. Therefore, more attention should be paid to patients with severe atrophied deltoid muscle who are at a high risk for poor practical consequences subsequent to RTSA.

  18. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 2. Appendices. Technical report, September 1977-October 1979

    International Nuclear Information System (INIS)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE. This volume contains the appendices

  19. Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior

    International Nuclear Information System (INIS)

    Monteleone, S.

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting

  20. Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting.

  1. Bibliography, subject index, and author index of the literature examined by the radiation shielding information center. Volume 6. Reactor and weapons radiation shielding

    International Nuclear Information System (INIS)

    1980-05-01

    An indexed bibliography is presented of literature selected by the Radiation Shielding Information Center since the previous volume was published in 1978 in the area of radiation transport and shielding against radiation from nuclear reactors, x-ray machines, radioisotopes, nuclear weapons (including fallout), and low energy accelerators (e.g., neutron generators). The bibliography was typeset from data processed by computer from magnetic tape files. In addition to lists of literature titles by subject categories (accessions 4951-6200), an author index is given

  2. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 2. Appendices. Technical report, September 1977-October 1979

    Energy Technology Data Exchange (ETDEWEB)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE. This volume contains the appendices.

  3. Final Generic Environmental Impact Statement. Handling and storage of spent light water power reactor fuel. Volume 1. Executive summary and text

    International Nuclear Information System (INIS)

    1979-08-01

    The Generic Environmental Impact Statement on spent fuel storage was prepared by the Nuclear Regulatory Commission staff in response to a directive from the Commissioners published in the Federal Register, September 16, 1975 (40 FR 42801). The Commission directed the staff to analyze alternatives for the handling and storage of spent light water power reactor fuel with particular emphasis on developing long range policy. Accordingly, the scope of this statement examines alternative methods of spent fuel storage as well as the possible restriction or termination of the generation of spent fuel through nuclear power plant shutdown. Volume 1 includes the executive summary and the text

  4. RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume II. Program implementation

    International Nuclear Information System (INIS)

    1976-06-01

    A discussion is presented of the use of the RELAP4/MOD5 computer program in simulating the thermal-hydraulic behavior of light-water reactor systems when subjected to postulated transients such as a LOCA, pump failure, or nuclear excursion. The volume is divided into main sections which cover: (1) program description, (2) input data, (3) problem initialization, (4) user guidelines, (5) output discussion, (6) source program description, (7) implementation requirements, (8) data files, (9) description of PLOTR4M, (10) description of STH20, (11) summary flowchart, (12) sample problems, (13) problem definition, and (14) problem input

  5. Nuclear reactors. Introduction

    International Nuclear Information System (INIS)

    Boiron, P.

    1997-01-01

    This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)

  6. Predicting Stem Total and Assortment Volumes in an Industrial Pinus taeda L. Forest Plantation Using Airborne Laser Scanning Data and Random Forest

    Directory of Open Access Journals (Sweden)

    Carlos Alberto Silva

    2017-07-01

    Full Text Available Improvements in the management of pine plantations result in multiple industrial and environmental benefits. Remote sensing techniques can dramatically increase the efficiency of plantation management by reducing or replacing time-consuming field sampling. We tested the utility and accuracy of combining field and airborne lidar data with Random Forest, a supervised machine learning algorithm, to estimate stem total and assortment (commercial and pulpwood volumes in an industrial Pinus taeda L. forest plantation in southern Brazil. Random Forest was populated using field and lidar-derived forest metrics from 50 sample plots with trees ranging from three to nine years old. We found that a model defined as a function of only two metrics (height of the top of the canopy and the skewness of the vertical distribution of lidar points has a very strong and unbiased predictive power. We found that predictions of total, commercial, and pulp volume, respectively, showed an adjusted R2 equal to 0.98, 0.98 and 0.96, with unbiased predictions of −0.17%, −0.12% and −0.23%, and Root Mean Square Error (RMSE values of 7.83%, 7.71% and 8.63%. Our methodology makes use of commercially available airborne lidar and widely used mathematical tools to provide solutions for increasing the industry efficiency in monitoring and managing wood volume.

  7. The Attributable Proportion of Specific Leisure-Time Physical Activities to Total Leisure Activity Volume Among US Adults, National Health and Nutrition Examination Survey 1999-2006.

    Science.gov (United States)

    Watson, Kathleen Bachtel; Dai, Shifan; Paul, Prabasaj; Carlson, Susan A; Carroll, Dianna D; Fulton, Janet

    2016-11-01

    Previous studies have examined participation in specific leisure-time physical activities (PA) among US adults. The purpose of this study was to identify specific activities that contribute substantially to total volume of leisure-time PA in US adults. Proportion of total volume of leisure-time PA moderate-equivalent minutes attributable to 9 specific types of activities was estimated using self-reported data from 21,685 adult participants (≥ 18 years) in the National Health and Nutrition Examination Survey 1999-2006. Overall, walking (28%), sports (22%), and dancing (9%) contributed most to PA volume. Attributable proportion was higher among men than women for sports (30% vs. 11%) and higher among women than men for walking (36% vs. 23%), dancing (16% vs. 4%), and conditioning exercises (10% vs. 5%). The proportion was lower for walking, but higher for sports, among active adults than those insufficiently active and increased with age for walking. Compared with other racial/ethnic groups, the proportion was lower for sports among non-Hispanic white men and for dancing among non-Hispanic white women. Walking, sports, and dance account for the most activity time among US adults overall, yet some demographic variations exist. Strategies for PA promotion should be tailored to differences across population subgroups.

  8. Fasting gall bladder volume and lithogenicity in relation to glucose tolerance, total and intra-abdominal fat masses in obese non-diabetic subjects

    DEFF Research Database (Denmark)

    Hendel, H W; Højgaard, L; Andersen, T

    1998-01-01

    OBJECTIVE: To investigate whether total body fat mass or fat distribution and associated metabolic disturbances in glucose and lipid metabolism influence the well known gallstone pathogenetic factors in obese subjects in order to explain why some obese subjects develop gallstones and some do not...... with a specific radioimmunoassay. Insulin sensitivity was measured by the Minimal Model and glucose tolerance by an oral glucose tolerance test (OGTT). Serum lipid concentrations were measured by standard methods. RESULTS: The gallbladder volume in the fasting state increased with increasing intra-abdominal fat...... mass (P=0.006) and was increased in subjects with impaired glucose tolerance (41 vs 27 ml, P=0.001). The lithogenic index was > 1 in all subjects and correlated with total fat mass (P=0.04). CONCLUSION: Gallstone pathogenesis in obesity seems to be influenced by the total body fat mass and its regional...

  9. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 1: Plenary session; High burnup fuel; Containment and structural aging

    International Nuclear Information System (INIS)

    Monteleone, S.

    1997-01-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This first volume is divided into 3 sections: plenary session; high burnup fuel; and containment and structural aging. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  10. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 3: PRA and HRA; Probabilistic seismic hazard assessment and seismic siting criteria

    International Nuclear Information System (INIS)

    Monteleone, S.

    1997-02-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: PRA and HRA and probabilistic seismic hazard assessment and seismic siting criteria. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  11. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 1: Plenary session; High burnup fuel; Containment and structural aging

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1997-01-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This first volume is divided into 3 sections: plenary session; high burnup fuel; and containment and structural aging. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  12. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 3: PRA and HRA; Probabilistic seismic hazard assessment and seismic siting criteria

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1997-02-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: PRA and HRA and probabilistic seismic hazard assessment and seismic siting criteria. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  13. R and D on early detection of the Total Instantaneous Blockage for 4. Generation Reactors - Inventory of non-nuclear methods investigated by the CEA

    International Nuclear Information System (INIS)

    Paumel, K.; Jeannot, J.-P.; Vanderhaegen, M.; Massacret, N.; Jeanne, T.; Laffont, G.

    2013-06-01

    In the safety analysis for the core of the 4. Generation Reactors, the Total Instantaneous Blockage (TIB) is a hypothetic accident scenario involving the melting of the blocked subassembly with a risk of propagation to the neighbouring subassemblies. To avoid this latter consequence a detection system has to scram the reactor. For Superphenix or EFR project a Delayed Neutron Detection Integrated (DND I) was considered as efficient to limit the melting to the first neighbouring subassemblies. Nonetheless for the CFV core the objective of improving the safety leads to limit the melting to the blocked subassembly. For this purpose, the CEA has launched a program development to find a new detection method. This paper provides a brief review of the feedback of R and D, progress and program on the various early non-nuclear detection methods investigated by the CEA: - Temperature measurement at the subassemblies outlet by thermocouples. The advantage of this method is that it will require no additional instrumentation to that already present for continuous monitoring. - Temperature measurement at the subassemblies outlet by Optical Fibers Bragg Grating (OFBG). This technology has the electromagnetic immunity, compactness and short response time. - Temperature measurement at the subassemblies outlet by ultrasound. The measuring point is located closer to the head subassembly and the response time could be shorter. - Acoustic detection of sodium boiling. Boiling occurs early in the accident progress and the area to be monitored may be covered by few sensors. - Subassemblies loss of flow detection by eddy-current flowmeters. This method seems logically the easiest and the most immediate method to detect a blockage. To date, none of these methods has been fully demonstrated to be feasible. It should be noted that temperature measurement methods will probably consist of the detection of a low increase rate using specific signal processing. These methods have been compared

  14. Occupational radiation exposure at commercial nuclear power reactors and other facilities 1995: Twenty-eighth annual report. Volume 17

    International Nuclear Information System (INIS)

    Thomas, M.L.

    1997-01-01

    This report summarizes the occupational exposure data that are maintained in the US Nuclear Regulatory Commission's (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 1995 annual reports submitted by six of the seven categories of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. Since there are no geologic repositories for high-level waste currently licensed, only six categories will be considered in this report. In 1995, the annual collective dose per reactor for light water reactor licensees (LWRs) was 199 person-cSv (person-rem). This is the same value that was reported for 1994. The annual collective dose per reactor for boiling water reactors (BWRs) was 256 person-cSv (person-rem) and, for pressurized water reactors (PWRs), it was 170 person-cSv (person-rem). Analyses of transient worker data indicate that 17,153 individuals completed work assignments at two or more licensees during the monitoring year. The dose distributions are adjusted each year to account for the duplicate reporting of transient workers by multiple licensees. In 1995, the average measurable dose calculated from reported data was 0.26 cSv (rem). The corrected dose distribution resulted in an average measurable dose of 0.32 cSv (rem)

  15. Environmentally assisted cracking in light-water reactors: Semi-annual report, January - June 1997. Volume 24

    International Nuclear Information System (INIS)

    Chopra, O.K.; Chung, H.M.; Gruber, E.E.

    1998-04-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from January 1997 to June 1997. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Types 304 and 304L SS, and (c) EAC of Alloys 600 and 690. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during various portions of a tensile-loading cycle is equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated boiling water reactor (BWR) water at 288 C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in low-DO, simulated pressurized water reactor environments

  16. RAZORBACK - A Research Reactor Transient Analysis Code Version 1.0 - Volume 3: Verification and Validation Report.

    Energy Technology Data Exchange (ETDEWEB)

    Talley, Darren G.

    2017-04-01

    This report describes the work and results of the verification and validation (V&V) of the version 1.0 release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, the equation of motion for fuel element thermal expansion, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This V&V effort was intended to confirm that the code shows good agreement between simulation and actual ACRR operations.

  17. Analysis of the total system life cycle cost for the Civilian Radioactive Waste Management Program. Volume 1. The analysis and its results

    International Nuclear Information System (INIS)

    1986-04-01

    The total-system life-cycle cost (TSLCC) analysis for the Department of Energy's (DOE) Civilian Radioactive Waste Management Program is an ongoing activity that helps determine whether the revenue-producing mechanism established by the Nuclear Waste Policy Act of 1982 is sufficient to cover the cost of the program. This report provides cost estimates for the fourth evaluation of the adequacy of the fee. The total-system cost for the reference authorized-system program is estimated to be 24 to 32 billion (1985) dollars. The total-system cost for the reference improved-performance system is estimated to be 26 to 34 billion dollars. A number of sensitivity cases were analyzed. For the authorized system, the costs for the sensitivity cases studied range from 21 to 39 billion dollars. For the improved-performance system, which includes a facility for monitored retrievable storage, the total-system cost in the sensitivity cases is estimated to be as high as 41 billion dollars. The factors that affect costs more than any other single factor for both the authorized and the improved-performance systems are delays in repository startup. A preliminary analysis of the impact of extending the burnup of nuclear fuel in the reactor was also performed; its results indicate that the impact is insignificant: the total-system cost is essentially unchanged from the comparable constant-burnup cases. The current estimate of the the total-system cost for the reference authorized system is zero to 3 billion dollars (9%) higher than the estimate for the reference system in the January 1985 TSLCC analysis

  18. Effect of Risk Acceptance for Bundled Care Payments on Clinical Outcomes in a High-Volume Total Joint Arthroplasty Practice After Implementation of a Standardized Clinical Pathway.

    Science.gov (United States)

    Kee, James R; Edwards, Paul K; Barnes, Charles L

    2017-08-01

    The Bundled Payments for Care Improvement (BPCI) initiative and the Arkansas Payment Improvement (API) initiative seek to incentivize reduced costs and improved outcomes compared with the previous fee-for-service model. Before participation, our practice initiated a standardized clinical pathway (CP) to reduce length of stay (LOS), readmissions, and discharge to postacute care facilities. This practice implemented a standardized CP focused on patient education, managing patient expectations, and maximizing cost outcomes. We retrospectively reviewed all primary total joint arthroplasty patients during the initial 2-year "at risk" period for both BPCI and API and determined discharge disposition, LOS, and readmission rate. During the "at risk" period, the average LOS decreased in our total joint arthroplasty patients and our patients discharged home >94%. Patients within the BPCI group had a decreased discharge to home and decreased readmission rates after total hip arthroplasty, but also tended to be older than both API and nonbundled payment patients. While participating in the BPCI and API, continued use of a standardized CP in a high-performing, high-volume total joint practice resulted in maintenance of a low-average LOS. In addition, BPCI patients had similar outcomes after total knee arthroplasty, but had decreased rates of discharge to home and readmission after total hip arthroplasty. Copyright © 2017 Elsevier Inc. All rights reserved.

  19. Fort Hood Solar Total Energy Project. Volume II. Preliminary design. Part 2. System performance and supporting studies. Final report

    Energy Technology Data Exchange (ETDEWEB)

    None,

    1979-01-01

    The preliminary design developed for the Solar Total Energy System to be installed at Fort Hood, Texas, is presented. System performance analysis and evaluation are described. Feedback of completed performance analyses on current system design and operating philosophy is discussed. The basic computer simulation techniques and assumptions are described and the resulting energy displacement analysis is presented. Supporting technical studies are presented. These include health and safety and reliability assessments; solar collector component evaluation; weather analysis; and a review of selected trade studies which address significant design alternatives. Additional supporting studies which are generally specific to the installation site are reported. These include solar availability analysis; energy load measurements; environmental impact assessment; life cycle cost and economic analysis; heat transfer fluid testing; meteorological/solar station planning; and information dissemination. (WHK)

  20. Impact of airflow interaction on inhaled air quality and transport of contaminants in rooms with personalized and total volume ventilation

    DEFF Research Database (Denmark)

    Melikov, Arsen Krikor; Cermak, Radim; Kovar, O.

    2003-01-01

    The impact of airflow interaction on inhaled air quality and transport of contaminants between occupants was studied in regard to pollution from floor covering, human bioeffluents and exhaled air, with combinations of two personalized ventilation systems (PV) with mixing and displacement...... quality with personalized and mixing ventilation was higher or at least similar compared to mixing ventilation alone. In the case of PV combined with displacement ventilation, the interaction caused mixing of the room air, an increase in the transport of bioeffluents and exhaled air between occupants and...... ventilation. In total, 80 L/s of clean air supplied at 20°C was distributed between the ventilation systems at different combinations of personalized airflow rate. Two breathing thermal manikins were used to simulate occupants in a full-scale test room. Regardless of the airflow interaction, the inhaled air...

  1. Total energy cycle assessment of electric and conventional vehicles: an energy and environmental analysis. Volume 2: appendices A-D to technical report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-01-01

    This report compares the energy use, oil use and emissions of electric vehicles (EVs) with those of conventional, gasoline- powered vehicles (CVs) over the total life cycle of the vehicles. The various stages included in the vehicles` life cycles include vehicle manufacture, fuel production, and vehicle operation. Disposal is not included. An inventory of the air emissions associated with each stage of the life cycle is estimated. Water pollutants and solid wastes are reported for individual processes, but no comprehensive inventory is developed. Volume II contains additional details on the vehicle, utility, and materials analyses and discusses several details of the methodology.

  2. Exercise order affects the total training volume and the ratings of perceived exertion in response to a super-set resistance training session

    Directory of Open Access Journals (Sweden)

    Balsamo S

    2012-02-01

    Full Text Available Sandor Balsamo1–3, Ramires Alsamir Tibana1,2,4, Dahan da Cunha Nascimento1,2, Gleyverton Landim de Farias1,2, Zeno Petruccelli1,2, Frederico dos Santos de Santana1,2, Otávio Vanni Martins1,2, Fernando de Aguiar1,2, Guilherme Borges Pereira4, Jéssica Cardoso de Souza4, Jonato Prestes41Department of Physical Education, Centro Universitário UNIEURO, Brasília, 2GEPEEFS (Resistance training and Health Research Group, Brasília/DF, 3Graduate Program in Medical Sciences, School of Medicine, Universidade de Brasília (UnB, Brasília, 4Graduation Program in Physical Education, Catholic University of Brasilia (UCB, Brasília/DF, BrazilAbstract: The super-set is a widely used resistance training method consisting of exercises for agonist and antagonist muscles with limited or no rest interval between them – for example, bench press followed by bent-over rows. In this sense, the aim of the present study was to compare the effects of different super-set exercise sequences on the total training volume. A secondary aim was to evaluate the ratings of perceived exertion and fatigue index in response to different exercise order. On separate testing days, twelve resistance-trained men, aged 23.0 ± 4.3 years, height 174.8 ± 6.75 cm, body mass 77.8 ± 13.27 kg, body fat 12.0% ± 4.7%, were submitted to a super-set method by using two different exercise orders: quadriceps (leg extension + hamstrings (leg curl (QH or hamstrings (leg curl + quadriceps (leg extension (HQ. Sessions consisted of three sets with a ten-repetition maximum load with 90 seconds rest between sets. Results revealed that the total training volume was higher for the HQ exercise order (P = 0.02 with lower perceived exertion than the inverse order (P = 0.04. These results suggest that HQ exercise order involving lower limbs may benefit practitioners interested in reaching a higher total training volume with lower ratings of perceived exertion compared with the leg extension plus leg curl

  3. Total energy cycle assessment of electric and conventional vehicles: an energy and environmental analysis. Volume 4: peer review comments on technical report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-01-01

    This report compares the energy use, oil use and emissions of electric vehicles (EVs) with those of conventional, gasoline-powered vehicles (CVs) over the total life cycle of the vehicles. The various stages included in the vehicles` life cycles include vehicle manufacture, fuel production, and vehicle operation. Disposal is not included. An inventory of the air emissions associated with each stage of the life cycle is estimated. Water pollutants and solid wastes are reported for individual processes, but no comprehensive inventory is developed. Volume IV includes copies of all the external peer review comments on the report distributed for review in July 1997.

  4. Evaluation of nuclear facility decommissioning projects. Three Mile Island Unit 2 reactor building decontamination. Summary status report. Volume 2

    International Nuclear Information System (INIS)

    Doerge, D.H.; Miller, R.L.; Scotti, K.S.

    1986-05-01

    This document summarizes information relating to decontamination of the Three Mile Island Unit 2 (TMI-2) reactor building. The report covers activities for the period of June 1, 1979 through March 29, 1985. The data collected from activity reports, reactor containment entry records, and other sources were entered into a computerized data system which permits extraction/manipulation of specific information which can be used in planning for recovery from an accident similar to that experienced at TMI-2 on March 28, 1979. This report contains summaries of man-hours, manpower, and radiation exposures incurred during decontamination of the reactor building. Support activities conducted outside of radiation areas are excluded from the scope of this report. Computerized reports included in this document are: a chronological summary listing work performed relating to reactor building decontamination for the period specified; and summary reports for each major task during the period. Each task summary is listed in chronological order for zone entry and subtotaled for the number of personnel entries, exposures, and man-hours. Manually-assembled table summaries are included for: labor and exposures by department and labor and exposures by major activity

  5. System of large transport containers for waste from dismantling light water and gas-cooled nuclear reactors. Volume 1

    International Nuclear Information System (INIS)

    Price, M.S.T.

    1986-09-01

    General descriptions of the main types of reactors in the European Economic Community are given, a series of reference plants selected for further study. Estimates are made of the radioactive decommissioning wastes for each, including neutron-activated and contaminated materials. Regulations governing the transport of radioactive materials, both international and national, are reviewed. (U.K.)

  6. Operating experience feedback report: Reliability of safety-related steam turbine-driven standby pumps. Commercial power reactors, Volume 10

    International Nuclear Information System (INIS)

    Boardman, J.R.

    1994-10-01

    This report documents a detailed analysis of failure initiators, causes and design features for steam turbine assemblies (turbines with their related components, such as governors and valves) which are used as drivers for standby pumps in the auxiliary feedwater systems of US commercial pressurized water reactor plants, and in the high pressure coolant injection and reactor core isolation cooling systems of US commercial boiling water reactor plants. These standby pumps provide a redundant source of water to remove reactor core heat as specified in individual plant safety analysis reports. The period of review for this report was from January 1974 through December 1990 for licensee event reports (LERS) and January 1985 through December 1990 for Nuclear Plant Reliability Data System (NPRDS) failure data. This study confirmed the continuing validity of conclusions of earlier studies by the US Nuclear Regulatory Commission and by the US nuclear industry that the most significant factors in failures of turbine-driven standby pumps have been the failures of the turbine-drivers and their controls. Inadequate maintenance and the use of inappropriate vendor technical information were identified as significant factors which caused recurring failures

  7. Increased SRP reactor power

    International Nuclear Information System (INIS)

    MacAfee, I.M.

    1983-01-01

    Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

  8. Blood parasites, total plasma protein and packed cell volume of small wild mammals trapped in three mountain ranges of the Atlantic Forest in Southeastern Brazil

    Directory of Open Access Journals (Sweden)

    MAML. Silva

    Full Text Available A study of blood parasites in small wild non-flying mammals was undertaken in three areas of the Atlantic Forest in Southeastern Brazil: Serra de Itatiaia, RJ, Serra da Bocaina, SP and Serra da Fartura, SP, from June 1999 to May 2001. A total of 450 animals (15 species were captured in traps and it was observed in 15.5% of the blood smears the presence of Haemobartonella sp. and Babesia sp. in red blood cells. There was no statistically significant difference between parasited and non-parasited specimens regarding total plasma protein, packed cell volume and body weight, which strongly suggests that these specimens might be parasite reservoirs.

  9. System of large transport containers for waste from dismantling light water and gas-cooled nuclear reactors. Volume 2

    International Nuclear Information System (INIS)

    Price, M.S.T.; Lafontaine, I.

    1985-01-01

    The purpose of this volume is to assess the means of transportation of decommissioning wastes, costs of transport, radiological detriment attributable to transport and develops conceptual designs of large transport containers. The document ends with Conclusions and Recommendations

  10. Enhanced adsorption of benzene vapor on granular activated carbon under humid conditions due to shifts in hydrophobicity and total micropore volume.

    Science.gov (United States)

    Liu, Han-Bing; Yang, Bing; Xue, Nan-Dong

    2016-11-15

    A series of hydrophobic-modified (polydimethylsiloxane (PDMS) coating) activated carbons (ACs) were developed to answer a fundamental question: what are the determinants that dominate the adsorption on ACs under humid conditions? Using column experiments, an inter-comparison among bare-AC and PDMS-coated ACs was conducted regarding the association of surface characteristics and adsorption capacity. Primary outcomes occurred in two dominating markers, hydrophobicity and total micropore volume, which played a key role in water adsorption on ACs. However, their contributions to water adsorption on ACs substantially differed under different Pwater/Pair conditions. Hydrophobicity was the only contributor in Pwater/Pair=0.1-0.6, while the two markers contributed equally in Pwater/Pair=0.7-1.0. Furthermore, PDMS-coated AC had a significant increase in benzene adsorption capacities compared to bare-AC at 0-90% relative humidity, while these differences were not significant among PDMS-coated ACs. It is thus presumed that the balance between the two markers can be shifted to favor almost unchanged benzene adsorption capacities among PDMS-coated ACs over a large range of relative humidity. These findings suggest potential benefits of PDMS coating onto ACs in enhancing selective adsorption of hydrophobic volatile organic compounds under high humid conditions. To develop new porous materials with both high total micropore volume and hydrophobicity should thus be considered. Copyright © 2016 Elsevier B.V. All rights reserved.

  11. Resistance Training with Single vs. Multi-joint Exercises at Equal Total Load Volume: Effects on Body Composition, Cardiorespiratory Fitness, and Muscle Strength.

    Science.gov (United States)

    Paoli, Antonio; Gentil, Paulo; Moro, Tatiana; Marcolin, Giuseppe; Bianco, Antonino

    2017-01-01

    The present study aimed to compare the effects of equal-volume resistance training performed with single-joint (SJ) or multi-joint exercises (MJ) on VO 2 max, muscle strength and body composition in physically active males. Thirty-six participants were divided in two groups: SJ group ( n = 18, 182.1 ± 5.2, 80.03 ± 2.78 kg, 23.5 ± 2.7 years) exercised with only SJ exercises (e.g., dumbbell fly, knee extension, etc.) and MJ group ( n = 18, 185.3 ± 3.6 cm, 80.69 ± 2.98 kg, 25.5 ± 3.8 years) with only MJ exercises (e.g., bench press, squat, etc.). The total work volume (repetitions × sets × load) was equated between groups. Training was performed three times a week for 8 weeks. Before and after the training period, participants were tested for VO 2 max, body composition, 1 RM on the bench press, knee extension and squat. Analysis of covariance (ANCOVA) was used to compare post training values between groups, using baseline values as covariates. According to the results, both groups decreased body fat and increased fat free mass with no difference between them. Whilst both groups significantly increased cardiorespiratory fitness and maximal strength, the improvements in MJ group were higher than for SJ in VO 2 max (5.1 and 12.5% for SJ and MJ), bench press 1 RM (8.1 and 10.9% for SJ and MJ), knee extension 1 RM (12.4 and 18.9% for SJ and MJ) and squat 1 RM (8.3 and 13.8% for SJ and MJ). In conclusion, when total work volume was equated, RT programs involving MJ exercises appear to be more efficient for improving muscle strength and maximal oxygen consumption than programs involving SJ exercises, but no differences were found for body composition.

  12. Assessment of treatment response by total tumor volume and global apparent diffusion coefficient using diffusion-weighted MRI in patients with metastatic bone disease: a feasibility study.

    Directory of Open Access Journals (Sweden)

    Matthew D Blackledge

    Full Text Available We describe our semi-automatic segmentation of whole-body diffusion-weighted MRI (WBDWI using a Markov random field (MRF model to derive tumor total diffusion volume (tDV and associated global apparent diffusion coefficient (gADC; and demonstrate the feasibility of using these indices for assessing tumor burden and response to treatment in patients with bone metastases. WBDWI was performed on eleven patients diagnosed with bone metastases from breast and prostate cancers before and after anti-cancer therapies. Semi-automatic segmentation incorporating a MRF model was performed in all patients below the C4 vertebra by an experienced radiologist with over eight years of clinical experience in body DWI. Changes in tDV and gADC distributions were compared with overall response determined by all imaging, tumor markers and clinical findings at serial follow up. The segmentation technique was possible in all patients although erroneous volumes of interest were generated in one patient because of poor fat suppression in the pelvis, requiring manual correction. Responding patients showed a larger increase in gADC (median change = +0.18, range = -0.07 to +0.78 × 10(-3 mm2/s after treatment compared to non-responding patients (median change = -0.02, range = -0.10 to +0.05 × 10(-3 mm2/s, p = 0.05, Mann-Whitney test, whereas non-responding patients showed a significantly larger increase in tDV (median change = +26%, range = +3 to +284% compared to responding patients (median change = -50%, range = -85 to +27%, p = 0.02, Mann-Whitney test. Semi-automatic segmentation of WBDWI is feasible for metastatic bone disease in this pilot cohort of 11 patients, and could be used to quantify tumor total diffusion volume and median global ADC for assessing response to treatment.

  13. Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 5. Appendices

    Energy Technology Data Exchange (ETDEWEB)

    1976-05-01

    Volume V of the five-volume report consists of appendices, which provide supplementary information, with emphasis on characteristics of geologic formations that might be used for final storage or disposal. Appendix titles are: selected glossary; conversion factors; geologic isolation, including, (a) site selection factors for repositories of wastes in geologic media, (b) rock types--geologic occurrence, (c) glossary of geohydrologic terms, and (d) 217 references; the ocean floor; and, government regulations pertaining to the management of radioactive materials. (JGB)

  14. Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 5. Appendices

    International Nuclear Information System (INIS)

    1976-05-01

    Volume V of the five-volume report consists of appendices, which provide supplementary information, with emphasis on characteristics of geologic formations that might be used for final storage or disposal. Appendix titles are: selected glossary; conversion factors; geologic isolation, including, (a) site selection factors for repositories of wastes in geologic media, (b) rock types--geologic occurrence, (c) glossary of geohydrologic terms, and (d) 217 references; the ocean floor; and, government regulations pertaining to the management of radioactive materials

  15. Proposed nuclear weapons nonproliferation policy concerning foreign research reactor spent nuclear fuel: Appendix A, environmental justice analysis. Volume 2

    International Nuclear Information System (INIS)

    1995-03-01

    This is Appendix A to a draft Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel. This appendix addresses environmental justice for the acceptance of foreign research reactor spent nuclear fuel containing uranium enriched in the United States. Analyses of environmental justice concerns are provided in three areas: (1) potential ports of entry, (2) potential transportation routes from candidate ports of entry to interim management sites, and (3) areas surrounding potential interim management sites. These analyses lead to the conclusion that the alternatives analyzed in this Environmental Impact Statement (EIS) would result in no disproportionate adverse effects on minority populations or low-income communities surrounding the candidate ports, transport routes, or interim management sites

  16. Technology, safety and costs of decommissioning a Reference Boiling Water Reactor Power Station. Main report. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWe.

  17. Evaluation of nuclear facility decommissioning projects. Three Mile Island Unit 2 reactor defueling and disassembly. Summary status report. Volume 3

    International Nuclear Information System (INIS)

    Doerge, D.H.; Miller, R.L.; Scotti, K.S.

    1986-05-01

    This document summarizes information relating to the preparations for defueling the Three Mile Island Unit 2 (TMI-2) reactor and disassembly activities being performed concurrently with decontamination of the facility. Data have been collected from activity reports, reactor containment entry records, and other sources and entered in a computerized data sysem which permits extraction/manipulation of specific data which can be used in planning for recovery from a loss of coolant event similar to that experienced at TMI-2 on March 28, 1979. This report contains summaries of man-hours, manpower, and radiation exposures incurred during the period of April 23, 1979 to April 16, 1985, in the completion of activities related to preparation for reactor defueling. Support activities conducted outside of radiation areas are not included within the scope of this report. Computerized reports included in this document are: A chronological summary listing work performed for the period; and summary reports for each major task undertaken in connection with the specific scope of this report. Presented in chronological order for the referenced time period. Manually-assembled table summaries are included for: Labor and exposures by department; and labor and exposures by major activity

  18. Fukushima Daiichi Unit 1 Accident Progression Uncertainty Analysis and Implications for Decommissioning of Fukushima Reactors - Volume I.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Mattie, Patrick D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-01-01

    Sandia National Laboratories (SNL) has conducted an uncertainty analysis (UA) on the Fukushima Daiichi unit (1F1) accident progression with the MELCOR code. The model used was developed for a previous accident reconstruction investigation jointly sponsored by the US Department of Energy (DOE) and Nuclear Regulatory Commission (NRC). That study focused on reconstructing the accident progressions, as postulated by the limited plant data. This work was focused evaluation of uncertainty in core damage progression behavior and its effect on key figures-of-merit (e.g., hydrogen production, reactor damage state, fraction of intact fuel, vessel lower head failure). The primary intent of this study was to characterize the range of predicted damage states in the 1F1 reactor considering state of knowledge uncertainties associated with MELCOR modeling of core damage progression and to generate information that may be useful in informing the decommissioning activities that will be employed to defuel the damaged reactors at the Fukushima Daiichi Nuclear Power Plant. Additionally, core damage progression variability inherent in MELCOR modeling numerics is investigated.

  19. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Data manual. Part 3: Hardware component failure data; Volume 5, Revision 4

    International Nuclear Information System (INIS)

    Reece, W.J.; Gilbert, B.G.; Richards, R.E.

    1994-09-01

    This data manual contains a hard copy of the information in the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) Version 3.5 database, which is sponsored by the US Nuclear Regulatory Commission. NUCLARR was designed as a tool for risk analysis. Many of the nuclear reactors in the US and several outside the US are represented in the NUCLARR database. NUCLARR includes both human error probability estimates for workers at the plants and hardware failure data for nuclear reactor equipment. Aggregations of these data yield valuable reliability estimates for probabilistic risk assessments and human reliability analyses. The data manual is organized to permit manual searches of the information if the computerized version is not available. Originally, the manual was published in three parts. In this revision the introductory material located in the original Part 1 has been incorporated into the text of Parts 2 and 3. The user can now find introductory material either in the original Part 1, or in Parts 2 and 3 as revised. Part 2 contains the human error probability data, and Part 3, the hardware component reliability data

  20. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    The method of operating a water-cooled neutronic reactor having a graphite moderator is described which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40--60 volume percent of the mixture, in contact with the graphite moderator. 2 claims, 4 figures

  1. Mechanical spectral shift reactor

    International Nuclear Information System (INIS)

    Sherwood, D.G.; Wilson, J.F.; Salton, R.B.; Fensterer, H.F.

    1981-01-01

    A mechanical spectral shift reactor comprises apparatus for inserting and withdrawing water displacer elements from the reactor core for selectively changing the water-moderator volume in the core thereby changing the reactivity of the core. The apparatus includes drivemechanisms for moving the displacer elements relative to the core and guide mechanisms for guiding the displayer rods through the reactor vessel

  2. Mechanical spectral shift reactor

    International Nuclear Information System (INIS)

    Sherwood, D.G.; Wilson, J.F.; Salton, R.B.; Fensterer, H.F.

    1982-01-01

    A mechanical spectral shift reactor comprises apparatus for inserting and withdrawing water displacer elements from the reactor core for selectively changing the water-moderator volume in the core thereby changing the reactivity of the core. The apparatus includes drive mechanisms for moving the displacer elements relative to the core and guide mechanisms for guiding the displacer rods through the reactor vessel. (author)

  3. Spectral shift reactor control method

    International Nuclear Information System (INIS)

    Impink, A.J. Jr.

    1981-01-01

    A method of operating a nuclear reactor having a core and coolant displacer elements arranged in the core wherein is established a reator coolant temperature set point at which it is desired to operate said reactor and first reactor coolant temperature band limits are provided within which said set point is located and it is desired to operate said reactor charactrized in that said reactor coolant displacer elements are moved relative to the reactor core for adjusting the volume of reactor coolant in said core as said reactor coolant temperature approaches said first band limits thereby to maintain said reactor coolant temperature near said set point and within said first band limits

  4. Analysis of the total system life cycle cost for the Civilian Radioactive Waste Management Program: Volume 1, The analysis and its results

    International Nuclear Information System (INIS)

    1987-06-01

    This report provides cost estimates for the fifth evaluation of the adequacy of the fee and is consistent with the program strategy and plans. The total-system cost for the reference cases in the improved-performance system is estimated at $32.1 to $38.2 billion (expressed in constant 1986 dollars) over the entire life of the system...or $1.5 to $1.6 billion more than that of the authorized system (i.e., the system without an MRS facility). The current estimate of the total-system cost for the reference cases in the improved-performance system is $3.8 to $5.4 billion higher than the estimate for the same system in the 1986 TSLCC analysis. In the case with the maximum increase, nearly all of the higher cost is due to a $5.2-billion increase in the costs of development and evaluation (D and E); all other system costs are essentially unchanged. The cost difference between the improved-performance system and the authorized system is smaller than the difference estimated in last year's TSLCC analysis. Volume 2 presents the detailed results for the 1987 analysis of the total-system life cycle cost (TSLCC). It consists of four sections: Section A presents the yearly flows of waste between waste-management facilities for the 12 aggregate logistics cases that were studied; Section B presents the annual total-system costs for each of the 30 TSLCC cases by major cost category; Section C presents the annual costs for the disposal of 16,000 canisters of defense high-level waste (DHLW) by major cost category for each of the 30 TSLCC cases; and Section D presents a summary of the cost-allocation factors that were calculated to determine the defense waste share of the total-system costs

  5. Proposed nuclear weapons nonproliferation policy concerning foreign research reactor spent nuclear fuel: Appendix C, marine transport and associated environmental impacts. Volume 2

    International Nuclear Information System (INIS)

    1995-03-01

    This is Appendix C to a Draft Environmental Statement on a Proposed Nuclear Weapon Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel. Shipment of any material via ocean transport entails risks to both the ship's crew and the environment. The risks result directly from transportation-related accidents and, in the case of radioactive or other hazardous materials, also include exposure to the effects of the material itself. This appendix provides a description of the approach used to assess the risks associated with the transport of foreign research reactor spent nuclear fuel from a foreign port to a U.S. port(s) of entry. This appendix also includes a discussion of the shipping configuration of the foreign research reactor spent nuclear fuel, the possible types of vessels that could be used to make the shipments, the risk assessment methodology (addressing both incident-free and accident risks), and the results of the analyses. Analysis of activities in the port(s) is described in Appendix D. The incident-free and accident risk assessment results are presented in terms of the per shipment risk and total risks associated with the basic implementation of Management Alternative 1and other implementation alternatives. In addition, annual risks from incident-free transport are developed

  6. An approach to the verification of a fault-tolerant, computer-based reactor safety system: A case study using automated reasoning: Volume 1: Interim report

    International Nuclear Information System (INIS)

    Chisholm, G.H.; Kljaich, J.; Smith, B.T.; Wojcik, A.S.

    1987-01-01

    The purpose of this project is to explore the feasibility of automating the verification process for computer systems. The intent is to demonstrate that both the software and hardware that comprise the system meet specified availability and reliability criteria, that is, total design analysis. The approach to automation is based upon the use of Automated Reasoning Software developed at Argonne National Laboratory. This approach is herein referred to as formal analysis and is based on previous work on the formal verification of digital hardware designs. Formal analysis represents a rigorous evaluation which is appropriate for system acceptance in critical applications, such as a Reactor Safety System (RSS). This report describes a formal analysis technique in the context of a case study, that is, demonstrates the feasibility of applying formal analysis via application. The case study described is based on the Reactor Safety System (RSS) for the Experimental Breeder Reactor-II (EBR-II). This is a system where high reliability and availability are tantamount to safety. The conceptual design for this case study incorporates a Fault-Tolerant Processor (FTP) for the computer environment. An FTP is a computer which has the ability to produce correct results even in the presence of any single fault. This technology was selected as it provides a computer-based equivalent to the traditional analog based RSSs. This provides a more conservative design constraint than that imposed by the IEEE Standard, Criteria For Protection Systems For Nuclear Power Generating Stations (ANSI N42.7-1972)

  7. Decomposition of methane to hydrogen using nanosecond pulsed plasma reactor with different active volumes, voltages and frequencies

    International Nuclear Information System (INIS)

    Khalifeh, Omid; Mosallanejad, Amin; Taghvaei, Hamed; Rahimpour, Mohammad Reza; Shariati, Alireza

    2016-01-01

    Highlights: • CH 4 conversion into H 2 is investigated in a nanosecond pulsed DBD reactor. • The absence of CO and CO 2 in the product gas is highly favorable. • Effects of external electrode length, applied voltage and frequency are examined. • The maximum efficiency of 7.23% is achieved at the electrode length of 15 cm. • The maximum CH 4 conversion of 87.2% is obtained at discharge power 268.92 W. - Abstract: In this paper, the methane conversion into hydrogen is investigated experimentally in a nanosecond pulsed DBD reactor. In order to achieve pure hydrogen production with minimum power consumption, effects of some operating parameters including external electrode length, applied voltage and pulse repetition frequency have been evaluated. Results show that although higher CH 4 conversion and H 2 concentration can be obtained at longer electrode lengths, higher applied voltages and pulse repetition frequencies, these parameters should be optimized for efficient hydrogen production. Actually, the maximum CH 4 conversion of 87.2% and maximum hydrogen percentage of 80% are obtained at the external electrode length, discharge power, voltage and frequency of 15 cm, 268.92 W, 12 kV and 10 kHz, respectively. However, the maximum efficiency of 7.23% is achieved at the external electrode length of 15 cm, applied voltage of 6 kV, pulse repetition frequency of 0.9 kHz and discharge power of 4 W. Furthermore, at this condition, due to low temperature of discharge zone very little amount of solid carbon was observed on the inner electrode surface of the reactor.

  8. Enforcement actions: Significant actions resolved reactor licensees. Quarterly progress report, October--December 1994, Volume 13, No. 4, Part 1

    International Nuclear Information System (INIS)

    1995-02-01

    This compilation summarizes significant enforcement actions that have been resolved during one quarterly period (October--December 1994) and includes copies of letters Notices, and Orders sent by the Nuclear Regulatory Commission to reactor licensees with respect to these enforcement actions. It is anticipated that the information in this publication will be widely disseminated to managers and employees engaged in activities licensed by the NRC, so that actions can be taken to improve safety by avoiding future violations similar to those described in this publication

  9. Enforcement actions: Significant actions resolved, reactor licensees. Quarterly progress report, July--September 1994; Volume 13, Number 3, Part 1

    International Nuclear Information System (INIS)

    1994-12-01

    This compilation summarizes significant enforcement actions that have been resolved during one quarterly period (July--September 1994) and includes copies of letters, Notices, and Orders sent by the Nuclear Regulatory Commission to reactor licensees with respect to these enforcement actions. It is anticipated that the information in this publication will be widely disseminated to managers and employees engaged in activities licensed by the NRC, so that actions can be taken to improve safety by avoiding future violations similar to those described in this publication

  10. Enforcement actions: Significant actions resolved. Reactor licensees: Volume 14, No. 1, Part 1, Quarterly progress report January--March 1995

    International Nuclear Information System (INIS)

    1995-01-01

    This compilation summarizes significant enforcement actions that have been resolved during one quarterly period (January--March 1995) and includes copies of letters, Notices, and Orders sent by the Nuclear Regulatory Commission to reactor licensees with respect to these enforcement actions. It is anticipated that the information in this publication will be widely disseminated to managers and employees engaged in activities licensed by the NRC, so that actions can be taken to improve safety by avoiding future violations similar to those described in this publication

  11. Enforcement actions: Significant actions resolved, reactor licensees. Semiannual progress report, July--December 1997; Volume 16, Number 2, Part 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-04-01

    This compilation summarizes significant enforcement actions that have been resolved during the period (July--December 1997) and includes copies of letters, Notices, and Orders sent by the Nuclear Regulatory Commission to reactor licensees with respect to these enforcement actions. It is anticipated that the information in this publication will be widely disseminated to managers and employees engaged in activities licensed by the NRC, so that actions can be taken to improve safety by avoiding future violations similar to those described in this publication.

  12. Enforcement actions: Significant actions resolved reactor licensees. Volume 14, No. 2, Part 2, Quarterly progress report, April--June 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-08-01

    This compilation summarizes significant enforcement actions that have been resolved during one quarterly period (April--June 1995) and includes copies of letters, Notices, and Orders sent by the Nuclear Regulatory Commission to reactor licensees with respect to these enforcement actions. It is anticipated that the information in this publication will be widely disseminated to managers and employees engaged in activities licensed by the NRC, so that actions can be taken to improve safety by avoiding future violations similar to those described in this publication.

  13. Enforcement actions: Significant actions resolved reactor licensees. Quarterly progress report, October--December 1994, Volume 13, No. 4, Part 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-02-01

    This compilation summarizes significant enforcement actions that have been resolved during one quarterly period (October--December 1994) and includes copies of letters Notices, and Orders sent by the Nuclear Regulatory Commission to reactor licensees with respect to these enforcement actions. It is anticipated that the information in this publication will be widely disseminated to managers and employees engaged in activities licensed by the NRC, so that actions can be taken to improve safety by avoiding future violations similar to those described in this publication.

  14. Enforcement actions: Significant actions resolved reactor licensees. Volume 14, No. 2, Part 2, Quarterly progress report, April--June 1995

    International Nuclear Information System (INIS)

    1995-08-01

    This compilation summarizes significant enforcement actions that have been resolved during one quarterly period (April--June 1995) and includes copies of letters, Notices, and Orders sent by the Nuclear Regulatory Commission to reactor licensees with respect to these enforcement actions. It is anticipated that the information in this publication will be widely disseminated to managers and employees engaged in activities licensed by the NRC, so that actions can be taken to improve safety by avoiding future violations similar to those described in this publication

  15. Enforcement actions: Significant actions resolved reactor licensees. Volume 13, No. 1, Part 1: Quarterly progress report, January--March 1994

    International Nuclear Information System (INIS)

    1994-06-01

    This compilation summarizes significant enforcement actions that have been resolved during one quarterly period (January--March 1994) and includes copies of letters, Notices, and Orders sent by the Nuclear Regulatory Commission to reactor licensees with respect to these enforcement actions. It is anticipated that the information in this publication will be widely disseminated to managers and employees engaged in activities licensed by the NRC, so that actions can be taken to improve safety by avoiding future violations similar to these described in this publication

  16. Enforcement actions: Significant actions resolved, reactor licensees. Semiannual progress report, January--June 1997; Volume 16, Number 1, Part 2

    International Nuclear Information System (INIS)

    1997-09-01

    This compilation summarizes significant enforcement actions that have been resolved during the period (January--June 1997) and includes copies of letters, Notices, and Orders sent by the Nuclear Regulatory Commission to reactor licensees with respect to these enforcement actions. It is anticipated that the information in this publication will be widely disseminated to managers and employees engaged in activities licensed by the NRC, so that actions can be taken to improve safety by avoiding future violations similar to those described in this publication

  17. Enforcement actions: Significant actions resolved, reactor licensees. Semiannual progress report, July--December 1997; Volume 16, Number 2, Part 2

    International Nuclear Information System (INIS)

    1998-04-01

    This compilation summarizes significant enforcement actions that have been resolved during the period (July--December 1997) and includes copies of letters, Notices, and Orders sent by the Nuclear Regulatory Commission to reactor licensees with respect to these enforcement actions. It is anticipated that the information in this publication will be widely disseminated to managers and employees engaged in activities licensed by the NRC, so that actions can be taken to improve safety by avoiding future violations similar to those described in this publication

  18. Twenty-second water reactor safety information meeting: Proceedings. Volume 3: Primary systems integrity; Structural and seismic engineering; Aging research, products and applications

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24--26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  19. Twenty-second water reactor safety information meeting: Proceedings. Volume 1: Plenary session; Advanced instrumentation and control hardware and software; Human factors research; IPE and PRA

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24--26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  20. Twenty-second water reactor safety information meeting: Proceedings. Volume 3: Primary systems integrity; Structural and seismic engineering; Aging research, products and applications

    International Nuclear Information System (INIS)

    Monteleone, S.

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24--26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  1. Twenty-second water reactor safety information meeting: Proceedings. Volume 1: Plenary session; Advanced instrumentation and control hardware and software; Human factors research; IPE and PRA

    International Nuclear Information System (INIS)

    Monteleone, S.

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24--26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  2. Modelling of an intermediate pressure microwave oxygen discharge reactor: from stationary two-dimensional to time-dependent global (volume-averaged) plasma models

    International Nuclear Information System (INIS)

    Kemaneci, Efe; Graef, Wouter; Rahimi, Sara; Van Dijk, Jan; Kroesen, Gerrit; Carbone, Emile; Jimenez-Diaz, Manuel

    2015-01-01

    A microwave-induced oxygen plasma is simulated using both stationary and time-resolved modelling strategies. The stationary model is spatially resolved and it is self-consistently coupled to the microwaves (Jimenez-Diaz et al 2012 J. Phys. D: Appl. Phys. 45 335204), whereas the time-resolved description is based on a global (volume-averaged) model (Kemaneci et al 2014 Plasma Sources Sci. Technol. 23 045002). We observe agreement of the global model data with several published measurements of microwave-induced oxygen plasmas in both continuous and modulated power inputs. Properties of the microwave plasma reactor are investigated and corresponding simulation data based on two distinct models shows agreement on the common parameters. The role of the square wave modulated power input is also investigated within the time-resolved description. (paper)

  3. Twenty-First Water Reactor Safety Information Meeting. Volume 3, Primary system integrity; Aging research, products and applications; Structural and seismic engineering; Seismology and geology: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1994-04-01

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25-27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Selected papers were indexed separately for inclusion in the Energy Science and Technology Database.

  4. A total Ammonium Reactor (NHxR) for In Situ Mobile Measurements: A Critical Tool to Understand Aerosol Formation, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — We will develop, demonstrate, and optimize a front-end ammonium reactor (NHxR) for the fast, precise, and accurate measurement of gas-phase ammonia (NH3) and...

  5. Seismic design technology for breeder reactor structures. Volume 2. Special topics in soil/structure interaction analyses

    International Nuclear Information System (INIS)

    Reddy, D.P.

    1983-04-01

    This volume is divided into six chapters: definition of seismic input ground motion, review of state-of-the-art procedures, analysis guidelines, rock/structure interaction analysis example, comparison of two- and three-dimensional analyses, and comparison of analyses using FLUSH and TRI/SAC Codes

  6. Re-irradiation after gross total resection of recurrent glioblastoma. Spatial pattern of recurrence and a review of the literature as a basis for target volume definition

    Energy Technology Data Exchange (ETDEWEB)

    Straube, Christoph; Elpula, Greeshma [Technische Universitaet Muenchen (TUM), Department of Radiation Oncology, Klinikum rechts der Isar, Muenchen (Germany); Gempt, Jens; Gerhardt, Julia; Meyer, Bernhard [Technische Universitaet Muenchen (TUM), Department of Neurosurgery, Klinikum rechts der Isar, Muenchen (Germany); Bette, Stefanie; Zimmer, Claus [Technische Universitaet Muenchen (TUM), Department of Neuroradiology, Klinikum rechts der Isar, Muenchen (Germany); Schmidt-Graf, Friederike [Technische Universitaet Muenchen (TUM), Department of Neurology, Klinikum rechts der Isar, Muenchen (Germany); Combs, Stephanie E. [Technische Universitaet Muenchen (TUM), Department of Radiation Oncology, Klinikum rechts der Isar, Muenchen (Germany); Helmholtz Zentrum Muenchen, Institute for Innovative Radiotherapy (iRT), Department of Radiation Sciences (DRS), Oberschleissheim (Germany)

    2017-11-15

    Currently, patients with gross total resection (GTR) of recurrent glioblastoma (rGBM) undergo adjuvant chemotherapy or are followed up until progression. Re-irradiation, as one of the most effective treatments in macroscopic rGBM, is withheld in this situation, as uncertainties about the pattern of re-recurrence, the target volume, and also the efficacy of early re-irradiation after GTR exist. Imaging and clinical data from 26 consecutive patients with GTR of rGBM were analyzed. The spatial pattern of recurrences was analyzed according to the RANO-HGG criteria (''response assessment in neuro-oncology criteria for high-grade gliomas''). Progression-free (PFS) and overall survival (OS) were analyzed by the Kaplan-Meier method. Furthermore, a systematic review was performed in PubMed. All but 4 patients underwent adjuvant chemotherapy after GTR. Progression was diagnosed in 20 of 26 patients and 70% of recurrent tumors occurred adjacent to the resection cavity. The median extension beyond the edge of the resection cavity was 20 mm. Median PFS was 6 months; OS was 12.8 months. We propose a target volume containing the resection cavity and every contrast enhancing lesion as the gross tumor volume (GTV), a spherical margin of 5-10 mm to generate the clinical target volume (CTV), and a margin of 1-3 mm to generate the planning target volume (PTV). Re-irradiation of this volume is deemed to be safe and likely to prolong PFS. Re-irradiation is worth considering also after GTR, as the volumes that need to be treated are limited and re-irradiation has already proven to be a safe treatment option in general. The strategy of early re-irradiation is currently being tested within the GlioCave/NOA 17/Aro 2016/03 trial. (orig.) [German] Patienten mit einem rezidivierten Glioblastom (rGBM) werden, wenn eine komplette Resektion (GTR) des makroskopischen Rezidivs durchgefuehrt wurde, aktuell meist systemisch adjuvant behandelt oder einer engmaschigen Nachsorge

  7. High-throughput analysis using non-depletive SPME: challenges and applications to the determination of free and total concentrations in small sample volumes.

    Science.gov (United States)

    Boyacı, Ezel; Bojko, Barbara; Reyes-Garcés, Nathaly; Poole, Justen J; Gómez-Ríos, Germán Augusto; Teixeira, Alexandre; Nicol, Beate; Pawliszyn, Janusz

    2018-01-18

    In vitro high-throughput non-depletive quantitation of chemicals in biofluids is of growing interest in many areas. Some of the challenges facing researchers include the limited volume of biofluids, rapid and high-throughput sampling requirements, and the lack of reliable methods. Coupled to the above, growing interest in the monitoring of kinetics and dynamics of miniaturized biosystems has spurred the demand for development of novel and revolutionary methodologies for analysis of biofluids. The applicability of solid-phase microextraction (SPME) is investigated as a potential technology to fulfill the aforementioned requirements. As analytes with sufficient diversity in their physicochemical features, nicotine, N,N-Diethyl-meta-toluamide, and diclofenac were selected as test compounds for the study. The objective was to develop methodologies that would allow repeated non-depletive sampling from 96-well plates, using 100 µL of sample. Initially, thin film-SPME was investigated. Results revealed substantial depletion and consequent disruption in the system. Therefore, new ultra-thin coated fibers were developed. The applicability of this device to the described sampling scenario was tested by determining the protein binding of the analytes. Results showed good agreement with rapid equilibrium dialysis. The presented method allows high-throughput analysis using small volumes, enabling fast reliable free and total concentration determinations without disruption of system equilibrium.

  8. Proceedings of the international conference on mathematics and computations, reactor physics, and environmental analyses. Volume 1 and 2

    International Nuclear Information System (INIS)

    Anon.

    1995-01-01

    The International Conference on Mathematics and Computations, Reactor Physics, and Environmental Analyses marks the sixteenth biennial topical meeting of the Mathematics and Computation (M ampersand C) Division of the American Nuclear Society (ANS). This conference combines many traditional features of M ampersand C conferences with several new aspects. The meeting is, for the first time, being held in Portland, Oregon and sponsored by the ANS Eastern Washington Section. Three of the cosponsors - the ANS Reactor Physics Division, the European Nuclear Society, and the Atomic Energy Society of Japan - have participated in a series of such meetings, with very successful results. The fourth cosponsor, the ANS Environmental Science Division, is participating for the first time as a cosponsor of a M ampersand C topical meeting, as a result of the M ampersand C Division's decision to formally include the area of environmental analyses as a major focus of the conference, another 'first.' Separate abstracts have been submitted to the energy database for contributions to this conference

  9. Licensed operating reactors. Status summary report, data as of 6-30-84. Volume 8, No. 7

    International Nuclear Information System (INIS)

    1984-08-01

    This monthly report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States. The percentage computations, the vendor capacity factors, and actual vs potential energy production are computed using actual data for the period of consideration. The percentages listed in power generation are computed as an arithmetic average. The factors for the life-span of each unit (the Cumulative column) are reported by the utility and are not entirely re-computed by NRC. Utility power production data are checked for consistency with previously submitted statistics

  10. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

    International Nuclear Information System (INIS)

    1994-07-01

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR section 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE's application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design

  11. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 2: Appendices

    International Nuclear Information System (INIS)

    1994-07-01

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR section 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE's application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design

  12. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-01

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

  13. Propagation calculation for reactor cases

    Energy Technology Data Exchange (ETDEWEB)

    Yang Yanhua [School of Power and Energy Engineering, Shanghai Jiao Tong Univ., Shanghai (China); Moriyama, K.; Maruyama, Y.; Nakamura, H.; Hashimoto, K. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-11-01

    The propagation of steam explosion for real reactor geometry and conditions are investigated by using the computer code JASMINE-pro. The ex-vessel steam explosion is considered, which is described as follow: during the accident of reactor core meltdown, the molten core melts a hole at the bottom of reactor vessel and causes the higher temperature core fuel being leaked into the water pool below reactor vessel. During the melt-water mixing interaction process, the high temperature melt evaporates the cool water at an extreme high rate and might induce a steam explosion. A steam explosion could experience first the premixing phase and then the propagation explosion phase. For a propagation calculation, we should know the information about the initial fragmentation time, the total melt mass, premixing region size, initial void fraction and distribution of the melt volume fraction, and so on. All the initial conditions used in this calculation are based on analyses from some simple assumptions and the observation from the experiments. The results show that the most important parameter for the initial condition of this phase is the total mass and its initial distribution. This gives the requirement for a premixing calculation. On the other hand, for higher melt volume fraction case, the fragmentation is strong so that the local pressure can exceed over the EOS maximum pressure of the code, which lead to the incorrect calculation or divergence of the calculation. (Suetake, M.)

  14. Resistance Training with Single vs. Multi-joint Exercises at Equal Total Load Volume: Effects on Body Composition, Cardiorespiratory Fitness, and Muscle Strength

    Directory of Open Access Journals (Sweden)

    Antonio Paoli

    2017-12-01

    Full Text Available The present study aimed to compare the effects of equal-volume resistance training performed with single-joint (SJ or multi-joint exercises (MJ on VO2max, muscle strength and body composition in physically active males. Thirty-six participants were divided in two groups: SJ group (n = 18, 182.1 ± 5.2, 80.03 ± 2.78 kg, 23.5 ± 2.7 years exercised with only SJ exercises (e.g., dumbbell fly, knee extension, etc. and MJ group (n = 18, 185.3 ± 3.6 cm, 80.69 ± 2.98 kg, 25.5 ± 3.8 years with only MJ exercises (e.g., bench press, squat, etc.. The total work volume (repetitions × sets × load was equated between groups. Training was performed three times a week for 8 weeks. Before and after the training period, participants were tested for VO2max, body composition, 1 RM on the bench press, knee extension and squat. Analysis of covariance (ANCOVA was used to compare post training values between groups, using baseline values as covariates. According to the results, both groups decreased body fat and increased fat free mass with no difference between them. Whilst both groups significantly increased cardiorespiratory fitness and maximal strength, the improvements in MJ group were higher than for SJ in VO2max (5.1 and 12.5% for SJ and MJ, bench press 1 RM (8.1 and 10.9% for SJ and MJ, knee extension 1 RM (12.4 and 18.9% for SJ and MJ and squat 1 RM (8.3 and 13.8% for SJ and MJ. In conclusion, when total work volume was equated, RT programs involving MJ exercises appear to be more efficient for improving muscle strength and maximal oxygen consumption than programs involving SJ exercises, but no differences were found for body composition.

  15. Low-level radioactive waste from commercial nuclear reactors. Volume 2. Treatment, storage, disposal, and transportation technologies and constraints

    Energy Technology Data Exchange (ETDEWEB)

    Jolley, R.L.; Dole, L.R.; Godbee, H.W.; Kibbey, A.H.; Oyen, L.C.; Robinson, S.M.; Rodgers, B.R.; Tucker, R.F. Jr.

    1986-05-01

    The overall task of this program was to provide an assessment of currently available technology for treating commercial low-level radioactive waste (LLRW), to initiate development of a methodology for choosing one technology for a given application, and to identify research needed to improve current treatment techniques and decision methodology. The resulting report is issued in four volumes. Volume 2 discusses the definition, forms, and sources of LLRW; regulatory constraints affecting treatment, storage, transportation, and disposal; current technologies used for treatment, packaging, storage, transportation, and disposal; and the development of a matrix relating treatment technology to the LLRW stream as an aid for choosing methods for treating the waste. Detailed discussions are presented for most LLRW treatment methods, such as aqueous processes (e.g., filtration, ion exchange); dewatering (e.g., evaporation, centrifugation); sorting/segregation; mechanical treatment (e.g., shredding, baling, compaction); thermal processes (e.g., incineration, vitrification); solidification (e.g., cement, asphalt); and biological treatment.

  16. Low-level radioactive waste from commercial nuclear reactors. Volume 2. Treatment, storage, disposal, and transportation technologies and constraints

    International Nuclear Information System (INIS)

    Jolley, R.L.; Dole, L.R.; Godbee, H.W.; Kibbey, A.H.; Oyen, L.C.; Robinson, S.M.; Rodgers, B.R.; Tucker, R.F. Jr.

    1986-05-01

    The overall task of this program was to provide an assessment of currently available technology for treating commercial low-level radioactive waste (LLRW), to initiate development of a methodology for choosing one technology for a given application, and to identify research needed to improve current treatment techniques and decision methodology. The resulting report is issued in four volumes. Volume 2 discusses the definition, forms, and sources of LLRW; regulatory constraints affecting treatment, storage, transportation, and disposal; current technologies used for treatment, packaging, storage, transportation, and disposal; and the development of a matrix relating treatment technology to the LLRW stream as an aid for choosing methods for treating the waste. Detailed discussions are presented for most LLRW treatment methods, such as aqueous processes (e.g., filtration, ion exchange); dewatering (e.g., evaporation, centrifugation); sorting/segregation; mechanical treatment (e.g., shredding, baling, compaction); thermal processes (e.g., incineration, vitrification); solidification (e.g., cement, asphalt); and biological treatment

  17. Phase 1 study of metallic cask systems for spent fuel management from reactor to repository. Volume I. Phase 1 study summary

    International Nuclear Information System (INIS)

    1986-02-01

    It was proposed to perform a systems evaluation of metallic cask systems in order to define and examine the use of various metallic cask concepts or combination of concepts for the overall inventory management of spent fuel starting with its discharge from reactors to its emplacement in geologic repositories. This systems evaluation occurs in three phases. This three phase systems evaluation leads to a definition and recommendation of a sound and practical metallic cask system to accomplish efficient and effective management of spent fuel in the back end of the nuclear fuel cycle. Phase 1 Study objectives: establish system-wide functional criteria and assumptions; perform the systems engineering needed to define the metallic cask concepts and their feasibility; perform a screening evaluation of the technical and economic merits of the concepts; and recommend those to be included for a more detailed systems evaluation in Phase 2. Phase 2 Study objectives: refine the system-wide functional criteria and assumptions; perform the design engineering needed to enhance the validity and workability of those concepts recommended in Phase 1; and perform a more detailed systems evaluation. Phase 3 Study objectives: conclude the systems evaluation and develop an implementation plan. Volume I presents an overview of the detailed systems evaluation presented in Volume II

  18. Development Of A Method For Measurement Of Total Neutron Cross Sections Based On The Neutron Transmission Method Using A He-3 Counter On Filtered Neutron Beams At Dalat Research Reactor

    International Nuclear Information System (INIS)

    Tran Tuan Anh; Dang Lanh; Nguyen Canh Hai; Nguyen Xuan Hai; Pham Kien; Nguyen Thuy Nham; Pham Ngoc Son; Ho Huu Thang

    2007-01-01

    Determination of total neutron cross sections and average resonance parameters in the energy range from tens keV to hundreds keV is important for fast reactors calculations and designs because this energy range gives the most output of all neutron induced reactions in the spectrum of fast reactors. Besides, the total neutron cross section measurement is also one of the methods for determination of s, p and d-wave neutron strength functions. The purpose of this project is to develop a method for measurement of total neutron cross sections based on the neutron transmission technique using a He-3 counter. The average total neutron cross sections of 238 U were obtained from neutron transmission measurements on filtered neutron beams of 55 keV and 144 keV at the horizontal channel No.4 of the Dalat research reactor. The present results have been compared with the previous measurements, and the evaluated data from ENDF/B-6.8 library. (author)

  19. [Total brain T2-hyperintense lesion-volume and the axonal damage in the normal-appearing white matter of brainstem in early lapsing-remitting multiple sclerosis].

    Science.gov (United States)

    Pascual-Lozano, A M; Martínez-Bisbal, M C; Boscá-Blasco, I; Valero-Merino, C; Coret-Ferrer, F; Martí-Bonmatí, L; Martínez-Granados, B; Celda, B; Casanova-Estruch, B

    To evaluate the relationship between the total brain T2-hyperintense lesion volume (TBT2LV) and the axonal damage in the normal-appearing white matter of brainstem measured by 1H-MRS in a group of early relapsing-remitting multiple sclerosis patients. 40 relapsing-remitting multiple sclerosis patients and ten sex- and age-matched healthy subjects were prospectively studied for two years. T2-weighted MR and 1H-MRS imaging were acquired at time of recruitment and at year two. The TBT2LV was calculated with a semiautomatic program; N-acetylaspartate (NAA), creatine (Cr) and choline (Cho) resonances areas were integrated with jMRUI program and the ratios were calculated for four volume elements that represented the brainstem. At basal study we obtained an axonal loss (as a decrement of NAA/ Cho ratio) in the group of patients compared with controls (p = 0.017); this axonal loss increased at the second year of the follow-up for patients (NAA/Cho decrease, p = 0.004, and NAA/Cr decrease, p = 0.002) meanwhile control subjects had no significant metabolic changes. Higher lesion load was correlated with a poor clinical outcome, being the correlation between the basal TBT2LV and the Expanded Disability Status Scale at second year (r = 0.299; p = 0.05). Besides, axonal loss was not homogeneous for all multiple sclerosis patients, being stronger in the subgroup of patients with high basal TBT2LV (p = 0.043; ANOVA). Our data suggest that axonal damage is early in multiple sclerosis and higher in patients high basal TBT2LV, suggesting a possible relationship between these two phenomena.

  20. Transitioning to the direct anterior approach in total hip arthroplasty. Is it a true muscle sparing approach when performed by a low volume hip replacement surgeon?

    Science.gov (United States)

    Nistor, Dan-Viorel; Caterev, Sergiu; Bolboacă, Sorana-Daniela; Cosma, Dan; Lucaciu, Dan Osvald Gheorghe; Todor, Adrian

    2017-11-01

    We conducted this study to establish if the transition from a lateral approach (LA) to the direct anterior approach (DAA) for a low volume hip arthroplasty surgeon during the steep learning curve can be performed maintaining the muscle sparing approach of the DAA without increasing the complication rates. In this controlled, prospective, randomized clinical study we investigated 70 patients (35 DAA, 35 LA) with similar demographics that underwent a total hip arthroplasty. Assessment of the two approaches consisted of determining the invasiveness through serum markers for muscle damage (i.e. myoglobin, creatine kinase and lactate dehydrogenase), the operative parameters such as post-operative pain and rescue medication consumption, the component positioning and complication rates. Post-operative myoglobin levels were higher (p < 0.001) in the LA group (326.42 ± 84.91 ng/mL) as compared to the DAA group (242.80 ± 71.03 ng/mL), but with no differences regarding other biomarkers for muscle damage. Pain levels were overall lower in the DAA group, with a statistical and clinical difference during surgery day (p < 0.001) associated with lower (p < 0.001) rescue medication consumption (median 1 (1; 3) mg morphine vs. 3 (2; 4) mg morphine). Most patients in the LA group reported chronic post-operative pain throughout all three evaluated months, while the majority of patients in the DAA group reported no pain after week six. Component positioning did not differ significantly between groups and neither did complication rates. The DAA can be transitioned from the LA safely, without higher complication rates while maintaining its muscle spearing advantages when performed by a low volume hip arthroplasty surgeon.

  1. Occupational radiation exposure at commercial nuclear power reactors and other facilities 1992. Twenty-fifth annual report, Volume 14

    International Nuclear Information System (INIS)

    Raddatz, C.T.

    1993-12-01

    This report summarizes the occupational radiation exposure information that has been reported to the NRC's Radiation Exposure Information Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC licensees during the years 1969 through 1992. The bulk of the data presented in the report was obtained from annual radiation exposure reports submitted in accordance with the requirements of 10CFR20.407 and the technical specifications of nuclear power plants. Data on workers terminating their employment at certain NRC licensed facilities were obtained from reports submitted pursuant to 10CFR20.408. The 1992 annual reports submitted by about 364 licensees indicated that approximately 204,365 individuals were monitored, 183,927 of whom were monitored by nuclear power facilities. They incurred an average individual dose of 0.16 rem (cSv) and an average measurable dose of about 0.30 (cSv). Termination radiation exposure reports were analyzed to reveal that about 74,566 individuals completed their employment with one or more of the 364 covered licensees during 1992. Some 71,846 of these individuals terminated from power reactor facilities, and about 9,724 of them were considered to be transient workers who received an average dose of 0.50 rem (cSv)

  2. Feasibility of and methodology for thermal annealing an embrittled reactor vessel. Volume 1. Program overview. Final report

    International Nuclear Information System (INIS)

    Mager, T.R.

    1983-01-01

    An EPRI sponsored program was carried out by Westinghouse to determine the extent of fracture toughness recovery as a function of annealing time and temperature for neutron embrittlement sensitive reactor vessel material and to develop an optimal thermal anneal procedure for field applications. Program materials were three weldments fabricated by Combustion Engineering, Inc., from the same heat of A533 Grade B Class 1 plate material and the same heat of MnMoNi weld wire. The only variables were the target copper level and the welding flux which was Linde Grade 80 and Linde 0091. Weldments of 0.22, 0.36, and 0.41 wt % copper were produced. It was concluded from this study that excellent recovery of all properties could be achieved by annealing at 850 0 F (454 0 C) and above for 168 hours. Such an annealing resulted in ductile-brittle transition temperature shift recovery of 80 to 100%, and reirradiation after this annealing indicated that the ductile-brittle transition temperature shift appears to continue at the rate which would have been expected had no anneal been performed. System limitations were identified for both wet and dry annealing methods

  3. The Reactor Analysis Support Package (RASP): Volume 10, Guidelines for developing a reload licensing capability:Final report

    International Nuclear Information System (INIS)

    Diamond, D.J.

    1988-08-01

    The EPRI Reactor Analysis Support Package (RASP) consists of computer codes and documentation for calculating core performance and plant transients. This report was written to help a utility to use these tools properly for reload design and safety evaluations. The emphasis is on the steps that the utility can take to develop a methodology that is approved by the Nuclear Regulatory Commission for reload licensing submittals. The report treats both the planning and the implementation of this type of calculational capability. With regard to planning, there is discussion defining objectives, resource requirements, organization and scheduling. In order to help the engineering staff implement the plan there is discussion of the development of a methodology for event analysis, qualification of the methods, and the writing of design control procedures and topical reports. The experience of utilities, and especially of GPU Nuclear Corporation (GPUN), in developing a reload licensing capability is cited throughout the report and extracts from GPUN design control procedures are included in the appendices. 16 refs., 23 figs., 9 tabs

  4. Occupational radiation exposure at commercial nuclear power reactors and other facilities, 1993. Volume 15, Twenty-six annual report

    International Nuclear Information System (INIS)

    Raddatz, C.T.

    1995-01-01

    This report the occupational radiation exposure information that has been reported to the NRC's Radiation Exposure Information Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC licensees during the years 1969 through 1993. The bulk of the data presented in the report was obtained from annual radiation exposure reports submitted in accordance with the requirements of 10 CFR 20.407 and the technical specifications of nuclear power plants. Data on workers terminating their employment at certain NRC licensed facilities were obtained from reports submitted pursuant to 10 CFR 20.408. The 1993 annual reports submitted by about 360 licensees indicated that approximately 189,711 individuals were monitored, 169,872 of whom were monitored by nuclear power facilities. They incurred an average individual dose of 0.16 rem (cSv) and an average measured dose of about 0.31 (cSv). Termination radiation exposure reports were analyzed to reveal that about 99,749 individuals completed their employment with one or more of the 360 covered licensees during 1993. Some 91,000 of these individuals terminated from power reactor facilities, and about 12,685 of them were considered to be transient workers who received an average dose of 0.49 rem (cSv)

  5. Evaluation of Packed-Bed Reactor and Continuous Stirred Tank Reactor for the Production of Colchicine Derivatives

    OpenAIRE

    Dubey, Kashyap Kumar; Kumar, Dhirendra; Kumar, Punit; Haque, Shafiul; Jawed, Arshad

    2013-01-01

    Bioconversion of colchicine into its pharmacologically active derivative 3-demethylated colchicine (3-DMC) mediated by P450BM3 enzyme is an economic and promising strategy for the production of this inexpensive and potent anticancer drug. Continuous stirred tank reactor (CSTR) and packed-bed reactor (PBR) of 3 L and 2 L total volumes were compared for the production of 3-demethylated colchicine (3-DMC) a colchicine derivative using Bacillus megaterium MTCC*420 under aerobic conditions. Statis...

  6. Bioconversion reactor

    Science.gov (United States)

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  7. Prognostic value of metabolic tumour volume and total lesion glycolysis in 18F-FDG PET/CT scans in locally advanced breast cancer staging.

    Science.gov (United States)

    Jiménez-Ballvé, A; García García-Esquinas, M; Salsidua-Arroyo, O; Serrano-Palacio, A; García-Sáenz, J A; Ortega Candil, A; Fuentes Ferrer, M E; Rodríguez Rey, C; Román-Santamaría, J M; Moreno, F; Carreras-Delgado, J L

    To determine whether metabolic tumour volume (MTV) and total lesion glycolysis (TLG) are able to predict recurrence risk in locally advanced breast cancer (LABC) patients. Retrospective study of LABC patients who undertook neoadjuvant, local and adjuvant treatment and follow up. A 18 F-FDG PET/CT study for initial staging was performed analysing in this study different metabolic parameters (MTV, TLG, SUVmax and SUVmed) both in the primary tumour (T) as well as in axillary nodes (N) and whole-body (WB). Forty females were included between January 2010-2011; follow up until January 2015 was completed. The average follow-up was 46 months. Twenty percent presented recurrence: local disease (n=2) and distant metastasis (n=6); 3 patients died (38% of the patients which recurred and 7.5% from the total). SUVmax, MTV and TLG, in T, N and WB, were higher in those patients with recurrence. The MTV and TLG parameters in the tumour (T) were related to the recurrence rate (P=.020 and P=.028, respectively); whereas SUVmax in the lymph nodes (N) was significantly related (P=.008) to the recurrence rate. The best cut-off points to predict recurrence where: MTV T ≥19.3cm 3 , TLG T≥74.4g and SUVmax N≥13.8, being 10-12 times more likely to recidivate when these thresholds where exceeded. Tumour grade was the only clinical-pathological variable which was related to recurrence probability (p=.035). In this study of LABC patients the metabolic parameters which have a better relationship with recurrence rate are: MTV and TLG in the primary tumour, SUVmax in the regional lymph node disease and whole-body PET data. Copyright © 2016 Elsevier España, S.L.U. y SEMNIM. All rights reserved.

  8. Implementation of a Total Hip Arthroplasty Care Pathway at a High-Volume Health System: Effect on Length of Stay, Discharge Disposition, and 90-Day Complications.

    Science.gov (United States)

    Featherall, Joseph; Brigati, David P; Faour, Mhamad; Messner, William; Higuera, Carlos A

    2018-06-01

    Standardized care pathways are evidence-based algorithms for optimizing an episode of care. Despite the theoretical promise of care pathways, there is an inconsistent literature demonstrating improvements in patient care. The authors hypothesized that implementing a care pathway, across 11 hospitals, would decrease hospital length of stay (LOS), decrease postoperative complications at 90 days, and increase discharges to home. A multidisciplinary team developed an evidence-based care pathway for total hip arthroplasty (THA) perioperative care. All patients receiving THA in 2013 (pre-protocol, historical control), 2014 (transition), and 2015 (full protocol implementation) were included in the analysis. Multivariable regression assessed the relationship of the care pathway to 90-day postoperative complications, LOS, and discharge disposition. Cost savings were estimated using previously published postarthroplasty episode and per diem hospital costs. A total of 6090 primary THAs were conducted during the study period. After adjusting for the covariates, the full protocol implementation was associated with a decrease in LOS (mean ratio, 0.747; 95% confidence interval [CI; 0.727, 0.767]) and an increase in discharges to home (odds ratio, 2.079; 95% CI [1.762, 2.456]). The full protocol implementation was not associated with a change in 90-day complications (odds ratio, 1.023; 95% CI [0.841, 1.245]). Payer-perspective-calculated theoretical cost savings, including both index admission and postdischarge costs, were $2533 per patient. The THA care pathway implementation was successful in reducing LOS and increasing discharges to home. The care pathway was not associated with a change in 90-day complications; further targeted interventions in this area are needed. Despite care standardization efforts, high-volume hospitals and surgeons had higher performance. Extrapolation of theoretical cost savings indicates that widespread THA care pathway adoption could lead to national

  9. Modelo de predição para o volume total de Quaruba (Vochysia inundata ducke via análise de fatores e regressão

    Directory of Open Access Journals (Sweden)

    Mário Diego Rocha Valente

    2011-04-01

    Full Text Available Neste trabalho propôs-se um método para a construção de um modelo de regressão para determinar o Volume de Madeira Total da espécie florestal (Vochysia inundata ducke Quaruba, em função de suas características (Diâmetro à Altura do Peito (DAP, Idade e Altura. O modelo foi determinado utilizando-se a técnica estatística multivariada de Análise de Fatores com do Método das Componentes Principais via Rotação Ortogonal do Tipo Varimax para Extração dos Fatores, procurando contornar o problema da Multicolinearidade. Por fim desenvolveu-se um modelo de Regressão Linear Simples com base nos Escores Fatoriais. O modelo determinado apresentou-se de fácil interpretação e utilização, usando-se um fator e proporcionando um bom ajuste (R² = 96 % aos dados e uma boa capacidade preditiva. Ele atendeu a todas as suposições teóricas para sua existência e utilização.

  10. Adaptations to Short, Frequent Sessions of Endurance and Strength Training Are Similar to Longer, Less Frequent Exercise Sessions When the Total Volume Is the Same.

    Science.gov (United States)

    Kilen, Anders; Hjelvang, Line B; Dall, Niels; Kruse, Nanna L; Nordsborg, Nikolai B

    2015-11-01

    The hypothesis that the distribution of weekly training across several short sessions, as opposed to fewer longer sessions, enhances maximal strength gain without compromising maximal oxygen uptake was evaluated. Twenty-nine subjects completed an 8-week controlled parallel-group training intervention. One group ("micro training" [MI]: n = 21) performed nine 15-minute training sessions weekly, whereas a second group ("classical training" [CL]: n = 8) completed exactly the same training on a weekly basis but as three 45-minute sessions. For each group, each session comprised exclusively strength, high-intensity cardiovascular training or muscle endurance training. Both groups increased shuttle run performance (MI: 1,373 ± 133 m vs. 1,498 ± 126 m, p ≤ 0.05; CL: 1,074 ± 213 m vs. 1,451 ± 202 m, p training intervention. In conclusion, similar training adaptations can be obtained with short, frequent exercise sessions or longer, less frequent sessions where the total volume of weekly training performed is the same.

  11. Review of light--water reactor safety studies. Volume 3 of health and safety impacts of nuclear, geothermal, and fossil-fuel electric generation in California

    International Nuclear Information System (INIS)

    Nero, A.V.; Farnaam, M.R.K.

    1977-01-01

    This report summarizes and compares important studies of light-water nuclear reactor safety, emphasizing the Nuclear Regulatory Commission's Reactor Safety Study, work on risk assessment funded by the Electric Power Research Institute, and the Report of the American Physical Society study group on light-water reactor safety. These reports treat risk assessment for nuclear power plants and provide an introduction to the basic issues in reactor safety and the needs of the reactor safety research program. Earlier studies are treated more briefly. The report includes comments on the Reactor Safety Study. The manner in which these studies may be used and alterations which would increase their utility are discussed

  12. The conceptual flowsheet of effluent treatment during total gelation of uranium process for preparing ceramic UO2 particles of high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Quan Ying; Chen Xiaotong; Wang Yang; Liu Bing; Tang Yaping; Tang Chunhe

    2014-01-01

    Today, more and more people pay attention to the environmental protection and ecological environment. Along with the development of nuclear industry, many radioactive effluents may be discharged into environment, which can lead to the pollutions of water, atmosphere and soil. So radioactive effluents including low-activity and medium-level wastes solution treatments have been becoming one of significant subjects. High temperature gas-cooled reactor (HTR) is one of advanced nuclear reactors owing to its reliability, security and broad application in which the fabrication of spherical fuel element is a key technology. During the production of spherical fuel elements, the radioactive effluent treatment is necessary. Referring to the current treatment technologies and methods, the conceptual flowsheet of low-level radioactive effluent treatment during preparing spherical fuel elements was summarized which met the 'Zero Emission' demand. (authors)

  13. Corrosion of graphitic high temperature reactor materials in steam/helium mixtures at total pessures of 3-55 bar and temperatures of 900-1150 C (1173-1423K)

    International Nuclear Information System (INIS)

    Hinssen, H.K.; Loenissen, K.J.; Katscher, W.; Moormann, R.

    1993-03-01

    In course of accident examination for (HTR), experiments on the corrosion behavior of graphitic reactor materials in steam have been performed a total pressures of 3-55bar and temperatures of 900-1150 C (1173-1423K); these experiments and their evaluation are documented here. Reactor materials examined are the structure graphite V483T2 and the fuel element matrices A3-27 and A3-3. In all experiments, the steam partial pressure was 474mbar (inert gas helium). The dependence of reaction rates and density profiles on burn-off, total pressure and temperature has been examined. Experimental reaction rates depending on burn-off are fitted by theoretical curves, a procedure, which allows rate comparison for a well defined burn-off. Comparing rates as a function of total pressure, V483T2 shows a linear dependence on 1√p total , whereas for matrix materials a pressure independent rate was found for p total 4mm for A3-3. (orig.) [de

  14. Reactor core for LMFBR type reactors

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

    1987-01-01

    Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

  15. ESTIMATIVA DO VOLUME TOTAL DE MADEIRA EM ESPÉCIES DE EUCALIPTO A PARTIR DE IMAGENS DE SATÉLITE LANDSAT

    Directory of Open Access Journals (Sweden)

    Elias Fernando Berra

    2012-01-01

    Full Text Available Models relating spectral answers with biophysical parameters aim estimate variables, such as wood volume, without the necessity of frequent field measurements. The objective was to develop models to estimatewood volume by Landsat 5 TM images, supported by regional forest inventory data. The image was georeferenced and converted to spectral reflectance. After, the images-index NDVI (Normalized Difference Vegetation Index and SR (Simple Ratio was generated. The reflectance values of both spectral bands (TM1, TM2, TM3 e TM4 and indices (NDVI and SR was related with the wood volume. The biggest correlation with volume was with the NDVI and SR indices. The variables selection was made by Stepwise method, which returned three regression models as significant to explain the variation in volume. Finally,the best fitted model was selected (volume = -830,95 + 46,05 × (SR + 107,47 × (TM2, which was applied on the Landsat image where the pixels had started to represent the estimated volume in m³/ha on the Eucalyptus sp. production units. This model, significant at 95 % confidence level, explains 68 % of the wood volume variation.

  16. Shock absorber in combination with a nuclear reactor core structure

    International Nuclear Information System (INIS)

    Housman, J.J.

    1976-01-01

    This invention relates to the provision of shock absorbers for use in blind control rod passages of a nuclear reactor core structure which are not subject to degradation. The shock absorber elements are made of a porous brittle carbonaceous material, a porous brittle ceramic material, or a porous brittle refractory oxide and have a void volume of between 30% and 70% of the total volume of the element for energy absorption by fracturing due to impact loading by a control rod. (UK)

  17. Twenty-third water reactor safety information meeting: Volume 2, Human factors research; Advanced I and C hardware and software; Severe accident research; Probabilistic risk assessment topics; Individual plant examination: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1996-03-01

    This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 2, present topics in human factors research, advanced instrumentation and control hardware and software, severe accident research, probabilistic risk assessment, and individual plant examination. Individual papers have been cataloged separately.

  18. Sandia Pulse Reactor-IV Project

    International Nuclear Information System (INIS)

    Reuscher, J.A.

    1983-01-01

    Sandia National Laboratories has developed, designed and operated fast burst reactors for over 20 years. These reactors have been used for a variety of radiation effects programs. During this period, programs have required larger irradiation volumes primarily to expose complex electronic systems to postulated threat environments. As experiment volumes increased, a new reactor was built so that these components could be tested. The Sandia Pulse Reactor-IV is a logical evolution of the two decades of fast burst reactor development at Sandia

  19. Automatic total kidney volume measurement on follow-up magnetic resonance images to facilitate monitoring of autosomal dominant polycystic kidney disease progression.

    Science.gov (United States)

    Kline, Timothy L; Korfiatis, Panagiotis; Edwards, Marie E; Warner, Joshua D; Irazabal, Maria V; King, Bernard F; Torres, Vicente E; Erickson, Bradley J

    2016-02-01

    Renal imaging examinations provide high-resolution information about the anatomic structure of the kidneys and are used to measure total kidney volume (TKV) in autosomal dominant polycystic kidney disease (ADPKD) patients. TKV has become the gold-standard image biomarker for ADPKD progression at early stages of the disease and is used in clinical trials to characterize treatment efficacy. Automated methods to segment the kidneys and measure TKV are desirable because of the long time requirement for manual approaches such as stereology or planimetry tracings. However, ADPKD kidney segmentation is complicated by a number of factors, including irregular kidney shapes and variable tissue signal at the kidney borders. We describe an image processing approach that overcomes these problems by using a baseline segmentation initialization to provide automatic segmentation of follow-up scans obtained years apart. We validated our approach using 20 patients with complete baseline and follow-up T1-weighted magnetic resonance images. Both manual tracing and stereology were used to calculate TKV, with two observers performing manual tracings and one observer performing repeat tracings. Linear correlation and Bland-Altman analysis were performed to compare the different approaches. Our automated approach measured TKV at a level of accuracy (mean difference ± standard error = 0.99 ± 0.79%) on par with both intraobserver (0.77 ± 0.46%) and interobserver variability (1.34 ± 0.70%) of manual tracings. All approaches had excellent agreement and compared favorably with ground-truth manual tracing with interobserver, stereological and automated approaches having 95% confidence intervals ∼ ± 100 mL. Our method enables fast, cost-effective and reproducible quantification of ADPKD progression that will facilitate and lower the costs of clinical trials in ADPKD and other disorders requiring accurate, longitudinal kidney quantification. In addition, it will hasten the routine use of

  20. Baseline Metabolic Tumor Volume and Total Lesion Glycolysis Are Associated With Survival Outcomes in Patients With Locally Advanced Pancreatic Cancer Receiving Stereotactic Body Radiation Therapy

    Energy Technology Data Exchange (ETDEWEB)

    Dholakia, Avani S. [Department of Radiation Oncology and Molecular Radiation Sciences, Johns Hopkins University School of Medicine, Baltimore, Maryland (United States); Chaudhry, Muhammad; Leal, Jeffrey P. [Department of Radiology, Johns Hopkins University School of Medicine, Baltimore, Maryland (United States); Chang, Daniel T. [Department of Radiation Oncology, Stanford University School of Medicine, Stanford, California (United States); Raman, Siva P. [Department of Radiology, Johns Hopkins University School of Medicine, Baltimore, Maryland (United States); Hacker-Prietz, Amy [Department of Radiation Oncology and Molecular Radiation Sciences, Johns Hopkins University School of Medicine, Baltimore, Maryland (United States); Su, Zheng; Pai, Jonathan [Department of Radiation Oncology, Stanford University School of Medicine, Stanford, California (United States); Oteiza, Katharine E.; Griffith, Mary E. [Department of Radiation Oncology and Molecular Radiation Sciences, Johns Hopkins University School of Medicine, Baltimore, Maryland (United States); Wahl, Richard L. [Department of Radiology, Johns Hopkins University School of Medicine, Baltimore, Maryland (United States); Tryggestad, Erik [Department of Radiation Oncology and Molecular Radiation Sciences, Johns Hopkins University School of Medicine, Baltimore, Maryland (United States); Pawlik, Timothy [Department of Surgery, Johns Hopkins University School of Medicine, Baltimore, Maryland (United States); Laheru, Daniel A. [Department of Oncology, Johns Hopkins University School of Medicine, Baltimore, Maryland (United States); Wolfgang, Christopher L. [Department of Surgery, Johns Hopkins University School of Medicine, Baltimore, Maryland (United States); Koong, Albert C. [Department of Radiation Oncology, Stanford University School of Medicine, Stanford, California (United States); and others

    2014-07-01

    Purpose: Although previous studies have demonstrated the prognostic value of positron emission tomography (PET) parameters in other malignancies, the role of PET in pancreatic cancer has yet to be well established. We analyzed the prognostic utility of PET for patients with locally advanced pancreatic cancer (LAPC) undergoing fractionated stereotactic body radiation therapy (SBRT). Materials and Methods: Thirty-two patients with LAPC in a prospective clinical trial received up to 3 doses of gemcitabine, followed by 33 Gy in 5 fractions of 6.6 Gy, using SBRT. All patients received a baseline PET scan prior to SBRT (pre-SBRT PET). Metabolic tumor volume (MTV), total lesion glycolysis (TLG), and maximum and peak standardized uptake values (SUV{sub max} and SUV{sub peak}) on pre-SBRT PET scans were calculated using custom-designed software. Disease was measured at a threshold based on the liver SUV, using the equation Liver{sub mean} + [2 × Liver{sub sd}]. Median values of PET parameters were used as cutoffs when assessing their prognostic potential through Cox regression analyses. Results: Of the 32 patients, the majority were male (n=19, 59%), 65 years or older (n=21, 66%), and had tumors located in the pancreatic head (n=27, 84%). Twenty-seven patients (84%) received induction gemcitabine prior to SBRT. Median overall survival for the entire cohort was 18.8 months (95% confidence interval [CI], 15.7-22.0). An MTV of 26.8 cm{sup 3} or greater (hazard ratio [HR] 4.46, 95% CI 1.64-5.88, P<.003) and TLG of 70.9 or greater (HR 3.08, 95% CI 1.18-8.02, P<.021) on pre-SBRT PET scan were associated with inferior overall survival on univariate analysis. Both pre-SBRT MTV (HR 5.13, 95% CI 1.19-22.21, P=.029) and TLG (HR 3.34, 95% CI 1.07-10.48, P=.038) remained independently associated with overall survival in separate multivariate analyses. Conclusions: Pre-SBRT MTV and TLG are potential predictive factors for overall survival in patients with LAPC and may assist in

  1. Baseline Metabolic Tumor Volume and Total Lesion Glycolysis Are Associated With Survival Outcomes in Patients With Locally Advanced Pancreatic Cancer Receiving Stereotactic Body Radiation Therapy

    International Nuclear Information System (INIS)

    Dholakia, Avani S.; Chaudhry, Muhammad; Leal, Jeffrey P.; Chang, Daniel T.; Raman, Siva P.; Hacker-Prietz, Amy; Su, Zheng; Pai, Jonathan; Oteiza, Katharine E.; Griffith, Mary E.; Wahl, Richard L.; Tryggestad, Erik; Pawlik, Timothy; Laheru, Daniel A.; Wolfgang, Christopher L.; Koong, Albert C.

    2014-01-01

    Purpose: Although previous studies have demonstrated the prognostic value of positron emission tomography (PET) parameters in other malignancies, the role of PET in pancreatic cancer has yet to be well established. We analyzed the prognostic utility of PET for patients with locally advanced pancreatic cancer (LAPC) undergoing fractionated stereotactic body radiation therapy (SBRT). Materials and Methods: Thirty-two patients with LAPC in a prospective clinical trial received up to 3 doses of gemcitabine, followed by 33 Gy in 5 fractions of 6.6 Gy, using SBRT. All patients received a baseline PET scan prior to SBRT (pre-SBRT PET). Metabolic tumor volume (MTV), total lesion glycolysis (TLG), and maximum and peak standardized uptake values (SUV max and SUV peak ) on pre-SBRT PET scans were calculated using custom-designed software. Disease was measured at a threshold based on the liver SUV, using the equation Liver mean + [2 × Liver sd ]. Median values of PET parameters were used as cutoffs when assessing their prognostic potential through Cox regression analyses. Results: Of the 32 patients, the majority were male (n=19, 59%), 65 years or older (n=21, 66%), and had tumors located in the pancreatic head (n=27, 84%). Twenty-seven patients (84%) received induction gemcitabine prior to SBRT. Median overall survival for the entire cohort was 18.8 months (95% confidence interval [CI], 15.7-22.0). An MTV of 26.8 cm 3 or greater (hazard ratio [HR] 4.46, 95% CI 1.64-5.88, P<.003) and TLG of 70.9 or greater (HR 3.08, 95% CI 1.18-8.02, P<.021) on pre-SBRT PET scan were associated with inferior overall survival on univariate analysis. Both pre-SBRT MTV (HR 5.13, 95% CI 1.19-22.21, P=.029) and TLG (HR 3.34, 95% CI 1.07-10.48, P=.038) remained independently associated with overall survival in separate multivariate analyses. Conclusions: Pre-SBRT MTV and TLG are potential predictive factors for overall survival in patients with LAPC and may assist in tailoring therapy

  2. Biodegradation of chlorinated hydrocarbons in a vapor phase reactor

    International Nuclear Information System (INIS)

    Ensley, B.D.

    1992-01-01

    A bench scale gas lift loop reactor was constructed to evaluate the feasibility of trichloroethylene (TCE) degradative microorganisms being used to treat TCE contaminated air. Two different microorganisms were used as biocatalysts in this reactor. After proper operating conditions were established for use of this reactor/biocatalyst combination, both microorganisms could degrade 95% of inlet TCE at air flow rates of up to 3% of the total reactor volume per minute. TCE concentrations of between 300 μg/L (60ppmv) and 3000 μg/L (600 ppmv) were degraded with 95% or better efficiency. Preliminary economic evaluations suggest that bioremediation may be the low cost alternative for treating certain TCE contaminated air streams and field trials of a scaled-up reactor system based on this technology are currently underway

  3. The total right/left-volume index: a new and simplified cardiac magnetic resonance measure to evaluate the severity of Ebstein anomaly of the tricuspid valve: a comparison with heart failure markers from various modalities.

    Science.gov (United States)

    Hösch, Olga; Sohns, Jan Martin; Nguyen, Thuy-Trang; Lauerer, Peter; Rosenberg, Christina; Kowallick, Johannes Tammo; Kutty, Shelby; Unterberg, Christina; Schuster, Andreas; Faßhauer, Martin; Staab, Wieland; Paul, Thomas; Lotz, Joachim; Steinmetz, Michael

    2014-07-01

    The classification of clinical severity of Ebstein anomaly still remains a challenge. The aim of this study was to focus on the interaction of the pathologically altered right heart with the anatomically-supposedly-normal left heart and to derive from cardiac magnetic resonance (CMR) a simple imaging measure for the clinical severity of Ebstein anomaly. Twenty-five patients at a mean age of 26±14 years with unrepaired Ebstein anomaly were examined in a prospective study. Disease severity was classified using CMR volumes and functional measurements in comparison with heart failure markers from clinical data, ECG, laboratory and cardiopulmonary exercise testing, and echocardiography. All examinations were completed within 24 hours. A total right/left-volume index was defined from end-diastolic volume measurements in CMR: total right/left-volume index=(RA+aRV+fRV)/(LA+LV). Mean total right/left-volume index was 2.6±1.7 (normal values: 1.1±0.1). This new total right/left-volume index correlated with almost all clinically used biomarkers of heart failure: brain natriuretic peptide (r=0.691; P=0.0003), QRS (r=0.432; P=0.039), peak oxygen consumption/kg (r=-0.479; P=0.024), ventilatory response to carbon dioxide production at anaerobic threshold (r=0.426; P=0.048), the severity of tricuspid regurgitation (r=0.692; P=0.009), tricuspid valve offset (r=0.583; P=0.004), and tricuspid annular plane systolic excursion (r=0.554; P=0.006). Previously described severity indices ([RA+aRV]/[fRV+LA+LV]) and fRV/LV end-diastolic volume corresponded only to some parameters. In patients with Ebstein anomaly, the easily acquired index of right-sided to left-sided heart volumes from CMR correlated well with established heart failure markers. Our data suggest that the total right/left-volume index should be used as a new and simplified CMR measure, allowing more accurate assessment of disease severity than previously described scoring systems. © 2014 American Heart Association, Inc.

  4. RA Reactor operation and maintenance (I-IX), Part IV, Task 3.08/04, Refurbishment of the RA reactor

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-12-01

    This volume contains reports describing maintenance and repair work of the RA reactor instrumentation, equipment of the reactor dosimetry control system, and equipment for regulation and control systems

  5. Engineering and planning for reactor 105-C interim safe storage project subcontract no. 0100C-SC-G0001 conceptual design report. Volume 1

    International Nuclear Information System (INIS)

    1996-04-01

    The 105-C Reactor, one of eight surplus production reactors at the Hanford Site, has been proposed by the U.S. Department of Energy, Richland, Operations Office to be the first large-scale technology demonstration project in the decontamination and decommissioning (D ampersand D) focus area as part of the project for dismantlement and interim safe storage. The 105-C Reactor will be placed in an interim safe storage condition, then undergo the decontamination and decommissioning phase. After D ampersand D, the reactor will be placed in long- term safe storage. This report provides the conceptual design for these activities

  6. Chronic Use of Aspirin and Total White Matter Lesion Volume: Results from the Women's Health Initiative Memory Study of Magnetic Resonance Imaging Study.

    Science.gov (United States)

    Holcombe, Andrea; Ammann, Eric; Espeland, Mark A; Kelley, Brendan J; Manson, JoAnn E; Wallace, Robert; Robinson, Jennifer

    2017-10-01

    To investigate the relationship between aspirin and subclinical cerebrovascular heath, we evaluated the effect of chronic aspirin use on white matter lesions (WML) volume among women. Chronic aspirin use was assessed in 1365 women who participated in the Women's Health Initiative Memory Study of Magnetic Resonance Imaging. Differences in WML volumes between aspirin users and nonusers were assessed with linear mixed models. A number of secondary analyses were performed, including lobe-specific analyses, subgroup analyses based on participants' overall risk of cerebrovascular disease, and a dose-response relationship analysis. The mean age of the women at magnetic resonance imaging examination was 77.6 years. Sixty-one percent of participants were chronic aspirin users. After adjusting for demographic variables and comorbidities, chronic aspirin use was nonsignificantly associated with 4.8% (95% CI: -6.8%, 17.9%) larger WML volumes. These null findings were confirmed in secondary and sensitivity analyses, including an active comparator evaluation where aspirin users were compared to users of nonaspirin nonsteroidal anti-inflammatory drugs or acetaminophen. There was a nonsignificant difference in WML volumes between aspirin users and nonusers. Further, our results suggest that chronic aspirin use may not have a clinically significant effect on WML volumes in women. Published by Elsevier Inc.

  7. Mechanical spectral shift reactor

    International Nuclear Information System (INIS)

    Doshi, P.K.; George, R.A.; Dollard, W.J.

    1982-01-01

    A mechanical spectral shift arrangement for controlling a nuclear reactor includes a plurality of reactor coolant displacer members which are inserted into a reactor core at the beginning of the core life to reduce the volume of reactor coolant-moderator in the core at start-up. However, as the reactivity of the core declines with fuel depletion, selected displacer members are withdrawn from the core at selected time intervals to increase core moderation at a time when fuel reactivity is declining. (author)

  8. Reactor operation environmental information document

    Energy Technology Data Exchange (ETDEWEB)

    Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

    1989-12-01

    The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

  9. BWR type reactor

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1983-01-01

    Purpose : To flatten the radial power distribution in the reactor core thereby improve the thermal performance of the reactor core by making the moderator-fuel ratio of fuel assemblies different depending on their position in the reactor core. Constitution : The volume of fuels disposed in the peripheral area of the reactor core is decreased by the increase of the volume of moderators in fuel assemblies disposed in the peripheral area of the reactor core to thereby make the moderator-fuel volume greater in the peripheral area than that in the central area. The moderator-fuel ratio adjustment is attained by making the number of water rods greater, decreasing the diameter of fuel pellets or decreasing the number of fuel pins in fuel assemblies disposed at the peripheral area of the reactor core as compared with fuel assemblies disposed at the central area of the reactor core. In this way, the infinite multiplication factors of fuels can be increased to thereby improve the reactor core performance. (Aizawa, K.)

  10. Some local dilution transient in a pressurized water reactor

    International Nuclear Information System (INIS)

    Jacobson, S.

    1989-01-01

    Reactivity accidents are important in the safety analysis of a pressurized water reactor. In this anlysis ejected control rod, steam line break, start of in-active loop and boron dilution accidents are usually dealt with. However, in the analysis is not included what reactivity excursions might happen when a zone,depleted of boron passes the reactor core. This thesis investigates during what operation and emergency conditions diluted zones might exist in a pressurized water reactor and what should be the maximum volumes for then. The limiting transport means are also established in terms of reactivty addition, for the depleted zones. In order to describe the complicated mixing process in the reactor vessel during such a transportation, a typical 3-loop reactor vessel has been modulated by means of TRAC-PF1's VESSEL component. Three cases have been analysed. In the first case the reactor is in a cold condition and the ractor coolant has boron concentration of 2000 ppm. To the reactor vessel is injected an clean water colume of 14 m 3 . In the two other cases the reactor is close to hot shutdown and borated to 850 ppm. To the reactor vessel is added 41 and 13 m 3 clean water, respectively. In the thesis is shown what spatial distribution the depleted zone gets when passing through the reactor vessel in the three cases. The boron concentration in the first case did not decrease the values which would bring the reactor to critical condition. For case two was shown by means of TRAC's point kinetics model that the reactor reaches prompt criticality after 16.03 seconds after starting of the reactor coolant pump. Another prompt criticality occured two seconds later. The total energy developed during the two power escalations were about 55 GJ. A comparision with the criteria used to evaluate the ejected control rod reactivity transient showed that none of these criteria were exceeded. (64 figs.)

  11. Total and regional brain volumes in a population-based normative sample from 4 to 18 years: the NIH MRI Study of Normal Brain Development.

    Science.gov (United States)

    2012-01-01

    Using a population-based sampling strategy, the National Institutes of Health (NIH) Magnetic Resonance Imaging Study of Normal Brain Development compiled a longitudinal normative reference database of neuroimaging and correlated clinical/behavioral data from a demographically representative sample of healthy children and adolescents aged newborn through early adulthood. The present paper reports brain volume data for 325 children, ages 4.5-18 years, from the first cross-sectional time point. Measures included volumes of whole-brain gray matter (GM) and white matter (WM), left and right lateral ventricles, frontal, temporal, parietal and occipital lobe GM and WM, subcortical GM (thalamus, caudate, putamen, and globus pallidus), cerebellum, and brainstem. Associations with cross-sectional age, sex, family income, parental education, and body mass index (BMI) were evaluated. Key observations are: 1) age-related decreases in lobar GM most prominent in parietal and occipital cortex; 2) age-related increases in lobar WM, greatest in occipital, followed by the temporal lobe; 3) age-related trajectories predominantly curvilinear in females, but linear in males; and 4) small systematic associations of brain tissue volumes with BMI but not with IQ, family income, or parental education. These findings constitute a normative reference on regional brain volumes in children and adolescents.

  12. Best-practices guidelines for L2PSA development and applications. Volume 2 - Best practices for the Gen II PWR, Gen II BWR L2PSAs. Extension to Gen III reactors

    International Nuclear Information System (INIS)

    Raimond, E.; Durin, T.; Rahni, N.; Meignen, R.; Cranga, M.; Pichereau, F.; Bentaib, A.; Guigueno, Y.; Loeffler, H.; Mildenberger, O.; Lajtha, G.; Santamaria, C.S.; Dienstbier, J.; Rydl, A.; Holmberg, J.E.; Lindholm, I.; Maennistoe, I.; Pauli, E.M.; Dirksen, G.; Grindon, L.; Peers, K.; Hulqvist, G.; Parozzi, F.; Polidoro, F.; Cazzoli, E.; Vitazkova, J.; Burgazzi, L.; Oury, L.; Ngatchou, C.; Siltanen, S.; Niemela, I.; Routamo, T.; Helstroem, P.; Bassi, C.; Brinkman, H.; Seidel, A.; Schubert, B.; Wohlstein, R.; Guentay, S.; Vincon, L.

    2010-01-01

    The objective of this coordinated action was to develop best practice guidelines for the performance of Level 2 PSA methodologies with a view of harmonisation at EU level and to allow meaningful and practical uncertainty evaluations in a Level 2 PSA. Specific relationships with community in charge of nuclear reactor safety (utilities, safety authorities, vendors, and research or services companies) have been established in order to define the current needs in terms of guidelines for level 2 PSA development and applications. An international workshop was organised in Hamburg, with the support of VATTENFALL, in November 2008. The level 2 PSA experts from the ASAMPSA2 project partners have proposed some guidelines for the development and application of L2PSA based on their experience and on information available from international cooperation (EC Severe Accident network of Excellence - SARNET, IAEA standards, OECD-NEA publications and workshop) or open literature. The number of technical issues addressed in the guideline is very large and all are not covered with the same relevancy in the first version of the guideline. This version is submitted for external review in November 2010 by severe accident experts and PSA, especially, from SARNET and OECD-NEA members. The feedback of the external review will be dis cussed during an international open works hop planned in March 2011 and all outcomes will be taken into consideration in the final version of this guideline (June 2011). The guideline includes 3 volumes: - Volume 1 - General considerations on L2PSA. - Volume 2 - Technical recommendations for Gen II and III reactors. - Volume 3 - Specific considerations for future reactor (Gen IV). The recommendations formulated in the guideline should not be considered as 'mandatory' but should help the L2PSA developers to achieve high quality studies with limited time and resources. It may also help the L2PSA reviewers by positioning one specific study in comparison with some

  13. A review of the internal components of the second generation of Swedish BWRs in perspective of their importance for the total safety. A diploma work in reactor technology

    International Nuclear Information System (INIS)

    Appelgren, S.; Eriksson, Stefan

    1999-03-01

    An investigation has been done of the second generation of Swedish BWRs, Barsebaeck 1 and 2, and Oskarshamn 2, concerning the vessel internals and theirs significance for the reactor safety. The purpose with this pilot study has been to produce a support for the course of action and to be a source of information for more detailed analyses of the vessel internals. A number of accident scenarios have been depicted and discussed regarding how they might occur and what the consequences might be. It is postulated that they start on account of some vessel internals failing. To be able to develop these scenarios it was necessary to collect and go through a relative large number of analyses and calculations. These have consisted of design conditions, calculation of stress and damage reports. In design conditions are included the maximum loads that a component expect to be subjected to in the course of different postulated averages. The design conditions are the input to the calculation of stress. The damage reports treat and analyse the damages that the internals have been exposed to during the years. For each scenario that has been treated, a judgement has been done about why or why not it is probable to happen. The authors do not claim to have made a probability study along the lines that are commonly accepted. The internal parts that have been the subject for the study are the core head, the feed water spargers, the steam dryers, the core shroud and the core shroud support. Below are the results with argumentations and recommendations. Core head: the core head has the behaviour that contribute most to the complexity of the scenarios. Initiators of this kind of scenarios are postulated weaknesses in the extensions of the bolts fastening the shroud head to the core shroud. A collapse of the extensions of the bolts fastening the shroud head to the core shroud will have a great impact on the reactor safety. Very likely it would lead to absent core cooling and absent

  14. Some basic thermohydraulic calculation methods for the analysis of pressure transients in a multicompartment total containment enclosing a breached water reactor circuit

    International Nuclear Information System (INIS)

    Porter, W.H.L.

    1976-05-01

    This paper gives an appreciation and commentary of the basic calculation methods under development at AEE Winfrith for the analysis of multicompartment total containments. The assumptions introduced and the effects of their variation are important in establishing a parametric survey of the range of possible conditions which the containment may be required to meet. These aspects of the performance will be discussed as each individual factor in the train of events is examined in turn. (U.K.)

  15. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

    Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

  16. Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 4. Alternatives for waste isolation and disposal

    International Nuclear Information System (INIS)

    1976-05-01

    Volume IV of the five-volume report contains information on alternatives for final storage and disposal of radioactive wastes. Section titles include: basic concepts for geologic isolation; geologic storage alternatives; geologic disposal alternatives; extraterrestrial disposal; and, transmutation

  17. Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 4. Alternatives for waste isolation and disposal

    Energy Technology Data Exchange (ETDEWEB)

    1976-05-01

    Volume IV of the five-volume report contains information on alternatives for final storage and disposal of radioactive wastes. Section titles include: basic concepts for geologic isolation; geologic storage alternatives; geologic disposal alternatives; extraterrestrial disposal; and, transmutation. (JGB)

  18. Spectral shift reactor control method

    International Nuclear Information System (INIS)

    Impink, A.J.

    1982-01-01

    A method of operating a nuclear reactor having a core and coolant displacer elements arranged in the core where there is established a reactor coolant temperature set point at which it is desired to operate the reactor and first reactor coolant temperature band limits within which the set point is characterized. The reactor coolant displacer elements are moved relative to the reactor core for adjusting the volume of reactor coolant in the core as the reactor coolant temperature approaches the first band limits to maintain the reactor coolant temperature near the set point and within the first band limits. The reactivity charges associated with movement of respective coolant displacer element clusters is calculated and compared with a calculated derived reactivity charge in order to select the cluster to be moved. (author)

  19. Megawatt Class Nuclear Space Power Systems (MCNSPS) conceptual design and evaluation report. Volume 2, technologies 1: Reactors, heat transport, integration issues

    Science.gov (United States)

    Wetch, J. R.

    1988-01-01

    The objectives of the Megawatt Class Nuclear Space Power System (MCNSPS) study are summarized and candidate systems and subsystems are described. Particular emphasis is given to the heat rejection system and the space reactor subsystem.

  20. Preliminary design studies of the draining tanks for the Molten Salt Fast Reactor

    International Nuclear Information System (INIS)

    Merle-Lucotte, E.; Allibert, M.; Heuer, D.; Brovchenko, M.; Laureau, A.; Ghetta, V.; Rubiolo, P.

    2014-01-01

    reactor called the Molten Salt Fast Reactor (MSFR). The reference MSFR design is a 3000 MWth reactor with a total fuel salt volume of 18 m3, operated at a mean fuel temperature of 750 deg. C. The first confinement barrier of the reactor includes a salt draining system. In case of a planned reactor shut down or in case of accidents leading to an excessive increase of the temperature in the fuel circuit, the fuel configuration may be changed passively by gravitational draining of the fuel salt in dedicated draining tank located under the reactor and designed to provide adequate reactivity margins while insuring a passive cooling of the fuel salt to extract the residual heat from the short to the long term. The present preliminary assessment of this sub-critical draining system has been performed to identify the physical constraints and to give some orders of magnitude of characteristic time periods (authors)

  1. Defense Manpower Commission Staff Studies and Supporting Papers. Volume 2. The Total Force and Its Manpower Requirements Including Overviews of Each Service

    Science.gov (United States)

    1976-05-01

    J^^’.Si*!** \\ ir..’’T^-.’-T*TSfn titoa i iMBi’M, OTTŕ" ,^~" fraCk k^«^;-<^»J,..;^.a,L.^t»^^ri^fc ft WBMyLmH’.’JW*^Hi,.J , Jl,l|llliln|Kli|Pffl...also develop a means to inspect tube internals to insure they are clean 11. Develop a deep tank, high volume, high head hydraulic driven pump to...and procedures that have been implemented at this Command and havs en- hanced productivity are: - Power floor cleaners Pneumatic/ hydraulic ram tire

  2. One piece reactor removal

    International Nuclear Information System (INIS)

    Chia, Wei-Min; Wang, Song-Feng

    1993-01-01

    The strategy of Taiwan Research Reactor Renewal plan is to remove the old reactor block with One Piece Reactor Removal (OPRR) method for installing a new research reactor in original building. In this paper, the engineering design of each transportation works including the work method, the major equipments, the design policy and design criteria is described and discussed. In addition, to ensure the reactor block is safety transported for storage and to guarantee the integrity of reactor base mat is maintained for new reactor, operation safety is drawn special attention, particularly under seismic condition, to warrant safe operation of OPRR. ALARA principle and Below Regulatory Concern (BRC) practice were also incorporated in the planning to minimize the collective dose and the total amount of radioactive wastes. All these activities are introduced in this paper. (J.P.N.)

  3. An evaluation of state-of-the-art two-velocity two-phase flow models and their applicability to nuclear reactor transient analysis. Volume 3. Data comparisons. Final report

    International Nuclear Information System (INIS)

    McFadden, J.H.; Lyczkowski, R.W.; Niederauer, G.F.

    1976-02-01

    A state-of-the-art review is conducted in order to provide the nuclear industry with a publicly available assessment of two-velocity thermal-hydraulic models and their applicability to nuclear reactor technology. The two major objectives of this state-of-the-art evaluation were: (1) document the basic theory in a consistent self-contained report; and (2) apply a prototype 'two-velocity' code (UVUT) to a limited number of separate effect tests. Volume 3 presents the data comparisons

  4. Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

    International Nuclear Information System (INIS)

    Monteleone, S.

    1994-04-01

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors

  5. Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [Brookhaven National Lab., Upton, NY (United States)] [comp.

    1994-04-01

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors.

  6. Reactor core of FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Ichimiya, Masakazu.

    1994-01-01

    A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

  7. The Influence of RSG-GAS Primary Pump Operation Concerning the Rise Water Level of Reactor Pool in 15 MW Reactor Power

    International Nuclear Information System (INIS)

    Djunaidi

    2004-01-01

    The expansion of air volume in the delay chamber shows in rise water level of reactor pool during the operation. The rises of water level in the reactor pool is not quite from the expansion of air volume in the delay chamber, but some influence the primary pump operation. The purpose evaluated of influence primary pump is to know the influence primary pump power concerning the rise water level during the reactor operation. From the data collection during 15 MW power operation in the last core 42 the influence of primary pump operation concerning the rise water level in the reactor pool is 34.48 % from the total increased after operation during 12 days. (author)

  8. Total Thyroidectomy

    Directory of Open Access Journals (Sweden)

    Lopez Moris E

    2016-06-01

    Full Text Available Total thyroidectomy is a surgery that removes all the thyroid tissue from the patient. The suspect of cancer in a thyroid nodule is the most frequent indication and it is presume when previous fine needle puncture is positive or a goiter has significant volume increase or symptomes. Less frequent indications are hyperthyroidism when it is refractory to treatment with Iodine 131 or it is contraindicated, and in cases of symptomatic thyroiditis. The thyroid gland has an important anatomic relation whith the inferior laryngeal nerve and the parathyroid glands, for this reason it is imperative to perform extremely meticulous dissection to recognize each one of these elements and ensure their preservation. It is also essential to maintain strict hemostasis, in order to avoid any postoperative bleeding that could lead to a suffocating neck hematoma, feared complication that represents a surgical emergency and endangers the patient’s life.It is essential to run a formal technique, without skipping steps, and maintain prudence and patience that should rule any surgical act.

  9. Final generic environmental statement on the use of recycle plutonium in mixed oxide fuel in light water cooled reactors. Volume 2

    International Nuclear Information System (INIS)

    1976-08-01

    This environmental statement assesses the impacts of the implementation of plutonium recycle in the LWR industry. It is based on assumptions that are intended to reflect conservatively an acceptable level of the application of current technology. It is not intended to be a representation of the ''as low as reasonably achievable'' (ALARA) philosophy. This generic environmental statement discusses the anticipated effects of recycling plutonium in light water nuclear power reactors. It is based on about 30 years of experience with the element in the context of a projected light water nuclear power industry that is already substantial. A background perspective on plutonium, its safety, and its recycling as a reactor fuel is presented

  10. Volumetry based biomarker speed of growth: Quantifying the change of total tumor volume in whole-body magnetic resonance imaging over time improves risk stratification of smoldering multiple myeloma patients.

    Science.gov (United States)

    Wennmann, Markus; Kintzelé, Laurent; Piraud, Marie; Menze, Bjoern H; Hielscher, Thomas; Hofmanninger, Johannes; Wagner, Barbara; Kauczor, Hans-Ulrich; Merz, Maximilian; Hillengass, Jens; Langs, Georg; Weber, Marc-André

    2018-05-18

    The purpose of this study was to improve risk stratification of smoldering multiple myeloma patients, introducing new 3D-volumetry based imaging biomarkers derived from whole-body MRI. Two-hundred twenty whole-body MRIs from 63 patients with smoldering multiple myeloma were retrospectively analyzed and all focal lesions >5mm were manually segmented for volume quantification. The imaging biomarkers total tumor volume, speed of growth (development of the total tumor volume over time), number of focal lesions, development of the number of focal lesions over time and the recent imaging biomarker '>1 focal lesion' of the International Myeloma Working Group were compared, taking 2-year progression rate, sensitivity and false positive rate into account. Speed of growth, using a cutoff of 114mm 3 /month, was able to isolate a high-risk group with a 2-year progression rate of 82.5%. Additionally, it showed by far the highest sensitivity in this study and in comparison to other biomarkers in the literature, detecting 63.2% of patients who progress within 2 years. Furthermore, its false positive rate (8.7%) was much lower compared to the recent imaging biomarker '>1 focal lesion' of the International Myeloma Working Group. Therefore, speed of growth is the preferable imaging biomarker for risk stratification of smoldering multiple myeloma patients.

  11. Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 3. Alternatives for interim storage and transportation

    International Nuclear Information System (INIS)

    1976-05-01

    Volume III of the five-volume report contains information on alternatives for interim storage and transportation. Section titles are: interim storage of spent fuel elements; interim storage of chop-leach fuel bundle residues; tank storage of high-level liquid waste; interim storage of solid non-high-level wastes; interim storage of solidified high-level waste; and, transportation alternatives

  12. [Fuel Rod Consolidation Project]: The estimated total life cycle cost for the 30-year operation of prototypical consolidation demonstration equipment: Volume 4, Phase 2

    International Nuclear Information System (INIS)

    1987-01-01

    The Total Life Cycle Costs have been developed for the construction, operation and decommissioning of a single line of hot-cell-enclosed production consolidation equipment operating on spent fuel at the rate of 750 MTU/year for 30 years. The cost estimate is for a single production line that is part of an overall facility at either a Monitored Retrievable Storage or a Repository facility. This overall facility would include other capabilities and possibly other consolidation lines. However, no costs were included in the cost estimate for other portions of the plant, except that staff costs include an overhead charge that reflects the overhead support services in an overall facility

  13. Nuclear reactor

    International Nuclear Information System (INIS)

    Jolly, R.

    1979-01-01

    The support grid for the fuel rods of a liquid metal cooled fast breeder reactor has a regular hexagonal contour and contains a large number of unit cells arranged honeycomb fashion. The totality of these cells make up a hexagonal shape. The grid contains a number of strips of material, and there is a window in each of three sidewalls staggered by one sidewall. The other sidewalls have embossed protrusions, thus generating a guide lining or guide bead. The windows reduce the rigidity of the areas in the middle between the ends of the cells. (DG) [de

  14. Production of Radioisotopes in Pakistan Research Reactor: Past, Present and Future

    International Nuclear Information System (INIS)

    Mushtaq, A.

    2013-01-01

    Radioisotope production to service different sectors of economic significance constitutes an important ongoing activity of many national nuclear programs. Radioisotopes, formed by nuclear reactions on targets in a reactor or cyclotron, require further processing in almost all cases to obtain them in a form suitable for use. The availability of short-lived radionuclides from radionuclide generators provides an inexpensive and convenient alternative to in-house radioisotope production facilities such as cyclotrons and reactors. The reactor offers large volume for irradiation, simultaneous irradiation of several samples, economy of production and possibility to produce a wide variety of radioisotopes. The accelerator-produced isotopes relatively constitute a smaller percentage of total use. (author)

  15. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  16. RA reactor operation and maintenance

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-02-01

    This volume includes the final report on RA reactor operation and utilization of the experimental facilities in 1962, detailed analysis of the system for heavy water distillation and calibration of the system for measuring the activity of the air

  17. Reactor operation environmental information document

    Energy Technology Data Exchange (ETDEWEB)

    Bauer, L.R.; Hayes, D.W.; Hunter, C.H.; Marter, W.L.; Moyer, R.A.

    1989-12-01

    This volume is a reactor operation environmental information document for the Savannah River Plant. Topics include meteorology, surface hydrology, transport, environmental impacts, and radiation effects. 48 figs., 56 tabs. (KD)

  18. Low-level radioactive waste from commercial nuclear reactors. Volume 3. Bibliographic abstracts of significant source references. Part 2. Bibliography for treatment, storage, disposal and transportation regulatory constraints

    Energy Technology Data Exchange (ETDEWEB)

    Jolley, R.L.; Rodgers, B.R.

    1986-05-01

    The overall task of this program was to provide an assessment of currently available technology for treating commercial low-level radioactive waste (LLRW), to initiate development of a methodology for choosing one technology for a given application, and to identify research needed to improve current treatment techniques and decision methodology. The resulting report is issued in four volumes. Volume 3 of this series is a collection of abstracts of most of the reference documents used for this study. Because of the large volume of literature, the abstracts have been printed in two separate parts. Federal, state, and local regulations affect the decision process for selecting technology applications. Regulations may favor a particular technology and may prevent application of others. Volume 3, part 2 presents abstracts of the regulatory constraint documents that relate to all phases of LLRW management (e.g., treatment, packaging, storage, transportation, and disposal).

  19. Three Mile Island nuclear reactor accident of March 1979. Environmental radiation data: Volume V. A report to the President's Commission on the Accident at Three Mile Island

    International Nuclear Information System (INIS)

    Bretthauer, E.W.; Grossman, R.F.; Thome, D.J.; Smith, A.E.

    1981-03-01

    This report contains a listing of environmental radiation monitoring data collected in the vicinity of Three Mile Island (TMI) following the March 28, 1979 accident. These data were collected by the EPA, NRC, DOE, HHS, the Commonwealth of Pennsylvania, or the Bethlehem Steel Corporation. This volume consists of the following 2 volumes: Table 16 Summary of Metropolitan Edison Company (Met-Ed) sampling and analytical procedures; and Table 17 Computer printout of data collected by Met-Ed

  20. Low-level radioactive waste from commercial nuclear reactors. Volume 4. Proceedings of the workshop on research and development needs for treatment of low-level radioactive waste from commercial nuclear reactors

    International Nuclear Information System (INIS)

    Godbee, H.W.; Frederick, E.J.; Jolley, R.L.; Kibbey, A.H.; Rodgers, B.R.

    1986-05-01

    The overall task of this program was to provide an assessment of currently available technology for treating commercial low-level radioactive waste (LLRW), to initiate development of a methodology for choosing one technology for a given application, and to identify research needed to improve current treatment techniques and decision methodology. The resulting report is issued in four volumes. As part of this program, a workshop was conducted for determining research and development needs in LLRW treatment. Volume 4, the proceedings of this workshop, includes the formal presentations and both panel and general discussions dealing with such issues as disposal, compaction, and the ''below regulatory concern'' philosophy. Summaries of individual workshops dealing with specific aspects of LLRW treatment are also presented in this volume

  1. ASSESSMENT OF THE TOTAL PETROLEUM HYDROCARBON CONTENT OF AGRICULTURAL SOIL POLLUTED WITH DIFFERENT VOLUME OF CRUDE OIL DURING PLANT- MICROBE INTERACTION

    Directory of Open Access Journals (Sweden)

    Toochukwu Ekwutosi OGBULIE

    2014-06-01

    Full Text Available The effectiveness of plants in interaction with indigenous organisms in environmental clean –up was evaluated. The agricultural soil used for the study was polluted with 100ml, 200ml, 400ml and 800ml of Bonny light crude oil [100%]. Pre and post Microbial examination of the polluted soil identified the indigenous flora present in the soil to be Penicillum sp Aspergillus fumigatus, Aspergillus niger, Candida sp, Pseudomonas fluorescence, Acinetobacter baumanni, Bacillus mycoides, Klebsiella sp., Staphylococcus aureus and Escherichia coli though the absence of S aureus and E. coli was evident during the latter. Vigna unguiculata var unguiculata, Mucuna pruriens, Zea mays and Telfairia occidentalis were the test plant used. Gas chromatographic (GC analysis revealed the total petroleum hydrocarbon (TPH of polluted soil on comparison with the value of 10,380 kg/ mg for control sample, to be low. The high TPH obtained from samples polluted with higher concentration depicts that the numbers of plants to be cultivated for remediation could be a determining factor for a faster clean-up. Statistical analysis using analysis of variance (ANOVA model of SPSS software however, showed there was no significant difference in the degradation of crude oil in samples that are in the green house or field.

  2. Supercritical Water Reactors

    International Nuclear Information System (INIS)

    Bouchter, J.C.; Dufour, P.; Guidez, J.; Latge, C.; Renault, C.; Rimpault, G.

    2014-01-01

    The supercritical water reactor (SCWR) is one of the 6 concepts selected for the 4. generation of nuclear reactors. SCWR is a new concept, it is an attempt to optimize boiling water reactors by using the main advantages of supercritical water: only liquid phase and a high calorific capacity. The SCWR requires very high temperatures (over 375 C degrees) and very high pressures (over 22.1 MPa) to operate which allows a high conversion yield (44% instead of 33% for a PWR). Low volumes of coolant are necessary which makes the neutron spectrum shift towards higher energies and it is then possible to consider fast reactors operating with supercritical water. The main drawbacks of supercritical water is the necessity to use very high pressures which has important constraints on the reactor design, its physical properties (density, calorific capacity) that vary strongly with temperatures and pressures and its very high corrosiveness. The feasibility of the concept is not yet assured in terms of adequate materials that resist to corrosion, reactor stability, reactor safety, and reactor behaviour in accidental situations. (A.C.)

  3. Final generic environmental statement on the use of recycle plutonium in mixed oxide fuel in light water cooled reactors. Volume 3

    International Nuclear Information System (INIS)

    1976-08-01

    An assessment is presented of the health, safety and environmental effects of the entire light water reactor fuel cycle, considering the comparative effects of three major alternatives: no recycle, recycle of uranium only, and recycle of both uranium and plutonium. The assessment covers the period from 1975 through the year 2000 and includes the cumulative effects for the entire period as well as projections for specific years. Topics discussed include: the light water reactor with plutonium recycle; mixed oxide fuel fabrication; reprocessing plant operations; supporting uranium fuel cycle; transportation of radioactive materials; radioactive waste management; storage of plutonium; radiological health assessment; extended spent fuel storage; and blending of plutonium and uranium at reprocessing plants

  4. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 1. Main report. Technical report, September 1977-October 1979

    International Nuclear Information System (INIS)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE

  5. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 1. Main report. Technical report, September 1977-October 1979

    Energy Technology Data Exchange (ETDEWEB)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE.

  6. Final generic environmental statement on the use of recycle plutonium in mixed oxide fuel in light water cooled reactors. Volume 4

    International Nuclear Information System (INIS)

    1976-08-01

    Information concerning the use of plutonium recycle in water cooled reactors is presented under the following chapter headings: probable adverse environmental effects that cannot be avoided; means for mitigating adverse environmental effects; alternative dispositions of plutonium; relationship between local short term uses of man's environment and the maintenance and enhancement of long term productivity; irreversible and irretrievable commitments of resources; and economic analysis and cost-benefit balancing

  7. Total protein

    Science.gov (United States)

    ... page: //medlineplus.gov/ency/article/003483.htm Total protein To use the sharing features on this page, please enable JavaScript. The total protein test measures the total amount of two classes ...

  8. Evaluation of nuclear facility decommissioning projects. Three Mile Island Unit 2 reactor coolant system and systems decontamination. Summary status report. Volume 1

    International Nuclear Information System (INIS)

    Doerge, D.H.; Miller, R.L.; Scotti, K.S.

    1986-05-01

    This document summarizes information relating to the decontamination and restoration of the Three Mile Island Unit 2 reactor coolant system and other plant systems. Data have been collected from activity reports, reactor containment entry records, and other sources and entered in a computerized data system which permits extraction/manipulation of specific data which can be used in planning for recovery from a loss of coolant event similar to that experienced by the Three Mile Island Unit 2 on March 28, 1979. This report contains a summary of radiation exposures, manpower, and time spent in radiation areas during the referenced period. Support activities conducted outside of radiation areas are not included. Computer reports included are: A chronological listing of all activities related to decomtamination and restoration of the reactor coolant system and other plant systems for the period of April 5, 1979, through December 19, 1984; a summary of labor and exposures by department for the same time period; and summary reports for each major task undertaken in connection with this specific work scope during the referenced time period

  9. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Data manual. Part 2: Human error probability (HEP) data; Volume 5, Revision 4

    Energy Technology Data Exchange (ETDEWEB)

    Reece, W.J.; Gilbert, B.G.; Richards, R.E. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1994-09-01

    This data manual contains a hard copy of the information in the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) Version 3.5 database, which is sponsored by the US Nuclear Regulatory Commission. NUCLARR was designed as a tool for risk analysis. Many of the nuclear reactors in the US and several outside the US are represented in the NUCLARR database. NUCLARR includes both human error probability estimates for workers at the plants and hardware failure data for nuclear reactor equipment. Aggregations of these data yield valuable reliability estimates for probabilistic risk assessments and human reliability analyses. The data manual is organized to permit manual searches of the information if the computerized version is not available. Originally, the manual was published in three parts. In this revision the introductory material located in the original Part 1 has been incorporated into the text of Parts 2 and 3. The user can now find introductory material either in the original Part 1, or in Parts 2 and 3 as revised. Part 2 contains the human error probability data, and Part 3, the hardware component reliability data.

  10. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Data manual. Part 2: Human error probability (HEP) data; Volume 5, Revision 4

    International Nuclear Information System (INIS)

    Reece, W.J.; Gilbert, B.G.; Richards, R.E.

    1994-09-01

    This data manual contains a hard copy of the information in the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) Version 3.5 database, which is sponsored by the US Nuclear Regulatory Commission. NUCLARR was designed as a tool for risk analysis. Many of the nuclear reactors in the US and several outside the US are represented in the NUCLARR database. NUCLARR includes both human error probability estimates for workers at the plants and hardware failure data for nuclear reactor equipment. Aggregations of these data yield valuable reliability estimates for probabilistic risk assessments and human reliability analyses. The data manual is organized to permit manual searches of the information if the computerized version is not available. Originally, the manual was published in three parts. In this revision the introductory material located in the original Part 1 has been incorporated into the text of Parts 2 and 3. The user can now find introductory material either in the original Part 1, or in Parts 2 and 3 as revised. Part 2 contains the human error probability data, and Part 3, the hardware component reliability data

  11. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Data manual, Part 2: Human Error Probability (HEP) Data. Volume 5, Revision 4

    International Nuclear Information System (INIS)

    Reece, W.J.; Gilbert, B.G.; Richards, R.E.

    1994-09-01

    This data manual contains a hard copy of the information in the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) Version 3.5 database, which is sponsored by the US Nuclear Regulatory Commission. NUCLARR was designed as a tool for risk analysis. Many of the nuclear reactors in the US and several outside the US are represented in the NUCLARR database. NUCLARR includes both human error probability estimates for workers at the plants and hardware failure data for nuclear reactor equipment. Aggregations of these data yield valuable reliability estimates for probabilistic risk assessments and human reliability analyses. The data manual is organized to permit manual searches of the information if the computerized version is not available. Originally, the manual was published in three parts. In this revision the introductory material located in the original Part 1 has been incorporated into the text of Parts 2 and 3. The user can now find introductory material either in the original Part 1, or in Parts 2 and 3 as revised. Part 2 contains the human error probability data, and Part 3, the hardware component reliability data

  12. Optimization of the biological process using flat membrane bioreactors. Maximum treatment performance with minimum reactor volume; Optimizacion del proceso biologico con BRM de membrana plana. Maximo rendimiento de depuracion con minimo volumen de reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lluch Vallmithana, S.; Lopez Gavin, A.

    2006-07-01

    In a conventional activated sludge process, the membranes are inside the biological reactor where they drain the water through suction or a water column. This system can be operated with heavy loads and sludge of 12-14 g/l or more, and is not affected by problems of bulking or foaming. This makes it suitable for treating difficult industrial waste waters, providing treated water that is free of bacteria and viruses. Micro filtration membranes are flat without any rubbing between them. The membranes require infrequent chemical cleaning and do not need back washing. As no final sedimented is needed, the waste water treatment plant occupies less space. (Author)

  13. Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    International Nuclear Information System (INIS)

    1966-01-01

    This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities

  14. Volume reduction and solidification of liquid and solid low-level radioactive waste

    International Nuclear Information System (INIS)

    May, J.R.

    1979-01-01

    This paper presents a brief background of the development of a method of radioactive waste volume reduction using a unique fluidized bed calciner/incinerator. The volume reduction system is capable of processing a variety of liquid chemical wastes, spent ion exchange resin beads, filter treatment sludges, contaminated lubricating oils, and miscellaneous combustible solids such as paper, rags, protective clothing, wood, etc. All of these wastes are processed in one chemical reaction vessel. Detailed process data is presented that shows the system is capable of reducing the total volume of disposable radioactive waste generated by light water reactors by a factor of 10. Equally important to reducing the volume of power reactor radwaste is the final form of the stored or disposable radwaste. This paper also presents process data related to a new radwaste solidification system, presently being developed, that is particularly suited for immobilizing the granular solids and ashes resulting from volume reduction by calcination and/or incineration

  15. Nuclear reactor with a suspended vessel

    International Nuclear Information System (INIS)

    Lemercier, Guy.

    1977-01-01

    This invention relates to a nuclear reactor with a suspended vessel and applies in particular when this is a fast reactor, the core or active part of the reactor being inside the vessel and immersed under a suitable volume of flowing liquid metal to cool it by extracting the calories released by the nuclear fission in the fuel assemblies forming this core [fr

  16. Twenty-fifth water reactor safety information meeting: Proceedings. Volume 3: Thermal hydraulic research and codes; Digital instrumentation and control; Structural performance

    International Nuclear Information System (INIS)

    Monteleone, S.

    1998-04-01

    This three-volume report contains papers presented at the conference. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following: (1) thermal hydraulic research and codes; (2) digital instrumentation and control; (3) structural performance

  17. Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 2. Alternatives for waste treatment

    International Nuclear Information System (INIS)

    1976-05-01

    Volume II of the five-volume report is devoted to the description of alternatives for waste treatment. The discussion is presented under the following section titles: fuel reprocessing modifications; high-level liquid waste solidification; treatment and immobilization of chop-leach fuel bundle residues; treatment of noncombustible solid wastes; treatment of combustible wastes; treatment of non-high-level liquid wastes; recovery of transuranics from non-high-level wastes; immobilization of miscellaneous non-high-level wastes; volatile radioisotope recovery and off-gas treatment; immobilization of volatile radioisotopes; retired facilities (decontamination and decommissioning); and, modification and use of selected fuel reprocessing wastes

  18. Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 2. Alternatives for waste treatment

    Energy Technology Data Exchange (ETDEWEB)

    1976-05-01

    Volume II of the five-volume report is devoted to the description of alternatives for waste treatment. The discussion is presented under the following section titles: fuel reprocessing modifications; high-level liquid waste solidification; treatment and immobilization of chop-leach fuel bundle residues; treatment of noncombustible solid wastes; treatment of combustible wastes; treatment of non-high-level liquid wastes; recovery of transuranics from non-high-level wastes; immobilization of miscellaneous non-high-level wastes; volatile radioisotope recovery and off-gas treatment; immobilization of volatile radioisotopes; retired facilities (decontamination and decommissioning); and, modification and use of selected fuel reprocessing wastes. (JGB)

  19. A compilation of reports of the Advisory Committee on Reactor Safeguards, 1997 annual, U.S. Nuclear Regulatory Commission. Volume 19

    International Nuclear Information System (INIS)

    1998-04-01

    This compilation contains 67 ACRS reports submitted to the Commission, or to the Executive Director for Operations, during calendar year 1997. It also includes a report to the Congress on the NRC Safety Research Program. Specific topics include: (1) advanced reactor designs, (2) emergency core cooling systems, (3) fire protection, (4) generic letters and issues, (5) human factors, (6) instrumentation, control and protection systems, (7) materials engineering, (8) probabilistic risk assessment, (9) regulatory guides and procedures, rules, regulations, and (10) safety research, philosophy, technology and criteria

  20. A compilation of reports of the Advisory Committee on Reactor Safeguards, 1997 annual, U.S. Nuclear Regulatory Commission. Volume 19

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-04-01

    This compilation contains 67 ACRS reports submitted to the Commission, or to the Executive Director for Operations, during calendar year 1997. It also includes a report to the Congress on the NRC Safety Research Program. Specific topics include: (1) advanced reactor designs, (2) emergency core cooling systems, (3) fire protection, (4) generic letters and issues, (5) human factors, (6) instrumentation, control and protection systems, (7) materials engineering, (8) probabilistic risk assessment, (9) regulatory guides and procedures, rules, regulations, and (10) safety research, philosophy, technology and criteria.

  1. RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume II. Program implementation

    International Nuclear Information System (INIS)

    1976-09-01

    This portion of the RELAP4/MOD5 User's Manual presents the details of setting up and entering the reactor model to be evaluated. The input card format and arrangement is presented in depth, including not only cards for data but also those for editing and restarting. Problem initalization including pressure distribution and energy balance is discussed. A section entitled ''User Guidelines'' is included to provide modeling recommendations, analysis and verification techniques, and computational difficulty resolution. The section is concluded with a discussion of the computer output form and format

  2. Enforcement actions: Significant actions resolved, reactor licensees. Semiannual progress report, July--December 1995. Volume 14, Numbers 3 and 4, Part 2

    International Nuclear Information System (INIS)

    1996-02-01

    This compilation summarizes significant enforcement actions that have been resolved during the period (July--December 1995) and includes copies of letters, Notices, and Orders sent by the Nuclear Regulatory Commission to reactor licensees with respect to these enforcement actions. It is anticipated that the information in this publication will be widely disseminated to managers and employees engaged in activities licensed by the NRC, so that actions can be taken to improve safety by avoiding future violations similar to those described in this publication

  3. An evaluation of state-of-the-art two-velocity two-phase flow models and their applicability to nuclear reactor transient analysis. Volume 2. Theoretical bases. Final report

    International Nuclear Information System (INIS)

    Hughes, E.D.; Lyczkowski, R.W.; McFadden, J.H.

    1976-02-01

    A state-of-the-art review was conducted in order to provide the nuclear industry with a publicly available assessment of two velocity thermal-hydraulic models and their applicability to nuclear reactor technology. The two major objectives of this state-of-the-art evaluation were: (1) document the basic theory in a consistent self-contained report; and (2) apply a prototype 'two-velocity' code (UVUT) to a limited number of separate effect tests. The theoretical basis of the two-velocity models given in Volume 2 is divided into three parts; Part I is the derivation of the basic differential equations; Part II describes in detail, the constitutive models required for closure of the system of equations; and Part III presents the numerical solution schemes

  4. Automation system for measurement of gamma-ray spectra of induced activity for multi-element high-volume neutron activation analysis at the IBR-2 reactor of FLNP at JINR

    International Nuclear Information System (INIS)

    Pavlov, S.S.; Dmitriev, A.Yu.; Chepurchenko, I.A.; Frontas'eva, M.V.

    2014-01-01

    The automation system for measurement of induced activity of gamma-ray spectra for multi-element high-volume neutron activation analysis (NAA) was designed, developed and implemented at the IBR-2 reactor. The system consists of three devices of automatic sample changers for three Canberra HPGe detector-based gamma spectrometry systems. Each sample changer consists of two-axis linear positioning module M202A by DriveSet (DriveSet.de) company and disk with 45 slots for containers with samples. Control of automatic sample changer is performed by the Xemo S360U controller by Systec (systec.de) company. Positioning accuracy can reach 0.1 mm. Special software performs automatic changing of samples and measurement of gamma spectra at constant interaction with the NAA database. The system is unique and can be recommended for other laboratories as one of the possible ways of the NAA integrated automation

  5. Three Mile Island nuclear reactor accident of March 1979. Environmental radiation data: Volume III. A report to the President's Commission on the Accident at Three Mile Island

    International Nuclear Information System (INIS)

    Bretthauer, E.W.; Grossman, R.F.; Thome, D.J.; Smith, A.E.

    1981-03-01

    This report contains a listing of environmental radiation monitoring data collected in the vicinity of Three Mile Island (TMI) following the March 28, 1979 accident. These data were collected by the EPA, NRC, DOE, HHS, the Commonwealth of Pennsylvania, or the Bethlehem Steel Corporation. This volume consists of Table 9 Computer printout of environmental data collected NRC

  6. Radwaste '86: proceedings volume

    International Nuclear Information System (INIS)

    Ainslie, L.C.

    1986-12-01

    The volume contains all the papers presented at the above Conference, which was held in Cape Town, South Africa from 7 to 12 September 1986. A total of 55 contributions cover the full spectrum of the theme of the Conference, which was subdivided into four sessions. Conditioning, treatment and management of radioactive waste: 12 papers reporting on experiences in various countries, as well as specialist topics such as the extraction of radioactive contaminants from reactor pool water. Containment, safe handling and long-term integrity of ILLW packages: 2 papers dealing with cask design. Transport and storage of radwaste and spent fuel: 7 papers ranging from broad overviews to specific operations in different parts of the world. Radioactive waste disposal and environmental impact: 32 papers covering topics from site selection, design and operation, to modelling and monitoring studies. South Africa's Vaalputs radioactive waste disposal facility is comprehensively described. The volume is a useful reference for anyone interested in the disposal of radioactive waste, especially in arid environments, as well as its treatment and management prior to disposal, and will appeal to a wide range of disciplines including engineers, geologists, geophysicists, life scientists and environmentalists. Of particular interest would be the intensive studies undertaken in South Africa prior to the establishment of a radioactive waste repository in that country

  7. Total algorithms

    NARCIS (Netherlands)

    Tel, G.

    We define the notion of total algorithms for networks of processes. A total algorithm enforces that a "decision" is taken by a subset of the processes, and that participation of all processes is required to reach this decision. Total algorithms are an important building block in the design of

  8. 3D Volume and Morphology of Perennial Cave Ice and Related Geomorphological Models at Scăriloara Ice Cave, Romania, from Structure from Motion, Ground Penetrating Radar and Total Station Surveys

    Science.gov (United States)

    Hubbard, J.; Onac, B. P.; Kruse, S.; Forray, F. L.

    2017-12-01

    Research at Scăriloara Ice Cave has proceeded for over 150 years, primarily driven by the presence and paleoclimatic importance of the large perennial ice block and various ice speleothems located within its galleries. Previous observations of the ice block led to rudimentary volume estimates of 70,000 to 120,000 cubic meters (m3), prospectively placing it as one of the world's largest cave ice deposits. The cave morphology and the surface of the ice block are now recreated in a total station survey-validated 3D model, produced using Structure from Motion (SfM) software. With the total station survey and the novel use of ArcGIS tools, the SfM validation process is drastically simplified to produce a scaled, georeferenced, and photo-texturized 3D model of the cave environment with a root-mean-square error (RMSE) of 0.24 m. Furthermore, ground penetrating radar data was collected and spatially oriented with the total station survey to recreate the ice block basal surface and was combined with the SfM model to create a model of the ice block itself. The resulting ice block model has a volume of over 118,000 m3 with an uncertainty of 9.5%, with additional volumes left un-surveyed. The varying elevation of the ice block basal surface model reflect specific features of the cave roof, such as areas of enlargement, shafts, and potential joints, which offer further validation and inform theories on cave and ice genesis. Specifically, a large depression area was identified as a potential area of initial ice growth. Finally, an ice thickness map was produced that will aid in the designing of future ice coring projects. This methodology presents a powerful means to observe and accurately characterize and measure cave and cave ice morphologies with ease and affordability. Results further establish the significance of Scăriloara's ice block to paleoclimate research, provide insights into cave and ice block genesis, and aid future study design.

  9. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2002-01-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

  10. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  11. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

  12. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

  13. Automation of registration of sample weights for high-volume neutron activation analysis at the IBR-2 reactor of FLNP, JINR

    International Nuclear Information System (INIS)

    Dmitriev, A.Yu.; Dmitriev, F.A.

    2015-01-01

    The 'Weight' software tool was created at FLNP JINR to automate the reading of analytical balance readouts and saving these values in the NAA database. The analytical balance connected to the personal computer is used to measure weight values. The 'Weight' software tool controls the reading of weight values and the exchange of information with the NAA database. The weighing process of a large amount of samples is reliably provided during high-volume neutron activation analysis. [ru

  14. Search for other natural fission reactors

    International Nuclear Information System (INIS)

    Apt, K.E.; Balagna, J.P.; Bryant, E.A.; Cowan, G.A.; Daniels, W.R.; Vidale, R.J.

    1977-01-01

    Precambrian uranium ores have been surveyed for evidence of other natural fission reactors. The requirements for formation of a natural reactor direct investigations to uranium deposits with large, high-grade ore zones. Massive zones with volumes approximately greater than 1 m 3 and concentrations approximately greater than 20 percent uranium are likely places for a fossil reactor if they are approximately greater than 0.6 b.a. old and if they contained sufficient water but lacked neutron-absorbing impurities. While uranium deposits of northern Canada and northern Australia have received most attention, ore samples have been obtained from the following worldwide locations: the Shinkolobwe and Katanga regions of Zaire; Southwest Africa; Rio Grande do Norte, Brazil; the Jabiluka, Nabarlek, Koongarra, Ranger, and El Sharana ore bodies of the Northern Territory, Australia; the Beaverlodge, Maurice Bay, Key Lake, Cluff Lake, and Rabbit Lake ore bodies and the Great Bear Lake region, Canada. The ore samples were tested for isotopic variations in uranium, neodymium, samarium, and ruthenium which would indicate natural fission. Isotopic anomalies were not detected. Criticality was not achieved in these deposits because they did not have sufficient 235 U content (a function of age and total uranium content) and/or because they had significant impurities and insufficient moderation. A uranium mill monitoring technique has been considered where the ''yellowcake'' output from appropriate mills would be monitored for isotopic alterations indicative of the exhumation and processing of a natural reactor

  15. Low-level radioactive waste from commercial nuclear reactors. Volume 1. Recommendations for technology developments with potential to significantly improve low-level radioactive waste management

    International Nuclear Information System (INIS)

    Rodgers, B.R.; Jolley, R.L.

    1986-02-01

    The overall task of this program was to provide an assessment of currently available technology for treating commercial low-level radioactive waste (LLRW), to initiate development of a methodology for choosing one technology for a given application, and to identify research needed to improve current treatment techniques and decision methodology. The resulting report is issued in four volumes. Volume 1 provides an executive summary and a general introduction to the four-volume set, in addition to recommendations for research and development (R and D) for low-level radioactive waste (LLRW) treatment. Generic, long-range, and/or high-risk programs identified and prioritized as needed R and D in the LLRW field include: (1) systems analysis to develop decision methodology; (2) alternative processes for dismantling, decontaminating, and decommissioning; (3) ion exchange; (4) incinerator technology; (5) disposal technology; (6) demonstration of advanced technologies; (7) technical assistance; (8) below regulatory concern materials; (9) mechanical treatment techniques; (10) monitoring and analysis procedures; (11) radical process improvements; (12) physical, chemical, thermal, and biological processes; (13) fundamental chemistry; (14) interim storage; (15) modeling; and (16) information transfer. The several areas are discussed in detail

  16. The Traveling Wave Reactor: Design and Development

    Directory of Open Access Journals (Sweden)

    John Gilleland

    2016-03-01

    Full Text Available The traveling wave reactor (TWR is a once-through reactor that uses in situ breeding to greatly reduce the need for enrichment and reprocessing. Breeding converts incoming subcritical reload fuel into new critical fuel, allowing a breed-burn wave to propagate. The concept works on the basis that breed-burn waves and the fuel move relative to one another. Thus either the fuel or the waves may move relative to the stationary observer. The most practical embodiments of the TWR involve moving the fuel while keeping the nuclear reactions in one place−sometimes referred to as the standing wave reactor (SWR. TWRs can operate with uranium reload fuels including totally depleted uranium, natural uranium, and low-enriched fuel (e.g., 5.5% 235U and below, which ordinarily would not be critical in a fast spectrum. Spent light water reactor (LWR fuel may also serve as TWR reload fuel. In each of these cases, very efficient fuel usage and significant reduction of waste volumes are achieved without the need for reprocessing. The ultimate advantages of the TWR are realized when the reload fuel is depleted uranium, where after the startup period, no enrichment facilities are needed to sustain the first reactor and a chain of successor reactors. TerraPower's conceptual and engineering design and associated technology development activities have been underway since late 2006, with over 50 institutions working in a highly coordinated effort to place the first unit in operation by 2026. This paper summarizes the TWR technology: its development program, its progress, and an analysis of its social and economic benefits.

  17. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  18. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  19. Changes in body weight, C-reactive protein, and total adiponectin in non-obese women after 12 months of a small-volume, home-based exercise program.

    Science.gov (United States)

    Mediano, Mauro Felippe Felix; Neves, Fabiana Alves; Cunha, Alessandra Cordeiro de Souza Rodrigues; Souza, Erica Patricia Garcia de; Moura, Anibal Sanchez; Sichieri, Rosely

    2013-01-01

    Our objective was to evaluate the effects of small-volume, home-based exercise combined with slight caloric restriction on the inflammatory markers C-reactive protein and adiponectin. In total, 54 women were randomly assigned to one of two groups for exercise intervention: the control or home-based exercise groups. Weight, waist and hip circumferences, and inflammatory markers were measured at baseline and after 6 and 12 months. Women allocated to the home-based exercise group received a booklet explaining the physical exercises to be practiced at home at least 3 times per week, 40 minutes per session, at low-to-moderate intensity. All participants received dietary counseling aimed at reducing caloric intake by 100-300 calories per day, with a normal distribution of macro-nutrients (26-28% of energy as fat). Clinicaltrials.gov: NCT01206413 RESULTS: The home-based exercise group showed a significantly greater reduction in weight and body mass index at six months, but no difference between groups was observed thereafter. With regard to the inflammatory markers, a greater but non-statistically significant reduction was found for C-reactive protein in the home-based exercise group at six months; however, this difference disappeared after adjusting for weight change. No differences in adiponectin were found at the 6- or 12-month follow-up. Small-volume, home-based exercise did not promote changes in inflammatory markers independent of weight change.

  20. Reactors of the world

    International Nuclear Information System (INIS)

    1971-01-01

    Basic data relating to 127 power reactors in 15 countries which are expected to be in operation at the end of this year, with a total installed electrical generating capacity of 35 340.15 MW(e), and a listing of 361 research reactors in 46 countries are given in the 1971 edition of the IAEA handbook, Power and Research Reactors in Member States, which has just been published. This edition, the fourth, was prepared especially for the Fourth International Conference on the Peaceful Uses of Atomic Energy. (author)

  1. Hydrodynamics of multi-phase packed bed micro-reactors

    NARCIS (Netherlands)

    Márquez Luzardo, N.M.

    2010-01-01

    Why to use packed bed micro-reactors for catalyst testing? Miniaturized packed bed reactors have a large surface-to-volume ratio at the reactor and particle level that favors the heat- and mass-transfer processes at all scales (intra-particle, inter-phase and inter-particle or reactor level). If the

  2. Anaerobic treatment of winery wastewater in fixed bed reactors.

    Science.gov (United States)

    Ganesh, Rangaraj; Rajinikanth, Rajagopal; Thanikal, Joseph V; Ramanujam, Ramamoorty Alwar; Torrijos, Michel

    2010-06-01

    The treatment of winery wastewater in three upflow anaerobic fixed-bed reactors (S9, S30 and S40) with low density floating supports of varying size and specific surface area was investigated. A maximum OLR of 42 g/l day with 80 +/- 0.5% removal efficiency was attained in S9, which had supports with the highest specific surface area. It was found that the efficiency of the reactors increased with decrease in size and increase in specific surface area of the support media. Total biomass accumulation in the reactors was also found to vary as a function of specific surface area and size of the support medium. The Stover-Kincannon kinetic model predicted satisfactorily the performance of the reactors. The maximum removal rate constant (U(max)) was 161.3, 99.0 and 77.5 g/l day and the saturation value constant (K(B)) was 162.0, 99.5 and 78.0 g/l day for S9, S30 and S40, respectively. Due to their higher biomass retention potential, the supports used in this study offer great promise as media in anaerobic fixed bed reactors. Anaerobic fixed-bed reactors with these supports can be applied as high-rate systems for the treatment of large volumes of wastewaters typically containing readily biodegradable organics, such as the winery wastewater.

  3. Nuclear reactors

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1977-01-01

    Reference is made to water cooled reactors and in particular to the cooling system of steam generating heavy water reactors (SGHWR). A two-coolant circuit is described for the latter. Full constructural details are given. (U.K.)

  4. Reactor decommissioning

    International Nuclear Information System (INIS)

    Lawton, H.

    1984-01-01

    A pioneering project on the decommissioning of the Windscale Advanced Gas-cooled Reactor, by the UKAEA, is described. Reactor data; policy; waste management; remote handling equipment; development; and recording and timescales, are all briefly discussed. (U.K.)

  5. BP volume reduction equipment

    International Nuclear Information System (INIS)

    Kitamura, Yoshinori; Muroo, Yoji; Hamanaka, Isao

    2003-01-01

    A new type of burnable poison (BP) volume reduction system is currently being developed. Many BP rods, a subcomponent of spent fuel assemblies are discharged from nuclear power reactors. This new system reduces the overall volume of BP rods. The main system consists of BP rod cutting equipment, equipment for the recovery of BP cut pieces, and special transport equipment for the cut rods. The equipment is all operated by hydraulic press cylinders in water to reduce operator exposure to radioactivity. (author)

  6. RA Reactor

    International Nuclear Information System (INIS)

    1978-02-01

    In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation [sr

  7. Multiregion reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de; Nair, R.P.K.

    1979-08-01

    The study of reflected reactors can be done employing the multigroup diffusion method. The neutron conservation equations, inside the intervals, can be written by fluxes and group constants. A reflected reactor (one and two groups) for a slab geometry is studied, aplying the continuity of flux and current in the interface. At the end, the appropriated solutions for a infinite cylindrical reactor and for a spherical reactor are presented. (Author) [pt

  8. International Benchmark on Pressurised Water Reactor Sub-channel and Bundle Tests. Volume II: Benchmark Results of Phase I: Void Distribution

    International Nuclear Information System (INIS)

    Rubin, Adam; Avramova, Maria; Velazquez-Lozada, Alexander

    2016-03-01

    This report summarised the first phase of the Nuclear Energy Agency (NEA) and the US Nuclear Regulatory Commission Benchmark based on NUPEC PWR Sub-channel and Bundle Tests (PSBT), which was intended to provide data for the verification of void distribution models in participants' codes. This phase was composed of four exercises; Exercise 1: steady-state single sub-channel benchmark, Exercise 2: steady-state rod bundle benchmark, Exercise 3: transient rod bundle benchmark and Exercise 4: a pressure drop benchmark. The experimental data provided to the participants of this benchmark is from a series of void measurement tests using full-size mock-up tests for both Boiling Water Reactors (BWRs) and Pressurised Water Reactors (PWRs). These tests were performed from 1987 to 1995 by the Nuclear Power Engineering Corporation (NUPEC) in Japan and made available by the Japan Nuclear Energy Safety Organisation (JNES) for the purposes of this benchmark, which was organised by Pennsylvania State University. Twenty-one institutions from nine countries participated in this benchmark. Seventeen different computer codes were used in Exercises 1, 2, 3 and 4. Among the computer codes were porous media, sub-channel, systems thermal-hydraulic code and Computational Fluid Dynamics (CFD) codes. It was observed that the codes tended to overpredict the thermal equilibrium quality at lower elevations and under predict it at higher elevations. There was also a tendency to overpredict void fraction at lower elevations and underpredict it at high elevations for the bundle test cases. The overprediction of void fraction at low elevations is likely caused by the x-ray densitometer measurement method used. Under sub-cooled boiling conditions, the voids accumulate at heated surfaces (and are therefore not seen in the centre of the sub-channel, where the measurements are being taken), so the experimentally-determined void fractions will be lower than the actual void fraction. Some of the best

  9. Totally James

    Science.gov (United States)

    Owens, Tom

    2006-01-01

    This article presents an interview with James Howe, author of "The Misfits" and "Totally Joe". In this interview, Howe discusses tolerance, diversity and the parallels between his own life and his literature. Howe's four books in addition to "The Misfits" and "Totally Joe" and his list of recommended books with lesbian, gay, bisexual, transgender,…

  10. Guidance for the application of an assessment methodology for innovative nuclear energy systems. INPRO manual - Overview of the methodology. Vol. 1 of 9 of the final report of phase 1 of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) including a CD-ROM comprising all volumes

    International Nuclear Information System (INIS)

    2008-11-01

    The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was initiated in the year 2000, based on a resolution of the IAEA General Conference (GC(44)/RES/21). The main objectives of INPRO are (1) to help to ensure that nuclear energy is available to contribute in fulfilling energy needs in the 21st century in a sustainable manner, (2) to bring together both technology holders and technology users to consider jointly the international and national actions required to achieve desired innovations in nuclear reactors and fuel cycles; and (3) to create a forum to involve all relevant stakeholders that will have an impact on, draw from, and complement the activities of existing institutions, as well as ongoing initiatives at the national and international level. This document follows the guidelines of the INPRO report 'Methodology for the assessment of innovative nuclear reactors and fuel cycles, Report of Phase 1B (first part) of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)', IAEA-TECDOC-1434 (2004), together with its previous report Guidance for the evaluation for innovative nuclear reactors and fuel cycles, Report of Phase 1A of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO), IAEA-TECDOC-1362 (2003). This INPRO manual is comprised of an overview volume (laid out in this report), and eight additional volumes (available on a CD-ROM attached to the inside back cover of this report) covering the areas of economics (Volume 2), infrastructure (Volume 3), waste management (Volume 4), proliferation resistance (Volume 5), physical protection (Volume 6), environment (Volume 7), safety of reactors (Volume 8), and safety of nuclear fuel cycle facilities (Volume 9). The overview volume sets out the philosophy of INPRO and a general discussion of the INPRO methodology. This overview volume discusses the relationship of INPRO with the UN concept of sustainability to demonstrate how the

  11. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sato, Morihiko.

    1994-01-01

    Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

  12. Innovative Energy Planning and Nuclear Option Using CANDLE Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sekimoto, H; Nagata, A; Mingyu, Y [Tokyo Institute of Technology, Tokyo (Japan)

    2008-07-01

    A new reactor burn-up strategy CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move upward (or downward) along its core axis. This burn-up strategy can derive many merits. The change of excess reactivity along burn-up is theoretically zero for ideal equilibrium condition, and shim rods will not be required for this reactor. The reactor becomes free from accidents induced by unexpected control rods withdrawal. The core characteristics, such as power feedback coefficients and power peaking factor, are not changed during life of operation. Therefore, the operation of the reactor becomes much easier than the conventional reactors. The infinite-medium neutron multiplication factor of replacing fuel is less than unity. Therefore, the transportation and storage of replacing fuels becomes easy and safe, since they are free from criticality accidents. Small long life fast reactor with CANDLE burn-up concept has investigated with depleted uranium as a replacing fuel. Both core diameter and height are chosen to be 2.0 m, and the thermal power is 200 MW. Lead-bismuth is used as a coolant, and nitride (enriched N-15) fuel are employed. The velocity of burning region along burn-up is less than 1.0 cm/year that enables a long life design easily. The core averaged discharged fuel burn-up is about 40 percent. It is about ten times of light water reactor burn-up. The spent fuel volume becomes one-tenth of light water reactor spent fuel. If a light water reactor with a certain power output has been operated for 40 years, the CANDLE reactor can be operated for 2000 years with the same power output and with only depleted uranium left after fuel production for the light water reactor. The system does not need any reprocessing or enrichment. Therefore, the reactor operation becomes very safe, the waste

  13. Innovative Energy Planning and Nuclear Option Using CANDLE Reactors

    International Nuclear Information System (INIS)

    Sekimoto, H.; Nagata, A.; Mingyu, Y.

    2008-01-01

    A new reactor burn-up strategy CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move upward (or downward) along its core axis. This burn-up strategy can derive many merits. The change of excess reactivity along burn-up is theoretically zero for ideal equilibrium condition, and shim rods will not be required for this reactor. The reactor becomes free from accidents induced by unexpected control rods withdrawal. The core characteristics, such as power feedback coefficients and power peaking factor, are not changed during life of operation. Therefore, the operation of the reactor becomes much easier than the conventional reactors. The infinite-medium neutron multiplication factor of replacing fuel is less than unity. Therefore, the transportation and storage of replacing fuels becomes easy and safe, since they are free from criticality accidents. Small long life fast reactor with CANDLE burn-up concept has investigated with depleted uranium as a replacing fuel. Both core diameter and height are chosen to be 2.0 m, and the thermal power is 200 MW. Lead-bismuth is used as a coolant, and nitride (enriched N-15) fuel are employed. The velocity of burning region along burn-up is less than 1.0 cm/year that enables a long life design easily. The core averaged discharged fuel burn-up is about 40 percent. It is about ten times of light water reactor burn-up. The spent fuel volume becomes one-tenth of light water reactor spent fuel. If a light water reactor with a certain power output has been operated for 40 years, the CANDLE reactor can be operated for 2000 years with the same power output and with only depleted uranium left after fuel production for the light water reactor. The system does not need any reprocessing or enrichment. Therefore, the reactor operation becomes very safe, the waste

  14. BWR type reactor

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1988-01-01

    Purpose: To inhibit the lowering of the neutron moderation effect due to voids in the upper portion of the reactor core, thereby flatten the axial power distribution. Constitution: Although it has been proposed to enlarge the diameter at the upper portion of a water rod thereby increasing the moderator in the upper portion, since the water rod situates within the channel box, the increase in the capacity thereof is has certain limit. In the present invention, it is designed such that the volume of the region at the outside of the channel box for the fuel assembly to which non-boiling water in the non-boiling water region can enter is made greater in the upper portion than in the lower portion of the reactor core. Thus, if the moderator density in the upper portion of the reactor core should be decreased due to the generation of the voids, the neutron moderating effect in the upper portion of the reactor core is not lowered as compared with the lower portion of the reactor core and, accordingly, the axial power distribution can be flattening more as compared with that in the conventional nuclear reactors. (Takahashi, M.)

  15. Feasibility of and methodology for thermal annealing an embrittled reactor vessel. Volume 2. Detailed technical description of the work. Final report

    International Nuclear Information System (INIS)

    Mager, T.R.

    1982-11-01

    Program materials were three weldments fabricated from A533 Grade B class 1 plate material and Mn Mo Ni weld wire. Specimens fabricated from the three submerged arc weldments included Type A Charpy V-notch impact, small size tensile, and 1/2T compact tension specimens. After encapsulation, the specimens were irradiated at the UVAR to two fluence levels, 8 x 10 18 n/cm 2 and 1.5 x 10 19 n/cm 2 (E > 1 MeV). Specimens were subjected to sequences of irradiation and anneals and then tested. Metallurgial/mechanistic analyses were also performed. It was concluded that excellent recovery of all properties could be achieved by annealing at greater than or equal to 850 0 F (454 0 C) for 168 hours. Such an annealing resulted in ductile-brittle transition temperature shift recovery of 80 to 100%, and reirradiation after this annealing indicated that the ductile-brittle transition temperature shift appears to continue at the expected rate. Several drawbacks were identified for wet thermal annealing. A conceptual dry in-situ thermal annealing procedure was developed for thermal annealing embrittled reactor vessels

  16. Waste generated by the future decommissioning of the Magurele VVR-S Research Reactor

    International Nuclear Information System (INIS)

    Dragolici, F.; Turcanu, C.N.; Dragolici, A.C.

    2001-01-01

    Nuclear Research Reactor WWR-S from the National Institute of Research and Development for Physics and Nuclear Engineering 'Horia Hulubei', Bucharest-Magurele, was commissioned in July 1957 and it was shut down in December 1997. At the moment the reactor is in conservation state. During its operation this reactor worked at an average power of 2MW, almost 3216 h/year, producing a total thermal power of 230 x 10 3 MWh. No major modifications or improvements were made during the 40 years of operation to the essential parts of the reactor, respective to the primary cooling system, reactor vessel, active core and electronic devices. So, all components of the measure, control and protection systems are old, generally at the technical level of the 1950s, therefore a reason why in December 1997 the operation was ceased. At present, the reactor can be considered, by IAEA definition in the first stage (reactor shut down, but the vital functions are maintained and monitored). The survey is related to the second stage - restrictive use of the area. To develop a real decommissioning project, it was first necessary to evaluate the volume and the characteristics of the radioactive waste which will be generated. Radioactive waste generated during the decommissioning of Magurele WR-S research reactor may be classified as: Activated wastes (internal structures, horizontal channels and thermal column, biological shield); Contaminated wastes (primary circuit non-activated components, hot cells, some technological rooms as main hall, pumps room, radioactive material transfer areas, ventilation building and stack); Possibly contaminated materials from any area of reactor building and ventilation building. After 40 years of nuclear research activities, all such areas are suspected of contamination. The volume of wastes that will result from WWR-S Research Reactor decommissioning is summarized

  17. RA Reactor operation and maintenance (I-IX), Part IV, Task 3.08/04, Refurbishment of the RA reactor; Pogon i odrzavanje reaktora RA (I-IX), IV Deo, Zadatak 3.08/04 Remont reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Zecevic, V [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    This volume contains reports describing maintenance and repair work of the RA reactor instrumentation, equipment of the reactor dosimetry control system, and equipment for regulation and control systems.

  18. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  19. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

    Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

  20. Three Mile Island nuclear reactor accident of March 1979. Environmental radiation data: Volume IV. A report to the President's Commission on the Accident at Three Mile Island

    International Nuclear Information System (INIS)

    Bretthauer, E.W.; Grossman, R.F.; Thome, D.J.; Smith, A.E.

    1981-03-01

    This report contains a listing of environmental radiation monitoring data collected in the vicinity of Three Mile Island (TMI) following the March 28, 1979 accident. These data were collected by the EPA, NRC, DOE, HHS, the Commonwealth of Pennsylvania, or the Bethlehem Steel Corporation. The original report was printed in September 1979 and the update was released in December 1979. This volume consists of the following: Table 10 Summary of US Department of Energy (DOE) sampling and analytical procedures; Table 11 Computer printout of environmental data collected by DOE; Table 12 Summary of Commonwealth of Pennsylvania sampling and analytical procedures; Table 13 Computer printout of environmental data collected by the Commonwealth of Pennsylvania; Table 14 Summary of State of New Jersey sampling and analytical procedures; Table 15 Computer printout of data collected by the State of New Jersey