WorldWideScience

Sample records for reactor snf wps

  1. Technical, economical and legal aspects of repatriation of Russian-origin research reactor SNF to Russia

    International Nuclear Information System (INIS)

    Smirnov, A.; Kanashov, B.; Efarov, S.; Lebedev, A.; Kolupaev, D.

    2005-01-01

    The aim of the report is to find some principal decisions to implement an Agreement between the Governments of the Russian Federation and the USA on repatriation of the research reactor spent nuclear fuel (RR SNF) to the Russian Federation. The report presents some ideas and approaches to the transportation of the Russian-origin RR SNF from the technical, economical and legal viewpoints. The report summarizes the Russian experience and possibilities to fulfill the program under the Agreement. Some decisions are proposed related to application of the international transportation experience and the most advanced technologies for the RR SNF handling. At present, there is no any unified SNF transportation technology that is capable to implement the transportation program schedule set by the Agreement. The decision is in the comprehensive approach as well as in the development of mobile and flexible schemes and in implementation of parallel and combined shipments. (author)

  2. oWPS VERSUS cWPS

    CERN Document Server

    Mainaud Durand, H; Herty, A; Marin, A; Rude, V

    2012-01-01

    The strategy of the CLIC pre-alignment relies on Wire Positioning Sensors (WPS) measuring the radial and vertical offsets with respect to a stretched wire. A precision below 2 µm and an accuracy of 5 µm over a whole range of measurement of 10 mm per axis are required for these sensors. Two types of sensors, based on two different technologies are under development and study at CERN: the capacitive sensor (cWPS) is already in use for the monitoring of the position of the low beta triplets in the LHC and the optical sensor (oWPS) is currently under development with Open Source Instruments. The cWPS had to be upgraded in order to reach the specifications required by the CLIC alignment. The oWPS is a new development especially designed to the CLIC demands. The paper presents the two types of sensors, the developments, as well as the latest results obtained in validation tests. These two types of sensors are part of a common test setup: results of inter-comparison tests achieved on this setup are detailed.

  3. Receipt capability for foreign research reactor (FRR) spent nuclear fuel (SNF) at the Savannah River Site (SRS)

    International Nuclear Information System (INIS)

    Clark, William D. Jr.

    1997-01-01

    The United Stated Department of Energy began implementation of the ten year FRR SNF return policy in May, 1996. Seventeen months into the thirteen year return program, four shipments have been made, returning 863 assemblies of aluminum clad SNF to SRS. Five additional shipments containing over 1,200 assemblies are scheduled in fiscal year 1998. During negotiation of contracts with various reactor operators, it has become apparent that many facilities wish to delay the return of their SNF until the latter part of the program. This has raised concern on the part of the DOE that insufficient receipt capability will exist during the last three to five years of the program to ensure the return of all of the SNF. To help quantify this issue and ensure that it is addressed early in the program, a computer simulation model has been developed at SRS to facilitate the planning, scheduling, and analysis of SNF shipments to be received from offsite facilities. The simulation model, called OFFSHIP, greatly reduces the time and effort required to analyze the complex global transportation system that involves dozens of reactor facilities, multiple casks and fuel types, and time-dependent SNF inventories. OFFSHIP allows the user to input many variables including priorities, cask preferences, shipping date preferences, turnaround times, and regional groupings. User input is easily managed using a spreadsheet format and the output data is generated in a spreadsheet format to facilitate detailed analysis and prepare graphical results. The model was developed in Microsoft Visual Basic for Applications and runs native in Microsoft Excel. The receipt schedules produced by the model have been compared to schedules generated manually with consistent results. For the purposes of this presentation, four scenarios have been developed. The 'Base Case' accounts for those countries/facilities that DOE believes may not participate in the return program. The three additional scenarios look at the

  4. ZOO: la piattaforma WPS libera

    Directory of Open Access Journals (Sweden)

    Luca Delluchi

    2010-03-01

    Full Text Available ZOO: WPS serverZOO is a WPS server with powered calculus abilities on geospatial data directly in the web, deriving from the international collaboration of French and Japaneseresearchers. ZOO is composed by three main parts: the Kernel, i.e. the core of the software, the Services, i.e. program processes allowing to connect the differentZOO libraries, and the API, Javascript based libraries for creating and managing WPS services. While an exhaustive description of ZOO is beyond the scope of this paper, we do hope to stimulate discussion bout possibilities and challenges of webbased analysis of geospatial data.

  5. ZOO: la piattaforma WPS libera

    Directory of Open Access Journals (Sweden)

    Luca Delluchi

    2010-03-01

    ZOO libraries, and the API, Javascript based libraries for creating and managing WPS services. While an exhaustive description of ZOO is beyond the scope of this paper, we do hope to stimulate discussion bout possibilities and challenges of webbased analysis of geospatial data.

  6. An innovative way of thinking nuclear waste management - Neutron physics of a reactor directly operating on SNF.

    Science.gov (United States)

    Merk, Bruno; Litskevich, Dzianis; Bankhead, Mark; Taylor, Richard J

    2017-01-01

    A solution for the nuclear waste problem is the key challenge for an extensive use of nuclear reactors as a major carbon free, sustainable, and applied highly reliable energy source. Partitioning and Transmutation (P&T) promises a solution for improved waste management. Current strategies rely on systems designed in the 60's for the massive production of plutonium. We propose an innovative strategic development plan based on invention and innovation described with the concept of developments in s-curves identifying the current boundary conditions, and the evolvable objectives. This leads to the ultimate, universal vision for energy production characterized by minimal use of resources and production of waste, while being economically affordable and safe, secure and reliable in operation. This vision is transformed into a mission for a disruptive development of the future nuclear energy system operated by burning of existing spent nuclear fuel (SNF) without prior reprocessing. This highly innovative approach fulfils the sustainability goals and creates new options for P&T. A proof on the feasibility from neutronic point of view is given demonstrating sufficient breeding of fissile material from the inserted SNF. The system does neither require new resources nor produce additional waste, thus it provides a highly sustainable option for a future nuclear system fulfilling the requests of P&T as side effect. In addition, this nuclear system provides enhanced resistance against misuse of Pu and a significantly reduced fuel cycle. However, the new system requires a demand driven rethinking of the separation process to be efficient.

  7. An innovative way of thinking nuclear waste management - Neutron physics of a reactor directly operating on SNF.

    Directory of Open Access Journals (Sweden)

    Bruno Merk

    Full Text Available A solution for the nuclear waste problem is the key challenge for an extensive use of nuclear reactors as a major carbon free, sustainable, and applied highly reliable energy source. Partitioning and Transmutation (P&T promises a solution for improved waste management. Current strategies rely on systems designed in the 60's for the massive production of plutonium. We propose an innovative strategic development plan based on invention and innovation described with the concept of developments in s-curves identifying the current boundary conditions, and the evolvable objectives. This leads to the ultimate, universal vision for energy production characterized by minimal use of resources and production of waste, while being economically affordable and safe, secure and reliable in operation. This vision is transformed into a mission for a disruptive development of the future nuclear energy system operated by burning of existing spent nuclear fuel (SNF without prior reprocessing. This highly innovative approach fulfils the sustainability goals and creates new options for P&T. A proof on the feasibility from neutronic point of view is given demonstrating sufficient breeding of fissile material from the inserted SNF. The system does neither require new resources nor produce additional waste, thus it provides a highly sustainable option for a future nuclear system fulfilling the requests of P&T as side effect. In addition, this nuclear system provides enhanced resistance against misuse of Pu and a significantly reduced fuel cycle. However, the new system requires a demand driven rethinking of the separation process to be efficient.

  8. An innovative way of thinking nuclear waste management – Neutron physics of a reactor directly operating on SNF

    Science.gov (United States)

    Litskevich, Dzianis; Bankhead, Mark; Taylor, Richard J.

    2017-01-01

    A solution for the nuclear waste problem is the key challenge for an extensive use of nuclear reactors as a major carbon free, sustainable, and applied highly reliable energy source. Partitioning and Transmutation (P&T) promises a solution for improved waste management. Current strategies rely on systems designed in the 60’s for the massive production of plutonium. We propose an innovative strategic development plan based on invention and innovation described with the concept of developments in s-curves identifying the current boundary conditions, and the evolvable objectives. This leads to the ultimate, universal vision for energy production characterized by minimal use of resources and production of waste, while being economically affordable and safe, secure and reliable in operation. This vision is transformed into a mission for a disruptive development of the future nuclear energy system operated by burning of existing spent nuclear fuel (SNF) without prior reprocessing. This highly innovative approach fulfils the sustainability goals and creates new options for P&T. A proof on the feasibility from neutronic point of view is given demonstrating sufficient breeding of fissile material from the inserted SNF. The system does neither require new resources nor produce additional waste, thus it provides a highly sustainable option for a future nuclear system fulfilling the requests of P&T as side effect. In addition, this nuclear system provides enhanced resistance against misuse of Pu and a significantly reduced fuel cycle. However, the new system requires a demand driven rethinking of the separation process to be efficient. PMID:28749952

  9. SMILE: numerical evaluation of the WPS validation test

    International Nuclear Information System (INIS)

    Moinereau, D.; Studer, V.; Dahl, A.; Wadier, Y.

    2004-01-01

    The reactor pressure vessel (RPV) is an essential component liable to limit the lifetime duration of nuclear PWR power plants. The structural integrity assessment of RPV subjected to pressurized thermal shock (PTA) transients made at an European level does not take always into account the potential beneficial effect of the load history (warm pre-stress WPS). A three-year European Research and Development program (SMILE) started in January 2002 as part of the Fifth Framework Program of the European Atomic Energy Community (EURATOM) to evaluate this effect. The SMILE project is one of a ''cluster'' of Fifth Framework Projects in the area of Plant Life Management. It aims to give sufficient elements to model and to validate the beneficial WPS effect in a RPV structural integrity assessment. Finally, this project aims to harmonize the different approaches to lay the basis for European codes and standards regarding the inclusion of the warm pre-stress (WPS) effect in the RPV assessments. Within the framework of this project, an important experimental work has been conducted including WPS type tests on CT specimens and also a PTS type transient experiment on a large cracked cylinder. The present paper describes shortly the PTS type experiment and presents the corresponding analyses based on engineering methods, finite element elastic and elastic-plastic computations, and local approach to fracture. The results are in good agreement with the experimental result. Significant margins are underlined, with an effective significant increase of the material resistance regarding the risk of brittle failure. (orig.)

  10. DESIGN ANALYSIS FOR THE NAVAL SNF WASTE PACKAGE

    International Nuclear Information System (INIS)

    T.L. Mitchell

    2000-01-01

    The purpose of this analysis is to demonstrate the design of the naval spent nuclear fuel (SNF) waste package (WP) using the Waste Package Department's (WPD) design methodologies and processes described in the ''Waste Package Design Methodology Report'' (CRWMS MandO [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000b). The calculations that support the design of the naval SNF WP will be discussed; however, only a sub-set of such analyses will be presented and shall be limited to those identified in the ''Waste Package Design Sensitivity Report'' (CRWMS MandO 2000c). The objective of this analysis is to describe the naval SNF WP design method and to show that the design of the naval SNF WP complies with the ''Naval Spent Nuclear Fuel Disposal Container System Description Document'' (CRWMS MandO 1999a) and Interface Control Document (ICD) criteria for Site Recommendation. Additional criteria for the design of the naval SNF WP have been outlined in Section 6.2 of the ''Waste Package Design Sensitivity Report'' (CRWMS MandO 2000c). The scope of this analysis is restricted to the design of the naval long WP containing one naval long SNF canister. This WP is representative of the WPs that will contain both naval short SNF and naval long SNF canisters. The following items are included in the scope of this analysis: (1) Providing a general description of the applicable design criteria; (2) Describing the design methodology to be used; (3) Presenting the design of the naval SNF waste package; and (4) Showing compliance with all applicable design criteria. The intended use of this analysis is to support Site Recommendation reports and assist in the development of WPD drawings. Activities described in this analysis were conducted in accordance with the technical product development plan (TPDP) ''Design Analysis for the Naval SNF Waste Package (CRWMS MandO 2000a)

  11. SMILE: test to validate the WPS effect with a cylindrical thick-walled specimen

    International Nuclear Information System (INIS)

    Bezdikian, G.; Moinereau, D.; Roos, E.; Kerkhof, K.; Taylor, N.

    2004-01-01

    The Reactor Pressure Vessel (RPV) is an essential component, which is liable to limit the lifetime duration of PWR plants. The assessment of defects in RPV subjected to pressurized thermal shock (PTS) transients made at an European level generally does not necessarily consider the beneficial effect of the load history (Warm Pre-stress, WPS). The SMILE project - Structural Margin Improvements in aged embrittled RPV with Load history Effects - aims to give sufficient elements to demonstrate, to model and to validate the beneficial WPS effect. It also aims to harmonize the different approaches in the national codes and standards regarding the inclusion of the WPS effect in a RPV structural integrity assessment. The project includes significant experimental work on WPS type experiments with C(T) specimens and a PTS type transient experiment on a large component. This paper deals with the results of the PTS type transient experiment on a component-like, specimen subjected to WPS-loading, the so called Validation Test, carried out within the framework of work package WP4. The test specimen consists of a cylindrical thickwalled specimen with a thickness of 40 mm and an outer diameter of 160 mm, provided with an internal fully circumferential crack with a depth of about 15 mm. The specified load path type is Load-Cool-Unload-Fracture (LCUF). (orig.)

  12. Probability of Criticality for MOX SNF

    International Nuclear Information System (INIS)

    P. Gottlieb

    1999-01-01

    The purpose of this calculation is to provide a conservative (upper bound) estimate of the probability of criticality for mixed oxide (MOX) spent nuclear fuel (SNF) of the Westinghouse pressurized water reactor (PWR) design that has been proposed for use. with the Plutonium Disposition Program (Ref. 1, p. 2). This calculation uses a Monte Carlo technique similar to that used for ordinary commercial SNF (Ref. 2, Sections 2 and 5.2). Several scenarios, covering a range of parameters, are evaluated for criticality. Parameters specifying the loss of fission products and iron oxide from the waste package are particularly important. This calculation is associated with disposal of MOX SNF

  13. Verification of WPS effect under a PTS event

    International Nuclear Information System (INIS)

    Okamura, Hiroyuki; Yagawa, Genki; Urabe, Yoshio; Satoh, Masanobu; Tomimatsu, Minoru; Iida, Masato

    1993-01-01

    In WPS A1 test, K, was monotonically increased with thermal shock and constant tensile load by control of bending load. The specimen broke within the scatter band of K IC data. In WPS A2 test, initial specimen temperature was kept same as that of WPS A1 test and another PTS load was applied. No crack initiation took place when K I was within the scatter band of K IC at the decreasing stage. After that, bending load was increased up to brittle fracture at the temperature where WPS A1 test specimen broke. K I value at fracture for WPS A2 test specimen was beyond the upper bound of K IC data. That is, WPS effect was confirmed even for the low toughness steel like RPV wall under irradiation. Also K I value at fracture can be predicted by Chell's theory. By introducing WPS effect into PTS integrity analysis, much more temperature margin can be expected

  14. Illustration of the WPS benefit through BATMAN test series: Tests on large specimens under WPS loading configurations

    International Nuclear Information System (INIS)

    Yuritzinn, T.; Ferry, L.; Chapuliot, S.; Mongabure, P.; Moinereau, D.; Dahl, A.; Gilles, P.

    2008-01-01

    To study the effects of warm pre-stressing on the toughness of reactor pressure vessel steel, the 'Commissariat a l Energie Atomique', in collaboration with 'Electricite de France' and AREVA-NP, has made a study combining modeling and a series of experiments on large specimens submitted to a thermal shock or isothermal cooling. The tests were made on 18MND5 ferritic steel bars, containing a short or large fatigue pre-crack. The effect of 'warm pre-stressing' was confirmed, in the two cases of a fast thermal shock creating a gradient across the thickness of the bar and for gradual uniform cooling. In both cases, no propagation was observed during the thermal transient. Fracture occurred under low temperature conditions, at the end of the test when the tensile load was increased. The failure loads recorded were substantially higher than during pre-stressing. To illustrate the benefit of the WPS effect, numerical interpretations were performed using either global approach or local approach criteria. WPS effect and capability of models to predict it were then clearly shown. (authors)

  15. WPS criterion proposition based on experimental data base interpretation

    International Nuclear Information System (INIS)

    Chapuliot, S.; Izard, J.P.; Moinereau, D.; Marie, S.

    2011-01-01

    This article gives the background and the methodology developed to define a K J based criterion for brittle fracture of Reactor Pressure Vessel (RPV) submitted to Pressurized Thermal Shock (PTS), and taking into account Warm Pre Stressing effect (WPS). The first step of this methodology is the constitution of an experimental data base. This work was performed through bibliography and partnerships, and allows merging experimental results dealing with: -) Various ferritic steels; -) Various material states (as received, thermally aged, irradiated...); -) Various mode of fracture (cleavage, inter-granular, mixed mode); -) Various specimen geometry and size (CT, SENB, mock-ups); -) Various thermo-mechanical transients. Based on this experimental data base, a simple K J based limit is proposed and compared to experimental results. Parametric studies are performed in order to define the main parameters of the problem. Finally, a simple proposition based on a detailed analysis of tests results is performed. This proposition giving satisfactory results in every cases, it constitutes a good candidate for integration in French RSE-M code for in service assessment. (authors)

  16. Anticipating WPS PIN Vulnerability to Secure Wireless Network

    Directory of Open Access Journals (Sweden)

    Indra Dwi Rianto

    2013-12-01

    Full Text Available WiFi Protected Setup (WPS is a standardized function supported by numerous vendors of wireless routers and access point to help set up connection to a wireless local area network. It is designed to simplify the set up and generally enabled by default. Due to design flaw, the WPS or QSS PIN is susceptible to a brute forceattack. In this paper, we test the security vulnerability occurred, evaluate the performance and give recommendations to anticipate the attack.

  17. International topical meeting on research reactor fuel management (RRFM) - United States foreign research reactor (FRR) spent nuclear fuel (SNF) acceptance program: 2010 update

    International Nuclear Information System (INIS)

    Messick, C.E.; Taylor, J.L.; Niehus, M.T.; Landers, C.

    2010-01-01

    The Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel, adopted by the United States Department of Energy (DOE), in consultation with the Department of State (DOS) in May 1996, scheduled to expire May 12, 2016, to return research reactor fuel until May 12, 2019 to the U.S. is in its fourteenth year. This paper provides a brief update on the program, part of the National Nuclear Security Administration (NNSA), and discusses program initiatives and future activities. The goal of the program continues to be recovery of U.S.-origin nuclear materials, which could otherwise be used in weapons, while assisting other countries to enjoy the benefits of nuclear technology. The NNSA is seeking feedback from research reactor operators to help us understand ways to include eligible research reactors who have not yet participated in the program. (author)

  18. International topical meeting on research reactor fuel management (RRFM) - United States Foreign Research Reactor (FRR) Spent Nuclear Fuel (SNF) acceptance program: 2007 update

    International Nuclear Information System (INIS)

    Messick, C.E.; Taylor, J.L.

    2007-01-01

    The Nuclear Weapons Non-proliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel, adopted by The United States Department of Energy (DOE), in consultation with the Department of State in May 1996, has been extended to expire May 12, 2016, providing an additional 10 years to return fuel to the U.S. This paper provides a brief update on the program, now transferred to the National Nuclear Security Administration (NNSA), and discusses program initiatives and future activities. The goal of the program continues to be recovery of nuclear materials (27 countries have participated so far, returning a total of 7620 spent nuclear fuel elements), which could otherwise be used in weapons, while assisting other countries to enjoy the benefits of nuclear technology. More than ever before, DOE and reactor operators need to work together to schedule shipments as soon as possible, to optimize shipment efficiency over the remaining years of the program. The NNSA is seeking feedback from research reactor operators to help us understand ways to include eligible reactor who have not yet participated in the program

  19. DOE SNF technology development necessary for final disposal

    International Nuclear Information System (INIS)

    Hale, D.L.; Fillmore, D.L.; Windes, W.E.

    1996-01-01

    Existing technology is inadequate to allow safe disposal of the entire inventory of US Department of Energy (DOE) spent nuclear fuel (SNF). Needs for SNF technology development were identified for each individual fuel type in the diverse inventory of SNF generated by past, current, and future DOE materials production, as well as SNF returned from domestic and foreign research reactors. This inventory consists of 259 fuel types with different matrices, cladding materials, meat composition, actinide content, and burnup. Management options for disposal of SNF include direct repository disposal, possible including some physical or chemical preparation, or processing to produce a qualified waste form by using existing aqueous processes or new treatment processes. Technology development needed for direct disposal includes drying, mitigating radionuclide release, canning, stabilization, and characterization technologies. While existing aqueous processing technology is fairly mature, technology development may be needed to apply one of these processes to SNF different than for which the process was originally developed. New processes to treat SNF not suitable for disposal in its current form were identified. These processes have several advantages over existing aqueous processes

  20. Experiencing WPS services in several application domains: opportunities and challenges

    Science.gov (United States)

    lovergine, francesco paolo; tarantino, cristina; d'addabbo, annarita; adamo, patrizia; giuseppe, satalino; refice, alberto; blonda, palma; vicario, saverio

    2016-04-01

    Experiencing WPS services in several application domains: opportunities and challenges ====================================================================================== The implementation of OGC web services and specifically of WPS services revealed itself as a key aspect in order to encourage openess attitude of scientific investigators within several application domains. It can benefit scientific research under different regards, even considering the possibility to promote interoperability, modularity, and the possibility opened by web modeling and the workflow paradigm explotation. Nevertheless it is still a challenging activity and specifically processing services still seem being at an early stage of maturity. This work is about exploitation activities conducted within the GEO GEOSS AIP-8 call by focusing on several applications, such as biodiversity, flood monitoring and soil moisture computation, with implementations based on the pyWPS framework for WPS 1.0 as available at the time of this work. We will present results, lessons learnt and limits found in using those services for distributing demo processing models, along with pro and cons in our experience. References: Refice, A., Capolongo, D., Pasquariello, G., D'Addabbo, A., Bovenga, F., Nutricato, Lovergine F.P., R., Pietranera, L. (2014). SAR and InSAR for Flood Monitoring: Examples With COSMO-SkyMed Data. IEEE Journal of Selected Topics in Applied Earth Observations and Remote Sensing, 7(7), 2711--722. F. Mattia, G. Satalino, A. Balenzano, V. Pauwels, E. De Lathauwer, "GMES Sentinel-1 soil moisture algorithm development", Final report for the European Space Agency, ESA ESTEC Contract No. 4000101352/10 /NL/MP/ef, 30 Nov. 2011. V. Tomaselli, P. Dimopoulos, C. Marangi, A. S. Kallimanis, M. Adamo, C. Tarantino, M. Panitsa, M. Terzi, G. Veronico, F. Lovergine, H. Nagendra, R. Lucas, P. Mairota, C.A. Mucher, P. Blonda, "Translating land cover/land use classifications to habitat taxonomies for landscape

  1. Current state of WWER SNF storage in Russia and the perspectives

    International Nuclear Information System (INIS)

    Anisimov, O.; Kozlov, Y.; Razmashkin, N.; Safutin, V.; Tikhonov, N.

    2006-01-01

    In the Russian Federation WWER-440 Spent Nuclear Fuel (SNF) is reprocessed at RT-1 plant near Cheliabinsk. WWER-1000 SNF is supposed to be reprocessed at RT-2 plant, which will be built about 2020. The information on the capacity and fill up level of the at-reactor pools at NPP with WWER reactors considering its modification up to May 2005 is given. The regulatory requirements to all SNF 'wet' storage facilities; the principle design and engineering solutions as well as the complex of measures for radiation safety and the environmental protection of spent fuel storage are presented. WWER-440 SNF management, WWER-1000 SNF management and dry storage of WWER-1000 SNF are discussed. In the conclusion it is noted than neither Russia, nor any other country have the experience of construction of vault-type 'dry' storage facilities of such a capacity to store WWER-1000 SNF (9000 tU). The experience and design solutions approved earlier in creation of other dangerous facilities were used. The calculations were based on conservative assumptions allowing with a large assurance to guarantee the nuclear and radiation safety and the environmental protection. At present, a program is developed for scientific-technical support of the dry storage facility design and operation, aimed at the studies whose results will allow to optimize the taken technical decisions, simplify SNF management technology and, possibly, to reduce the cost of the storage facility itself

  2. Serbian SNF Repatriation Operation. Issues, Solving, Lesson

    Energy Technology Data Exchange (ETDEWEB)

    Smirnov, A. [Research and Development Company ' Sosny' , Moscow (Russian Federation)

    2011-07-01

    For now the removal of SNF from RA reactor site (PC NFS, Serbia) is the most time-consuming and technically complicated operation under RRRFR Program. The most efficient techniques and lessons learned from other projects of the RRRFR Program as well as new unique technical decisions were used. Two big challenges were resolved during implementation of Serbian Project: (1) preparation of damaged fuel located in the packages unsuitable for transport, taking into account insufficient infrastructure of RA reactor site and (2) removal of large amount of fuel in one multimodal shipment through several transit countries. The main attention was paid to safety justification of all activities. All approvals were obtained in Russia, Serbia and transit countries. Special canisters were designed for transportation of specific RA reactor fuel (of small dimensions, unidentifiable, damaged due to corrosion). The canister design was selected to be untight - it was the most expedient decision for that case from safety perspective. The technology and a set of equipment were designed for remote removal of the fuel from the existing package (aluminum barrels and reactor channels) and placing of the fuel into the new canisters. After fabrication and assembling of the equipment theoretical and practical training of the personnel was performed. Fuel repackaging took about 5 months. SNF was transported in TUK-19 and SKODA VPVR/M casks. The baskets of large capacity were designed and fabricated for SKODA VPVR/M casks. Special requirements to drying the packages and composition of gaseous medium inside were justified to ensure fire and explosion safety. Specialized ISO-containers and transfer equipment designed under Romanian Project were used together with TUK-19 casks. A forklift and mobile rail system were used to handle SKODA VPVR/M casks under conditions of low capacity of the cranes at the facility. Due to the tight schedule of RRRFR Program as well as geographical peculiarities of RA

  3. Evaluation of Neutron Poison Materials for DOE SNF Disposal Systems

    International Nuclear Information System (INIS)

    Vinson, D.W.; Caskey, G.R. Jr.; Sindelar, R.L.

    1998-09-01

    Aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors is being consolidated at the Savannah River Site (SRS) for ultimate disposal in the Mined Geologic Disposal System (MGDS). Most of the aluminum-based fuel material contains highly enriched uranium (HEU) (more than 20 percent 235U), which challenges the preclusion of criticality events for disposal periods exceeding 10,000 years. Recent criticality analyses have shown that the addition of neutron absorbing materials (poisons) is needed in waste packages containing DOE SNF canisters fully loaded with Al-SNF under flooded and degraded configurations to demonstrate compliance with the requirement that Keff less than 0.95. Compatibility of poison matrix materials and the Al-SNF, including their relative degradation rate and solubility, are important to maintain criticality control. An assessment of the viability of poison and matrix materials has been conducted, and an experimental corrosion program has been initiated to provide data on degradation rates of poison and matrix materials and Al-SNF materials under repository relevant vapor and aqueous environments. Initial testing includes Al6061, Type 316L stainless steel, and A516Gr55 in synthesized J-13 water vapor at 50 degrees C, 100 degrees C, and 200 degrees C and in condensate water vapor at 100 degrees C. Preliminary results are presented herein

  4. Commercial SNF Accident Release Fractions

    Energy Technology Data Exchange (ETDEWEB)

    J. Schulz

    2004-11-05

    The purpose of this analysis is to specify and document the total and respirable fractions for radioactive materials that could be potentially released from an accident at the repository involving commercial spent nuclear fuel (SNF) in a dry environment. The total and respirable release fractions are used to support the preclosure licensing basis for the repository. The total release fraction is defined as the fraction of total commercial SNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. Radionuclides are released from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses; this subset of the total release fraction is referred to as the respirable release fraction. Accidents may involve waste forms characterized as: (1) bare unconfined intact fuel assemblies, (2) confined intact fuel assemblies, or (3) canistered failed commercial SNF. Confined intact commercial SNF assemblies at the repository are contained in shipping casks, canisters, or waste packages. Four categories of failed commercial SNF are identified: (1) mechanically and cladding-penetration damaged commercial SNF, (2) consolidated/reconstituted assemblies, (3) fuel rods, pieces, and debris, and (4) nonfuel components. It is assumed that failed commercial SNF is placed into waste packages with a mesh screen at each end (CRWMS M&O 1999). In contrast to bare unconfined fuel assemblies, the

  5. Commercial SNF Accident Release Fractions

    International Nuclear Information System (INIS)

    Schulz, J.

    2004-01-01

    The purpose of this analysis is to specify and document the total and respirable fractions for radioactive materials that could be potentially released from an accident at the repository involving commercial spent nuclear fuel (SNF) in a dry environment. The total and respirable release fractions are used to support the preclosure licensing basis for the repository. The total release fraction is defined as the fraction of total commercial SNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. Radionuclides are released from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses; this subset of the total release fraction is referred to as the respirable release fraction. Accidents may involve waste forms characterized as: (1) bare unconfined intact fuel assemblies, (2) confined intact fuel assemblies, or (3) canistered failed commercial SNF. Confined intact commercial SNF assemblies at the repository are contained in shipping casks, canisters, or waste packages. Four categories of failed commercial SNF are identified: (1) mechanically and cladding-penetration damaged commercial SNF, (2) consolidated/reconstituted assemblies, (3) fuel rods, pieces, and debris, and (4) nonfuel components. It is assumed that failed commercial SNF is placed into waste packages with a mesh screen at each end (CRWMS M andO 1999). In contrast to bare unconfined fuel assemblies, the

  6. Web Service for Positional Quality Assessment: the Wps Tier

    Science.gov (United States)

    Xavier, E. M. A.; Ariza-López, F. J.; Ureña-Cámara, M. A.

    2015-08-01

    In the field of spatial data every day we have more and more information available, but we still have little or very little information about the quality of spatial data. We consider that the automation of the spatial data quality assessment is a true need for the geomatic sector, and that automation is possible by means of web processing services (WPS), and the application of specific assessment procedures. In this paper we propose and develop a WPS tier centered on the automation of the positional quality assessment. An experiment using the NSSDA positional accuracy method is presented. The experiment involves the uploading by the client of two datasets (reference and evaluation data). The processing is to determine homologous pairs of points (by distance) and calculate the value of positional accuracy under the NSSDA standard. The process generates a small report that is sent to the client. From our experiment, we reached some conclusions on the advantages and disadvantages of WPSs when applied to the automation of spatial data accuracy assessments.

  7. SNF Project Engineering Process Improvement Plan

    International Nuclear Information System (INIS)

    DESAI, S.P.

    2000-01-01

    This plan documents the SNF Project activities and plans to support its engineering process. It describes five SNF Project Engineering initiatives: new engineering procedures, qualification cards process; configuration management, engineering self assessments, and integrated schedule for engineering activities

  8. CONTAINMENT EVALUATION OF BREACHED AL-SNF FOR CASK TRANSPORT

    International Nuclear Information System (INIS)

    Vinson, D. W.; Sindelar, R. L.; Iyer, N. C.

    2005-01-01

    Aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors (FRR/DRR) is being shipped to the Savannah River Site. To enter the U.S., the cask with loaded fuel must be certified to comply with the requirements in the Title 10 of the U.S. Code of Federal Regulations, Part 71. The requirements include demonstration of containment of the cask with its contents under normal and accident conditions. Al-SNF is subject to corrosion degradation in water storage, and many of the fuel assemblies are ''failed'' or have through-clad damage. A methodology has been developed with technical bases to show that Al-SNF with cladding breaches can be directly transported in standard casks and maintained within the allowable release rates. The approach to evaluate the limiting allowable leakage rate, L R , for a cask with breached Al-SNF for comparison to its test leakage rate could be extended to other nuclear material systems. The approach for containment analysis of Al-SNF follows calculations for commercial spent fuel as provided in NUREG/CR-6487 that adopts ANSI N14.5 as a methodology for containment analysis. The material-specific features and characteristics of damaged Al-SNF (fuel materials, fabrication techniques, microstructure, radionuclide inventory, and vapor corrosion rates) that were derived from literature sources and/or developed in laboratory testing are applied to generate the four containment source terms that yield four separate cask cavity activity densities; namely, those from fines; gaseous fission product species; volatile fission product species; and fuel assembly crud. The activity values, A 2 , are developed per the guidance of 10CFR71. The analysis is performed parametrically to evaluate maximum number of breached assemblies and exposed fuel area for a proposed shipment in a cask with a test leakage rate

  9. Volumes, Masses, and Surface Areas for Shippingport LWBR Spent Nuclear Fuel in a DOE SNF Canister

    International Nuclear Information System (INIS)

    J.W. Davis

    1999-01-01

    The purpose of this calculation is to estimate volumes, masses, and surface areas associated with (a) an empty Department of Energy (DOE) 18-inch diameter, 15-ft long spent nuclear fuel (SNF) canister, (b) an empty DOE 24-inch diameter, 15-ft long SNF canister, (c) Shippingport Light Water Breeder Reactor (LWBR) SNF, and (d) the internal basket structure for the 18-in. canister that has been designed specifically to accommodate Seed fuel from the Shippingport LWBR. Estimates of volumes, masses, and surface areas are needed as input to structural, thermal, geochemical, nuclear criticality, and radiation shielding calculations to ensure the viability of the proposed disposal configuration

  10. Security preparation for receipt of SNF from the FRR to the INEEL

    International Nuclear Information System (INIS)

    Dahlquist, Rhonda L.

    1997-01-01

    This paper reports the key security-related activities associated with the Foreign Research Reactors (FRR) shipment. Starting with Transportation of the SNF in the country of origin to the final destination at the INEEL. Methodology for compliance will be addressed. The graded approach and a three-step system will be explained. This paper will be used as part of the planning to support the FRR Project for returning the Asia and European SNF back to the United States. (author)

  11. SNF shipping cask shielding analysis

    International Nuclear Information System (INIS)

    Johnson, J.O.; Pace, J.V. III.

    1996-01-01

    The Waste Management and Remedial Action Division has planned a modification sequence for storage facility 7827 in the Solid Waste Storage Area (SWSA). The modification cycle is: (1) modify an empty caisson, (2) transfer the spent nuclear fuel (SNF) of an occupied caisson to a hot cell in building 3525 for inspection and possible repackaging, and (3) return the package to the modified caisson in the SWSA. Although the SNF to be moved is in the solid form, it has different levels of activity. Thus, the following 5 shipping casks will be available for the task: the Loop Transport Carrier, the In- Pile Loop LITR HB-2 Carrier, the 6.5-inch HRLEL Carrier, the HFIR Hot Scrap Carrier, and the 10-inch ORR Experiment Removal Shield Cask. This report describes the shielding tasks for the 5 casks: determination of shielding characteristics, any streaming avenues, estimation of thermal limits, and shielding calculational uncertainty for use in the transportation plan

  12. EQ6 Calculations for Chemical Degradation Of N Reactor (U-Metal) Spent Nuclear Fuel Waste Packages

    International Nuclear Information System (INIS)

    P. Bernot

    2001-01-01

    The Monitored Geologic Repository (MGR) Waste Package Department of the Civilian Radioactive Waste Management System Management and Operating Contractor (CRWMS M and O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the N Reactor, a graphite moderated reactor at the Department of Energy's (DOE) Hanford Site (ref. 1). The N Reactor core was fueled with slightly enriched (0.947 wt% and 0.947 to 1.25 wt% 235 U in Mark IV and Mark IA fuels, respectively) U-metal clad in Zircaloy-2 (Ref. 1, Sec. 3). Both types of N Reactor SNF have been considered for disposal at the proposed Yucca Mountain site. For some WPs, the outer shell and inner shell may breach (Ref. 3) allowing the influx of water. Water in the WP will moderate neutrons, increasing the likelihood of a criticality event within the WP; and the water may, in time, gradually leach the fissile components from the WP, further affecting the neutronics of the system. This study presents calculations of the long-term geochemical behavior of WPs containing two multi-canister overpacks (MCO) with either six baskets of Mark IA or five baskets of Mark IV intact N Reactor SNF rods (Ref. 1, Sec. 4) and two high-level waste (HLW) glass pour canisters (GPCs) arranged according to the codisposal concept (Ref. 4). The specific study objectives were to determine: (1) The extent to which fissile uranium will remain in the WP after corrosion/dissolution of the initial WP configuration (2) The extent to which fissile uranium will be carried out of the degraded WP by infiltrating water (such that internal criticality is no longer possible, but the possibility of external criticality may be enhanced); and (3) The nominal chemical composition for the criticality evaluations of the WP design, and to suggest the range of parametric variations for additional evaluations. The scope of this calculation, the chemical compositions (and subsequent criticality evaluations) of the simulations, is limited to

  13. EQ6 Calculations for Chemical Degradation Of N Reactor (U-Metal) Spent Nuclear Fuel Waste Packages

    Energy Technology Data Exchange (ETDEWEB)

    P. Bernot

    2001-02-27

    The Monitored Geologic Repository (MGR) Waste Package Department of the Civilian Radioactive Waste Management System Management & Operating Contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the N Reactor, a graphite moderated reactor at the Department of Energy's (DOE) Hanford Site (ref. 1). The N Reactor core was fueled with slightly enriched (0.947 wt% and 0.947 to 1.25 wt% {sup 235}U in Mark IV and Mark IA fuels, respectively) U-metal clad in Zircaloy-2 (Ref. 1, Sec. 3). Both types of N Reactor SNF have been considered for disposal at the proposed Yucca Mountain site. For some WPs, the outer shell and inner shell may breach (Ref. 3) allowing the influx of water. Water in the WP will moderate neutrons, increasing the likelihood of a criticality event within the WP; and the water may, in time, gradually leach the fissile components from the WP, further affecting the neutronics of the system. This study presents calculations of the long-term geochemical behavior of WPs containing two multi-canister overpacks (MCO) with either six baskets of Mark IA or five baskets of Mark IV intact N Reactor SNF rods (Ref. 1, Sec. 4) and two high-level waste (HLW) glass pour canisters (GPCs) arranged according to the codisposal concept (Ref. 4). The specific study objectives were to determine: (1) The extent to which fissile uranium will remain in the WP after corrosion/dissolution of the initial WP configuration (2) The extent to which fissile uranium will be carried out of the degraded WP by infiltrating water (such that internal criticality is no longer possible, but the possibility of external criticality may be enhanced); and (3) The nominal chemical composition for the criticality evaluations of the WP design, and to suggest the range of parametric variations for additional evaluations. The scope of this calculation, the chemical compositions (and subsequent criticality evaluations) of the simulations, is limited

  14. An assessment of KW Basin radionuclide activity when opening SNF canisters

    International Nuclear Information System (INIS)

    Bergmann, D.W.; Mollerus, F.J.; Wray, J.L.

    1995-01-01

    N Reactor spent fuel is being stored in sealed canisters in the KW Basin. Some of the canisters contain damaged fuel elements. There is the potential for release of Cs 137, Kr 85, H3, and other fission products and transuranics (TRUs) when canisters are opened. Canister opening is required to select and transfer fuel elements to the 300 Area for examination as part of the Spent Nuclear Fuel (SNF) Characterization program. This report estimates the amount of radionuclides that can be released from Mark II spent nuclear fuel (SNF) canisters in KW Basin when canisters are opened for SNF fuel sampling as part of the SNF Characterization Program. The report also assesses the dose consequences of the releases and steps that can be taken to reduce the impacts of these releases

  15. Standardization of the WPS and Development of the Welding Process Management System for Nuclear Power Plants

    International Nuclear Information System (INIS)

    Cho, Hong Seok; Lee, Jang Wook; Cho, Ki Hyun; Choi, Sang Hoon

    2010-01-01

    The purpose of this study is to integrate existing WPS(welding procedure specification) and PQR(procedure qualification records) which are kept by every branch office respectively and to develop a specialized program that will assist in the creation of WPS and PQR in accordance with the ASME Sec. IX, III, B31.1, etc. These research results make possible not only to ensure structural integrity by applying WPS and PQR correctly but also to cut down on expenses by managing the welding process efficiently. Moreover, as the specialized program will be linked with ERP system between KHNP and KPS, an administration action of welding process will be dealt with on-line. The backgrounds of this study are as follows; · Need to apply WPS correctly and promptly · Need to cut down on expenses due to overlapping development WPS · Need to manage welding resources efficiently · Need to standardize welding QA(quality assurance) documents · Need to improve efficiency of administrative welding process

  16. Experimental investigations on the mechanisms of the warm prestress effect; Experimentelle Untersuchungen zu den Mechanismen des WPS-Effekts

    Energy Technology Data Exchange (ETDEWEB)

    Roos, E.; Alsmann, U. [Stuttgart Univ. (Germany). Staatliche Materialpruefungsanstalt

    1998-11-01

    The Government-supported, joint project by IWW Freiburg, BAM Berlin, IWW Magdeburg, and MPA Stuttgart is intended to yield insight into the mechanisms underlying the WPS, warm prestress effect, and to derive on this basis a quantitative description of the WPS effect. (orig./CB) [Deutsch] Ziel eines durch das BMBF gefoerderten Gemeinschaftsprojekts von IWW Freiburg, BAM Berlin, IWW Magdeburg und MPA Stuttgart ist es, die Mechanismen des WPS-(warm prestress)-Effekts zu erklaeren und auf dieser Basis eine quantitative Beschreibung des WPS-Effekts zu ermoeglichen. (orig./MM)

  17. Status of burnup credit for transport of SNF in the United States

    International Nuclear Information System (INIS)

    Parks, C.V.; Wagner, J.C.

    2004-01-01

    Allowing credit for the reduction in reactivity associated with fuel depletion can enable more cost-effective, higher-density storage, transportation, and disposal of spent nuclear fuel (SNF) while maintaining a subcritical margin sufficient to establish an adequate safety basis. This paper reviews the current status of burnup credit applied to the design and transport of SNF casks in the United States. The existing U.S. regulatory guidance on burnup credit is limited to pressurized-water-reactor (PWR) fuel and to allowing credit only for actinides in the SNF. By comparing loading curves against actual SNF discharge data for U.S. reactors, the potential benefits that can be realized using the current regulatory guidance with actinide-only burnup credit are illustrated in terms of the inventory allowed in high-capacity casks and the concurrent reduction in SNF shipments. The additional benefits that might be realized by extending burnup credit to credit for select fission products are also illustrated. The curves show that, although fission products in SNF provide a small decrease in reactivity compared with actinides, the additional negative reactivity causes the SNF inventory acceptable for transportation to increase from roughly 30% to approximately 90% when fission products are considered. A savings of approximately $150M in transport costs can potentially be realized for the planned inventory of the repository. Given appropriate experimental data to support code validation, a realistic best-estimate analysis of burnup credit that includes validated credit for fission products is the enhancement that will yield the most significant impact on future transportation plans

  18. An implementation of OGC WPS and BPEL4WS compliant dynamic geoprocessing services chain

    Science.gov (United States)

    Xie, Bin; Zhang, Denghui; Yu, Le; Zhang, Dengrong

    2008-12-01

    How to use web services quickly and efficiently is quite important in geospatial applications. A possible solution of sharing and integrating geospatial resources in opening web environment is to chain distributed and diversified geodata and geoprocessing by using web services. This paper presents an approach for chaining geoprocessing by employing Web Processing Service (WPS) and Business Process Execution Language for Web Services (BPEL4WS) under the service-oriented architecture (SOA) and Open Geospatial Consortium (OGC) standard. Workflow control model and SQL Server based register center are used in a prototype system for chaining geoprocessing web services which have been performed functionality decomposition and packed by using extended WPS.

  19. New implementation of OGC Web Processing Service in Python programming language. PyWPS-4 and issues we are facing with processing of large raster data using OGC WPS

    Directory of Open Access Journals (Sweden)

    J. Čepický

    2016-06-01

    Full Text Available The OGC® Web Processing Service (WPS Interface Standard provides rules for standardizing inputs and outputs (requests and responses for geospatial processing services, such as polygon overlay. The standard also defines how a client can request the execution of a process, and how the output from the process is handled. It defines an interface that facilitates publishing of geospatial processes and client discovery of processes and and binding to those processes into workflows. Data required by a WPS can be delivered across a network or they can be available at a server. PyWPS was one of the first implementations of OGC WPS on the server side. It is written in the Python programming language and it tries to connect to all existing tools for geospatial data analysis, available on the Python platform. During the last two years, the PyWPS development team has written a new version (called PyWPS-4 completely from scratch. The analysis of large raster datasets poses several technical issues in implementing the WPS standard. The data format has to be defined and validated on the server side and binary data have to be encoded using some numeric representation. Pulling raster data from remote servers introduces security risks, in addition, running several processes in parallel has to be possible, so that system resources are used efficiently while preserving security. Here we discuss these topics and illustrate some of the solutions adopted within the PyWPS implementation.

  20. SWI/SNF complex in disorder

    Science.gov (United States)

    Santen, Gijs W.E.; Kriek, Marjolein; van Attikum, Haico

    2012-01-01

    Heterozygous germline mutations in components of switch/sucrose nonfermenting (SWI/SNF) chromatin remodeling complexes were recently identified in patients with non-syndromic intellectual disability, Coffin-Siris syndrome and Nicolaides-Baraitser syndrome. The common denominator of the phenotype of these patients is severe intellectual disability and speech delay. Somatic and germline mutations in SWI/SNF components were previously implicated in tumor development. This raises the question whether patients with intellectual disability caused by SWI/SNF mutations in the germline are exposed to an increased risk of developing cancer. Here we compare the mutational spectrum of SWI/SNF components in intellectual disability syndromes and cancer, and discuss the implications of the results of this comparison for the patients. PMID:23010866

  1. Technology development for DOE SNF management

    International Nuclear Information System (INIS)

    Hale, D.L.; Einziger, R.E.; Murphy, J.R.

    1995-01-01

    This paper describes the process used to identify technology development needs for the same management of spent nuclear fuel (SNF) in the US Department of Energy (DOE) inventory. Needs were assessed for each of the over 250 fuel types stores at DOE sites around the country for each stage of SNF management--existing storage, transportation, interim storage, and disposal. The needs were then placed into functional groupings to facilitate integration and collaboration among the sites

  2. Database specification for the Worldwide Port System (WPS) Regional Integrated Cargo Database (ICDB)

    Energy Technology Data Exchange (ETDEWEB)

    Faby, E.Z.; Fluker, J.; Hancock, B.R.; Grubb, J.W.; Russell, D.L. [Univ. of Tennessee, Knoxville, TN (United States); Loftis, J.P.; Shipe, P.C.; Truett, L.F. [Oak Ridge National Lab., TN (United States)

    1994-03-01

    This Database Specification for the Worldwide Port System (WPS) Regional Integrated Cargo Database (ICDB) describes the database organization and storage allocation, provides the detailed data model of the logical and physical designs, and provides information for the construction of parts of the database such as tables, data elements, and associated dictionaries and diagrams.

  3. Canister storage building compliance assessment SNF project NRC equivalency criteria - HNF-SD-SNF-DB-003

    International Nuclear Information System (INIS)

    BLACK, D.M.

    1999-01-01

    This document presents the Project's position on compliance with the SNF Project NRC Equivalency Criteria - HNF-SD-SNF-DE-003, Spent Nuclear Fuel Project Path Forward Additional NRC Requirements. No non-compliances are shown. The compliance statements have been reviewed and approved by DOE. Open items are scheduled to be closed prior to project completion

  4. Assessment results of the Indonesian TRIGA SNF to be shipped to INEEL

    International Nuclear Information System (INIS)

    Jefimoff, J.; Robb, A.K.; Wendt, K.M.; Syarip, I.; Alfa, T.

    1997-01-01

    This paper describes the Training, Research, Isotope, General Atomics (TRIGA) spent nuclear fuel (SNF) examination performed by technical personnel from the Idaho National Engineering and Environmental Laboratory (INEEL) at the Bandung and Yogyakarta research reactor facilities in Indonesia. The examination was required before the SNF would be accepted for transportation to and storage at the INEEL. This paper delineates the Initial Preparations prior to the Indonesian foreign research reactor (FRR) fuel examination. The technical basis for the examination, the TRIGA SNF Acceptance Criteria, and the physical condition required for transportation, receipt and storage of the TRIGA SNF at the INEEL is explained. In addition to the initial preparations, preparation descriptions of the Work Plan For TRIGA Fuel Examination, the Underwater Examination Equipment used, and personnel Examination Team Training are included. Finally, the Fuel Examination and Results of the aluminum and stainless steel clad TRIGA fuel examination have been summarized. Lessons learned from all the activities completed to date is provided in an addendum. The initial preparations included: (1) coordination between the INEEL, FRR or Badan Tenaga Atom Nasional (BATAN), DOE-HQ, and the US State Department and Embassy; (2) incorporating Savannah River Site (SRS) FRR experience and lessons learned; (3) collecting both FRR facility and spent fuel data, and issuing a radionuclide report (Radionuclide Mass Inventory, Activity, Decay Heat, and Dose Rate Parametric Data for TRIGA Spent Nuclear Fuels) needed for transportation and fuel acceptance at the INEEL; and (4) preexamination work at the research reactor for the fuel examination

  5. SNF AGING SYSTEM DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    L.L. Swanson

    2005-01-01

    The purpose of this system description document (SDD) is to establish requirements that drive the design of the spent nuclear fuel (SNF) aging system and associated bases, which will allow the design effort to proceed. This SDD will be revised at strategic points as the design matures. This SDD identifies the requirements and describes the system design, as it currently exists, with emphasis on attributes of the design provided to meet the requirements. This SDD is an engineering tool for design control; accordingly, the primary audience and users are design engineers. This SDD is part of an iterative design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. The SDD follows the design with regard to the description of the system. The description provided in the SDD reflects the current results of the design process. Throughout this SDD, the term aging cask applies to vertical site-specific casks and to horizontal aging modules. The term overpack is a vertical site-specific cask that contains a dual-purpose canister (DPC) or a disposable canister. Functional and operational requirements applicable to this system were obtained from ''Project Functional and Operational Requirements'' (F andOR) (Curry 2004 [DIRS 170557]). Other requirements that support the design process were taken from documents such as ''Project Design Criteria Document'' (PDC) (BSC 2004 [DES 171599]), ''Site Fire Hazards Analyses'' (BSC 2005 [DIRS 172174]), and ''Nuclear Safety Design Bases for License Application'' (BSC 2005 [DIRS 171512]). The documents address requirements in the ''Project Requirements Document'' (PRD) (Canori and Leitner 2003 [DIRS 166275]). This SDD includes several appendices. Appendix A is a Glossary; Appendix B is a list of key system charts, diagrams, drawings, lists and additional supporting information; and Appendix C is a list of

  6. SNF AGING SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    L.L. Swanson

    2005-04-06

    The purpose of this system description document (SDD) is to establish requirements that drive the design of the spent nuclear fuel (SNF) aging system and associated bases, which will allow the design effort to proceed. This SDD will be revised at strategic points as the design matures. This SDD identifies the requirements and describes the system design, as it currently exists, with emphasis on attributes of the design provided to meet the requirements. This SDD is an engineering tool for design control; accordingly, the primary audience and users are design engineers. This SDD is part of an iterative design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. The SDD follows the design with regard to the description of the system. The description provided in the SDD reflects the current results of the design process. Throughout this SDD, the term aging cask applies to vertical site-specific casks and to horizontal aging modules. The term overpack is a vertical site-specific cask that contains a dual-purpose canister (DPC) or a disposable canister. Functional and operational requirements applicable to this system were obtained from ''Project Functional and Operational Requirements'' (F&OR) (Curry 2004 [DIRS 170557]). Other requirements that support the design process were taken from documents such as ''Project Design Criteria Document'' (PDC) (BSC 2004 [DES 171599]), ''Site Fire Hazards Analyses'' (BSC 2005 [DIRS 172174]), and ''Nuclear Safety Design Bases for License Application'' (BSC 2005 [DIRS 171512]). The documents address requirements in the ''Project Requirements Document'' (PRD) (Canori and Leitner 2003 [DIRS 166275]). This SDD includes several appendices. Appendix A is a Glossary; Appendix B is a list of key system charts

  7. Remote Qualification of HLS and WPS Systems in the LHC Tunnel

    CERN Document Server

    Mainaud Durand, Helene; Marin, Antonio; Rousseau, Michel; Sosin, Mateusz

    2014-01-01

    The position of the inner triplets of the LHC is monitored using Hydrostatic Levelling System (HLS) and Wire Positioning System (WPS). A regulation of these systems is needed to guarantee the sensors’ function. Such a regulation was done in-situ up to now, but the level of residual radiation at the level of the inner triplets will significantly increase with the next steps of LHC operation. Two systems have been designed to perform such a remote qualification: a filling/purging system for the HLS system and a wire displacer system for the WPS. In the paper, the requirements and the solutions proposed are described, with the emphasis on the conceptual design and the results obtained.

  8. Implementación de WPS en el firmware NodeMCU para el ESP8266

    OpenAIRE

    Candelario Elías, Julio

    2016-01-01

    Este trabajo se centra en el system on chip ESP8266, más en concreto en añadir el estándar WPS al firmware de NodeMCU. Para llevar a cabo este trabajo se han estudiado las características de este SOC y su funcionamiento, así como el estándar WPS y sus distintos modos de funcionamiento. Se ha creado un módulo nuevo en el NodeMCU que permite el uso de las funciones de la librería “libwps.a” proporcionada por la compañía Espressif en su sdk, las cuáles permitirán al NodeMCU hacer ...

  9. Spent Nuclear Fuel (SNF) Project Execution Plan

    International Nuclear Information System (INIS)

    LEROY, P.G.

    2000-01-01

    The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities

  10. Spent Nuclear Fuel (SNF) Project Execution Plan

    Energy Technology Data Exchange (ETDEWEB)

    LEROY, P.G.

    2000-11-03

    The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities.

  11. SNF project engineering process improvement plan

    International Nuclear Information System (INIS)

    DESAI, S.P.

    1999-01-01

    This Engineering Process Improvement Plan documents the activities and plans to be taken by the SNF Project to support its engineering process and to produce a consolidated set of engineering procedures that are fully compliant with the requirements of HNF-PRO-1819. All new procedures will be issued and implemented by September 30, 1999

  12. Functional identification of an Arabidopsis snf4 ortholog by screening for heterologous multicopy suppressors of snf4 deficiency in yeast

    DEFF Research Database (Denmark)

    Kleinow, T.; Bhalerao, R.; Breuer, F.

    2000-01-01

    Yeast Snf4 is a prototype of activating gamma-subunits of conserved Snf1/AMPK-related protein kinases (SnRKs) controlling glucose and stress signaling in eukaryotes. The catalytic subunits of Arabidopsis SnRKs, AKIN10 and AKIN11, interact with Snf4 and suppress the snf1 and snf4 mutations in yeast....... By expression of an Arabidopsis cDNA library in yeast, heterologous multicopy snf4 suppressors were isolated. In addition to AKIN10 and AKIN11, the deficiency of yeast snf4 mutant to grown on non-fermentable carbon source was suppressed by Arabidopsis Myb30, CAAT-binding factor Hap3b, casein kinase I, zinc......-finger factors AZF2 and ZAT10, as well as orthologs of hexose/UDP-hexose transporters, calmodulin, SMC1-cohesin and Snf4. Here we describe the characterization of AtSNF4, a functional Arabidopsis Snf4 ortholog, that interacts with yeast Snf1 and specifically binds to the C-terminal regulatory domain...

  13. Demonstrating the Feasibility of Molten Aluminum for Destroying Polymeric Encapsulants in SNF-Bearing Metallographic Mounts. Final Technical Report

    International Nuclear Information System (INIS)

    Dan Stout; Scott Ploger

    2004-01-01

    DOE-owned spent nuclear fuel (SNF) rods have been cross sectioned and mounted for metallography throughout the history of nuclear reactors. Many hundreds of these ''met mounts'' have accumulated in storage across the DOE complex. However, because of potential hydrogen generation from radiolysis of the polymeric encapsulants, the met mounts are problematic for eventual disposal in a geologic repository

  14. Design and implementation of distributed spatial computing node based on WPS

    International Nuclear Information System (INIS)

    Liu, Liping; Li, Guoqing; Xie, Jibo

    2014-01-01

    Currently, the research work of SIG (Spatial Information Grid) technology mostly emphasizes on the spatial data sharing in grid environment, while the importance of spatial computing resources is ignored. In order to implement the sharing and cooperation of spatial computing resources in grid environment, this paper does a systematical research of the key technologies to construct Spatial Computing Node based on the WPS (Web Processing Service) specification by OGC (Open Geospatial Consortium). And a framework of Spatial Computing Node is designed according to the features of spatial computing resources. Finally, a prototype of Spatial Computing Node is implemented and the relevant verification work under the environment is completed

  15. Interrelation of technologies for RW preparation and sites for final isolation of the wastes from pyrochemical processing of SNF

    Energy Technology Data Exchange (ETDEWEB)

    Gupalo, V.S.; Chistyakov, V.N. [JSC - Design-Prospecting and Scientific-Research Institute of Industrial Technology -, Kashirskoye Highway, 33, Moscow 115409 (Russian Federation); Kormilitsyn, M.V.; Kormilitsyna, L.A. [JSC - State Scientific Center - Research Institute of Atomic Reactors -, Ulyanovsk region, Dimitrovgrad - 10, 433510 (Russian Federation)

    2013-07-01

    For the justification of engineering solutions and practical testing of the radiochemical component of the perspective nuclear power complex with on-site variant of nuclear fuel cycle (NFC), it is planned to establish a multi-functional research-development complex (MFCRC) for radiochemical processing of spent nuclear fuels (SNF) from fast reactors. MFCRC is being established at the NIIAR site, it comprises technological process lines, where innovation pyro-electrochemical and hydrometallurgical technologies are realized, with an option for closing the inter-chain material flows for testing the combined radiochemically converted materials. The technological flowchart for processing at the MFCRC is subdivided into 3 segments: -) complex of the lead operations for dismantling the fuel elements (FE) and fuel assemblies (FA), -) pyrochemical extraction flowchart for processing SNF, and -) hydrometallurgical flowchart for processing SNF. The engineered solutions for the management and disposition of the radioactive wastes from MFCRC are reviewed.

  16. Will the world SNF be reprocessed in Russia?

    International Nuclear Information System (INIS)

    Gagarinski, A.

    2000-01-01

    Russia's possibilities in nuclear fuel reprocessing are well known. RT-1 plant with 400 tons/year in the Chelyabinsk region can provide reprocessing of fuel from Russian and Central European WWER-440 reactors, as well as from transport and research reactors. Former military complex Krasnoyarsk-26 with unique underground installations situated in rock galleries, already has an aqueous facility for storage of 6000 tons of spent nuclear fuel (SNF), half-built plant RT-2 for nuclear fuel reprocessing with 1500 tons/year capacity, as well as the projects of dry storage facility for 30000 tons of SNF and of MOX fuel production plant. Russian nuclear specialists understand well, that the economic efficiency of nuclear fuel reprocessing industry is shown only in case of large-scale production, which would require consolidation of the countries, which develop nuclear energy. They also understand, that Russia has all the possibilities to become one of the centers of such a consolidation and to use these possibilities for the benefit of the country. The idea of foreign nuclear fuel reprocessing (for a long time realized for East and Central European countries, which operate Soviet-design reactors) has existed in the specialists' minds, and sometimes has appeared in the mass media. On the other hand, rehabilitation of territories of nuclear fuel cycle enterprises in Russia continues, including the Karachai lake, which contains 120 million Curie of radioactivity. Unfortunately, Russia simply has no money for complete solution of the problems of radiation military legacy. During discussion of the budget for 2000, the Russian Minatom has made a daring step. A real program, how to find money needed for solving the 'radiation legacy' problem, was proposed. With this purpose, it was proposed to permit storage and further reprocessing of other countries' SNF on Russian territory. It is well known, that another countries' SNF is accepted for reprocessing by UK and France, and Russia

  17. The Snf1 Protein Kinase in the Yeast Saccharomyces cerevisiae

    DEFF Research Database (Denmark)

    Usaite, Renata

    2008-01-01

    4 on the regulation of glucose and galactose metabolism, I physiologically characterized Δsnf1, Δsnf4, and Δsnfsnf4 CEN.PK background yeast strains in glucose and glucose-galactose mixture batch cultivations (chapter 2). The results of this study showed that delayed induction of galactose...... that the stable isotope labeling approach is highly reproducible among biological replicates when complex protein mixtures containing small expression changes were analyzed. Where poor correlation between stable isotope labeling and spectral counting was found, the major reason behind the discrepancy was the lack...

  18. Spent Nuclear Fuel (SNF) Project Product Specification

    International Nuclear Information System (INIS)

    PAJUNEN, A.L.

    2000-01-01

    The process for removal of Spent Nuclear Fuel (SNF) from the K Basins has been divided into major sub-systems. The Fuel Retrieval System (FRS) removes fuel from the existing storage canisters, cleans it, and places it into baskets. The multi-canister overpack (MCO) loading system places the baskets into an MCO that has been pre-loaded in a cask. The cask, containing a loaded MCO, is then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the cask, and MCO, are transferred to the Canister Storage Building (CSB), where the MCO is removed from the cask, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The purpose of this document is to specify the process related characteristics of an MCO at the interface between major process systems. The characteristics are derived from the primary technical documents that form the basis for safety analysis and design calculations. This document translates the calculation assumptions into implementation requirements and describes the method of verifying that the requirement is achieved. These requirements are used to define validation test requirements and describe requirements that influence multiple sub-project safety analysis reports. This product specification establishes limits and controls for each significant process parameter at interfaces between major sub-systems that potentially affect the overall safety and/or quality of the SNF packaged for processing, transport, and interim dry storage. The product specifications in this document cover the SNF packaged in MCOs to be transported throughout the SNF Project. The description of the product specifications are organized in the document as follows: Section 2.0--Summary listing of product specifications at each major sub-system interface. Section 3.0--Summary description providing guidance as to how specifications are complied with by equipment design or processing within a major

  19. Spent Nuclear Fuel (SNF) Project Product Specification

    Energy Technology Data Exchange (ETDEWEB)

    PAJUNEN, A.L.

    2000-12-07

    The process for removal of Spent Nuclear Fuel (SNF) from the K Basins has been divided into major sub-systems. The Fuel Retrieval System (FRS) removes fuel from the existing storage canisters, cleans it, and places it into baskets. The multi-canister overpack (MCO) loading system places the baskets into an MCO that has been pre-loaded in a cask. The cask, containing a loaded MCO, is then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the cask, and MCO, are transferred to the Canister Storage Building (CSB), where the MCO is removed from the cask, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The purpose of this document is to specify the process related characteristics of an MCO at the interface between major process systems. The characteristics are derived from the primary technical documents that form the basis for safety analysis and design calculations. This document translates the calculation assumptions into implementation requirements and describes the method of verifying that the requirement is achieved. These requirements are used to define validation test requirements and describe requirements that influence multiple sub-project safety analysis reports. This product specification establishes limits and controls for each significant process parameter at interfaces between major sub-systems that potentially affect the overall safety and/or quality of the SNF packaged for processing, transport, and interim dry storage. The product specifications in this document cover the SNF packaged in MCOs to be transported throughout the SNF Project. The description of the product specifications are organized in the document as follows: Section 2.0--Summary listing of product specifications at each major sub-system interface. Section 3.0--Summary description providing guidance as to how specifications are complied with by equipment design or processing within a major

  20. Current status of development in dry pyro-electrochemical technology of SNF reprocessing

    International Nuclear Information System (INIS)

    Bychkov, A.V.; Skiba, O.V.; Kormilitsyn, M.V.

    2004-01-01

    The technology of SNF management in molten salts currently developed by a group of institutes headed by RIAR has had several stages of development: - basic research of uranium, plutonium and main FP properties (investigation and reprocessing of different kinds of SNF in 1960 - 1970); - development of the equipment and implementation of the pyro-electrochemical technology of granulated UPu fuel production. Development of the vibro-packing method and in-pile testing of vibro-packed fuel pins with granulated fuel as the most 'logical' continuation of reprocessing: implementation of the technology for BOR-60 and BN-600 (1980 - 1990); - development of closed fuel cycle elements. Checking of the technology using batches of SNF. In-pile tests. Feasibility study of the closed fuel cycle (CFC). Study of application of the technology to other objects (transmutation; nitride, cermet and other fuels) (1980 - 1990). The current status of the research is the following: - Basic research. Properties of uranium, plutonium, thorium, and neptunium in chloride melts have been studied in much detail. The data on physical chemistry and electrochemistry of the main FP is enough for understanding the processes. Detailed studies of americium, curium, and technetium chemistry are the essential investigation directions; - Engineering development. The technology and equipment bases have been developed for the processes of oxide fuel reprocessing and fabrication. The technology was checked using 5500 kg of pure fuel from different reactors and 20 kg of irradiated BN-350 and BOR-60 fuel. The bases of the technology have been provided and the feasibility study has been carried out for a full-scale plant of BN-800 CFC; - Industrial application: Since the technology is highly prepared, the activities on industrial application of U-Pu fuel are now underway. The BOR-60 reactor uses fuel obtained by the dry method, the design of the facility for implementation of CFC reactors is being developed. 9

  1. Interaction of DOE SNF and Packaging Materials

    International Nuclear Information System (INIS)

    Anderson, P.A.

    1998-01-01

    A sensitivity analysis was conducted to identify and evaluate potential destructive interactions between the materials in US Department of Energy (USDOE) spent nuclear fuels (SNFs) and their storage/disposal canisters. The technical assessment was based on the thermodynamic properties as well as the chemical and physical characteristics of the materials expected inside the canisters. No chemical reactions were disclosed that could feasibly corrode stainless steel canisters to the point of failure. However, the possibility of embrittlement (loss of ductility) of the stainless steel through contact with liquid metal fission products or hydrogen inside the canisters cannot be dismissed. Higher-than-currently-permitted internal gas pressures must also be considered. These results, based on the assessment of two representative 90-year-cooled fuels that are stored at 200C in stainless steel canisters with internal blankets of helium, may be applied to most of the fuels in the USDOE's SNF inventory

  2. SNF project engineering process improvement plan

    International Nuclear Information System (INIS)

    KELMENSON, R.L.

    1999-01-01

    This Engineering Process Improvement Plan documents the activities and plans to be taken by the SNF Project (the Project) to support its engineering process and to produce a consolidated set of engineering procedures that are fully compliant with the requirements of HNF-PRO-1819 (1819). These requirements are imposed on all engineering activities performed for the Project and apply to all life-cycle stages of the Project's systems, structures and components (SSCs). This Plan describes the steps that will be taken by the Project during the transition period to ensure that new procedures are effectively integrated into the Project's work process as these procedures are issued. The consolidated procedures will be issued and implemented by September 30, 1999

  3. Storage and treatment of SNF of Alfa class nuclear submarines: current status and problems

    International Nuclear Information System (INIS)

    Ignatiev, Sviatoslav; Zabudko, Alexey; Pankratov, Dmitry; Somov, Ivan; Suvorov, Gennady

    2007-01-01

    Available in abstract form only. Full text of publication follows: The current status and main problems associated with storage, defueling and following treatment of spent nuclear fuel (SNF) of Nuclear Submarines (NS) with heavy liquid metal cooled reactors are considered. In the final analysis these solutions could be realized in the form of separate projects to be funded through national and bi- and multilateral funding in the framework of the international collaboration of the Russian Federation on complex utilization of NS and rehabilitation of contaminated objects allocated in the North-West region of Russia. (authors)

  4. New implementation of OGC Web Processing Service in Python programming language. PyWPS-4 and issues we are facing with processing of large raster data using OGC WPS

    OpenAIRE

    J. Čepický; L. M. de Sousa

    2016-01-01

    The OGC® Web Processing Service (WPS) Interface Standard provides rules for standardizing inputs and outputs (requests and responses) for geospatial processing services, such as polygon overlay. The standard also defines how a client can request the execution of a process, and how the output from the process is handled. It defines an interface that facilitates publishing of geospatial processes and client discovery of processes and and binding to those processes into workflows. Data ...

  5. Aspen Forest Cover by Stratum/Plot (SNF)

    Data.gov (United States)

    National Aeronautics and Space Administration — Average percent coverage and standard deviation of each canopy stratum from subplots at each aspen site during the SNF study in the Superior National Forest, Minnesota

  6. Purification and characterization of the three Snf1-activating kinases of Saccharomyces cerevisiae

    OpenAIRE

    Elbing, Karin; McCartney, Rhonda R.; Schmidt, Martin C.

    2006-01-01

    Members of the Snf1/AMPK family of protein kinases are activated by distinct upstream kinases that phosphorylate a conserved threonine residue in the Snf1/AMPK activation loop. Recently, the identities of the Snf1- and AMPK-activating kinases have been determined. Here we describe the purification and characterization of the three Snf1-activating kinases of Saccharomyces cerevisiae. The identities of proteins associated with the Snf1-activating kinases were determined by peptide mass fingerpr...

  7. Spent Nuclear Fuel (SNF) Startup Plan to Operations

    International Nuclear Information System (INIS)

    GREGORY, J.R.

    2000-01-01

    This plan defines the approach that will be used to ensure the transition from initial startup to normal operations of the SNF operations--are performed in a safe, controlled, and deliberate manner. It provides a phased approach that bridges the operations between the completion of the ORR and the return to normal operations. This plan includes management oversight and administrative controls to be implemented and then reduced in a controlled manner until normal operations are authorized by SNF Management

  8. System/subsystem specifications for the Worldwide Port System (WPS) Regional Integrated Cargo Database (ICDB)

    Energy Technology Data Exchange (ETDEWEB)

    Rollow, J.P.; Shipe, P.C.; Truett, L.F. [Oak Ridge National Lab., TN (United States); Faby, E.Z.; Fluker, J.; Grubb, J.; Hancock, B.R. [Univ. of Tennessee, Knoxville, TN (United States); Ferguson, R.A. [Science Applications International Corp., Oak Ridge, TN (United States)

    1995-11-20

    A system is being developed by the Military Traffic Management Command (MTMC) to provide data integration and worldwide management and tracking of surface cargo movements. The Integrated Cargo Database (ICDB) will be a data repository for the WPS terminal-level system, will be a primary source of queries and cargo traffic reports, will receive data from and provide data to other MTMC and non-MTMC systems, will provide capabilities for processing Advance Transportation Control and Movement Documents (ATCMDs), and will process and distribute manifests. This System/Subsystem Specifications for the Worldwide Port System Regional ICDB documents the system/subsystem functions, provides details of the system/subsystem analysis in order to provide a communication link between developers and operational personnel, and identifies interfaces with other systems and subsystems. It must be noted that this report is being produced near the end of the initial development phase of ICDB, while formal software testing is being done. Following the initial implementation of the ICDB system, maintenance contractors will be in charge of making changes and enhancing software modules. Formal testing and user reviews may indicate the need for additional software units or changes to existing ones. This report describes the software units that are components of this ICDB system as of August 1995.

  9. Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying (CVD) Facility Operations Manual

    International Nuclear Information System (INIS)

    IRWIN, J.J.

    2000-01-01

    The mission of the Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying Facility (CVDF) is to achieve the earliest possible removal of free water from Multi-Canister Overpacks (MCOs). The MCOs contain metallic uranium SNF that have been removed from the 100K Area fuel storage water basins (i.e., the K East and K West Basins) at the US. Department of Energy Hanford Site in Southeastern Washington state. Removal of free water is necessary to halt water-induced corrosion of exposed uranium surfaces and to allow the MCOs and their SNF payloads to be safely transported to the Hanford Site 200 East Area and stored within the SNF Project Canister Storage Building (CSB). The CVDF is located within a few hundred yards of the basins, southwest of the 165KW Power Control Building and the 105KW Reactor Building. The site area required for the facility and vehicle circulation is approximately 2 acres. Access and egress is provided by the main entrance to the 100K inner area using existing roadways. The CVDF will remove free. water from the MCOs to reduce the potential for continued fuel-water corrosion reactions. The cold vacuum drying process involves the draining of bulk water from the MCO and subsequent vacuum drying. The MCO will be evacuated to a pressure of 8 torr or less and backfilled with an inert gas (helium). The MCO will be sealed, leak tested, and then transported to the CSB within a sealed shipping cask. (The MCO remains within the same shipping Cask from the time it enters the basin to receive its SNF payload until it is removed from the Cask by the CSB MCO handling machine.) The CVDF subproject acquired the required process systems, supporting equipment, and facilities. The cold vacuum drying operations result in an MCO containing dried fuel that is prepared for shipment to the CSB by the Cask transportation system. The CVDF subproject also provides equipment to dispose of solid wastes generated by the cold vacuum drying process and transfer process water removed

  10. SNF/HLW Transfer System Description Document

    International Nuclear Information System (INIS)

    W. Holt

    2005-01-01

    The purpose of this system description document (SDD) is to establish requirements that drive the design of the spent nuclear fuel (SNF)/high-level radioactive waste (HLW) transfer system and associated bases, which will allow the design effort to proceed to license application. This SDD will be revised at strategic points as the design matures. This SDD identifies the requirements and describes the system design, as it currently exists, with emphasis on attributes of the design provided to meet the requirements. This SDD is an engineering tool for design control. Accordingly, the primary audience and users are design engineers. This SDD is part of an iterative design process. It leads the design process with regard to the flowdown of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. The SDD follows the design with regard to the description of the system. The description provided in this SDD reflects the current results of the design process

  11. WPS-based technology for client-side remote sensing data processing

    Science.gov (United States)

    Kazakov, E.; Terekhov, A.; Kapralov, E.; Panidi, E.

    2015-04-01

    Server-side processing is principal for most of the current Web-based geospatial data processing tools. However, in some cases the client-side geoprocessing may be more convenient and acceptable. This study is dedicated to the development of methodology and techniques of Web services elaboration, which allow the client-side geoprocessing also. The practical objectives of the research are focused on the remote sensing data processing, which are one of the most resource-intensive data types. The idea underlying the study is to propose such geoprocessing Web service schema that will be compatible with the current serveroriented Open Geospatial Consortium standard (OGC WPS standard), and additionally will allow to run the processing on the client, transmitting processing tool (executable code) over the network instead of the data. At the same time, the unity of executable code must be preserved, and the transmitted code should be the same to that is used for server-side processing. This unity should provide unconditional identity of the processing results that performed using of any schema. The appropriate services are pointed by the authors as a Hybrid Geoprocessing Web Services (HGWSs). The common approaches to architecture and structure of the HGWSs are proposed at the current stage as like as a number of service prototypes. For the testing of selected approaches, the geoportal prototype was implemented, which provides access to created HGWS. Further works are conducted on the formalization of platform independent HGWSs implementation techniques, and on the approaches to conceptualization of theirs safe use and chaining possibilities. The proposed schema of HGWSs implementation could become one of the possible solutions for the distributed systems, assuming that the processing servers could play the role of the clients connecting to the service supply server. The study was partially supported by Russian Foundation for Basic Research (RFBR), research project No. 13

  12. Regulatory practices of radiation safety of SNF transportation in Russia

    International Nuclear Information System (INIS)

    Kuryndina, Lidia; Kuryndin, Anton; Stroganov, Anatoly

    2008-01-01

    This paper overviews current regulatory practices for the assurance of nuclear and radiation safety during railway transportation of SNF on the territory of Russian Federation from NPPs to longterm-storage of reprocessing sites. The legal and regulatory requirements (mostly compliant with IAEA ST-1), licensing procedure for NM transportation are discussed. The current procedure does not require a regulatory approval for each particular shipment if the SNF fully comply with the Rosatom's branch standard and is transported in approved casks. It has been demonstrated that SNF packages compliant with the branch standard, which is knowingly provide sufficient safety margin, will conform to the federal level regulations. The regulatory approval is required if a particular shipment does not comply with the branch standard. In this case, the shipment can be approved only after regulatory review of Applicant's documents to demonstrate that the shipment still conformant to the higher level (federal) regulations. The regulatory review frequently needs a full calculation test of the radiation safety assurance. This test can take a lot of time. That's why the special calculation tools were created in SEC NRS. These tools aimed for precision calculation of the radiation safety parameters by SNF transportation use preliminary calculated Green's functions. Such approach allows quickly simulate any source distribution and optimize spent fuel assemblies placement in cask due to the transport equation property of linearity relatively the source. The short description of calculation tools are presented. Also, the paper discusses foreseen implications related to transportation of mixed-oxide SNF. (author)

  13. Problems and solutions of the spent nuclear fuel (SNF) at Kozloduy NPP

    International Nuclear Information System (INIS)

    Jordanov, J.

    2003-01-01

    There are two options concerning spent nuclear fuel: to return it back to Russia for reprocessing or to store it on the site until we decide what to do with it. In both options prior to the shutting down of each reactor the Spent Fuel Pool thereto should be vacated (the filling in of the equipment at present is illustrated) and the Spent Fuel Storage Facility (SFSF) should also be vacated after the stop of the last nuclear facility on the site in order to be reequipped for permanent storage of the highly active wastes which will be returned in the country, if we submit the fuel for reprocessing; or of SNF, if we decide to leave them ultimately in Bulgaria. The difference is mainly in the quantities which will permanently remain here, respectively the volumes required for their storage and the funds necessary for the implementation of the processes. The pool volumes filling in both variants is also illustrated and the SFSF will be filled by 2008, if no fuel is transported.Costs of the SNF transport to Russia and investment costs of dry storage of SNF from pools 1 - 4 are present. The costs are visibly lower compared to those in the case of return of the fuel. However, these are only investments for construction and equipment of the buildings and storage containers. The costs related to their servicing are not included, and it should be taken into account that in approximately 50 years we will have to seek solution for their permanent storage. Despite the material costs to be incurred now for the implementation of the option with the return of the fuel, this is the more worthy way to resolve the problem. In accordance with the ethic principles in the nuclear energy, the burdens arising as a result of the use of nuclear facilities should be covered by the generation consuming the benefits from it

  14. Alteration to the SWI/SNF complex in human cancers

    Directory of Open Access Journals (Sweden)

    Vanessa S. Gordon

    2011-12-01

    Full Text Available The SWI/SNF complex is a key catalyst for gene expression and regulates a variety of pathways, many of which have anticancer roles. Its central roles in cellular growth control, DNA repair, differentiation, cell adhesion and development are often targeted, and inactivated, during cancer development and progression. In this review, we will discuss what is known about how SWI/SNF is inactivated, and describe the potential impact of abrogating this complex. BRG1 and BRM are the catalytic subunits which are essential for SWI/SNF function, and thus, it is not surprising that they are lost in a variety of cancer types. As neither gene is mutated when lost, the mechanism of suppression, as well as the impact of potential gene activity restoration, are reviewed.

  15. A User's Guide to the SNF ampersand INEL EIS

    International Nuclear Information System (INIS)

    1995-01-01

    This User's Guide is intended to help you find information in the SNF and INEL EIS (that's short for US Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Final Environmental Impact Statement). The first section of this Guide gives you a brief overview of the SNF ampersand INEL EIS., The second section is organized to help you find specific information in the Environmental Impact Statement -- whether you're interested in a management alternative, a particular site (such as Hanford), or a discipline (such as land use or water quality)

  16. Spent Nuclear Fuel (SNF) Bounding Drop Support Calculations

    International Nuclear Information System (INIS)

    CHENAULT, D.M.

    1999-01-01

    This report evaluates different drop heights, concrete and other impact media to which the transport package and/or the MCO is dropped. A prediction method is derived for estimating the resultant impact factor for determining the bounding drop case for the SNF Project

  17. Characterization of FRR SNF in Basin and Dry Storage Systems

    International Nuclear Information System (INIS)

    Brooks, H.M.; Sindelar, R.L.

    1998-09-01

    Since May 1996, over 1700 aluminum-based spent nuclear fuel (A1-SNF) assemblies have been inspected for corrosion and mechanical damage to determine if the cladding had been penetrated as part of the process for acceptance of the fuel at the Savannah River Site (SRS). The results of the release measurements are summarized in this paper

  18. 42 CFR 424.20 - Requirements for posthospital SNF care.

    Science.gov (United States)

    2010-10-01

    ... statements may be signed by— (1) The physician responsible for the case or, with his or her authorization, by a physician on the SNF staff or a physician who is available in case of an emergency and has knowledge of the case; or (2) A nurse practitioner or clinical nurse specialist, neither of whom has a...

  19. Fast Reactors and Nuclear Nonproliferation

    International Nuclear Information System (INIS)

    Avrorina, E.N.; Chebeskovb, A.N.

    2013-01-01

    Conclusion remarks: 1. Fast reactor start-up with U-Pu fuel: – dependent on thermal reactors, – no needs in U enrichment, – needs in SNF reprocessing, – Pu is a little suitable for NED, – practically impossible gun-type NED, – difficulties for implosion-type NED: necessary tests, advanced technologies, etc. – Pu in blankets is similar to WPu by isotopic composition, – Use of blanket for production isotopes (e.g. 233 U), – Combined reprocessing of SNF: altogether blanket and core, – Blanket elimination: decrease in Pu production – No pure Pu separation. 2. Fast reactor start-up with U fuel: - Needs in both U enrichment and SNF reprocessing, - Independent of thermal reactors, - Good Pu bred in the core let alone blankets, - NED of simple gun-type design, - Increase of needs in SWU, - Increased demands in U supply. 3. Fast reactors for export: - Uranium shortage, - To replace thermal reactors in future, - No blankets (depends on the country, though), - Fuel supply and SNF take back, - International centers for rendering services of NFC. Time has come to remove from FRs and their NFC the label unfairly identifying them as the most dangerous installations of nuclear power from the standpoint of being a proliferation problem

  20. A contingency safe, responsible, economic, increased capacity spent nuclear fuel (SNF) advance fuel cycle

    International Nuclear Information System (INIS)

    Levy, S.

    2008-01-01

    The purpose of this paper is to have an Advanced Light Water (LWR) fuel cycle and an associated development program to provide a contingency plan to the current DOE effort to license once-through spent Light Water Reactor (LWR) fuel for disposition at Yucca Mountain (YM). The intent is to fully support the forthcoming June 2008 DOE submittal to the Nuclear Regulatory Commission (NRC) based upon the latest DOE draft DOE/EIS-0250F-SID dated October 2007 which shows that the latest DOE YM doses would readily satisfy the anticipated NRC and Environmental Protection Agency (EP) standards. The proposed Advance Fuel Cycle can offer potential resolution of obstacles that might arise during the NRC review and, particularly, during the final hearings process to be held in Nevada. Another reason for the proposed concept is that a substantial capacity growth of the YM repository will be necessary to accommodate the SNF of Advance Light Water Reactors (ALWRs) currently under consideration for United States (U.S.) electricity production (1) and the results of the recently issued study by the Electric Power Research Institute (EPRI) to reduce CO 2 emissions (2). That study predicts that by 2030 U.S. nuclear power generation would grow by 64 Gigawatt electrical (GWe) and account for 25.5 percent of the overall U.S. electrical generation. The current annual SNF once-through fuel cycle accumulation would rise from 2000-2100 MT (Metric Tons) to about 3480 MT in 2030 and the total SNF inventory, would reach nearly 500,000 MT by 2100 if U. S. nuclear power continues to grow at 1.1 percent per year after 2030. That last projection does not account for any SNF reduction due to increased fuel burnup or any increased capacity needed 'to establish supply Global Nuclear Energy Partnership (GNEP,) arrangements among nations to provide nuclear fuel and taking back spent fuel for recycling without spreading enrichment and reprocessing technologies' (3). The anticipated capacity of 120 MT

  1. Spent Nuclear Fuel (SNF) Removal Campaign Plan

    International Nuclear Information System (INIS)

    PAJUNEN, A.L.

    2000-01-01

    The overall operation of the Spent Nuclear Fuel Project will include fuel removal, sludge removal, debris removal, and deactivation transition activities. Figure 1-1 provides an overview of the current baseline operating schedule for project sub-systems, indicating that a majority of fuel removal activities are performed over an approximately three-and-one-half year time period. The purpose of this document is to describe the strategy for operating the fuel removal process systems. The campaign plan scope includes: (1) identifying a fuel selection sequence during fuel removal activities, (2) identifying MCOs that are subjected to extra testing (process validation) and monitoring, and (3) discussion of initial MCO loading and monitoring in the Canister Storage Building (CSB). The campaign plan is intended to integrate fuel selection requirements for handling special groups of fuel within the basin (e.g., single pass reactor fuel), process validation activities identified for process systems, and monitoring activities during storage

  2. 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.)

  3. 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.)

  4. SNF fuel retrieval sub project safety analysis document

    International Nuclear Information System (INIS)

    BERGMANN, D.W.

    1999-01-01

    This safety analysis is for the SNF Fuel Retrieval (FRS) Sub Project. The FRS equipment will be added to K West and K East Basins to facilitate retrieval, cleaning and repackaging the spent nuclear fuel into Multi-Canister Overpack baskets. The document includes a hazard evaluation, identifies bounding accidents, documents analyses of the accidents and establishes safety class or safety significant equipment to mitigate accidents as needed

  5. SNF fuel retrieval sub project safety analysis document

    Energy Technology Data Exchange (ETDEWEB)

    BERGMANN, D.W.

    1999-02-24

    This safety analysis is for the SNF Fuel Retrieval (FRS) Sub Project. The FRS equipment will be added to K West and K East Basins to facilitate retrieval, cleaning and repackaging the spent nuclear fuel into Multi-Canister Overpack baskets. The document includes a hazard evaluation, identifies bounding accidents, documents analyses of the accidents and establishes safety class or safety significant equipment to mitigate accidents as needed.

  6. Purification and characterization of the three Snf1-activating kinases of Saccharomyces cerevisiae.

    Science.gov (United States)

    Elbing, Karin; McCartney, Rhonda R; Schmidt, Martin C

    2006-02-01

    Members of the Snf1/AMPK family of protein kinases are activated by distinct upstream kinases that phosphorylate a conserved threonine residue in the Snf1/AMPK activation loop. Recently, the identities of the Snf1- and AMPK-activating kinases have been determined. Here we describe the purification and characterization of the three Snf1-activating kinases of Saccharomyces cerevisiae. The identities of proteins associated with the Snf1-activating kinases were determined by peptide mass fingerprinting. These kinases, Sak1, Tos3 and Elm2 do not appear to require the presence of additional subunits for activity. Sak1 and Snf1 co-purify and co-elute in size exclusion chromatography, demonstrating that these two proteins form a stable complex. The Snf1-activating kinases phosphorylate the activation loop threonine of Snf1 in vitro with great specificity and are able to do so in the absence of beta and gamma subunits of the Snf1 heterotrimer. Finally, we showed that the Snf1 kinase domain isolated from bacteria as a GST fusion protein can be activated in vitro and shows substrate specificity in the absence of its beta and gamma subunits.

  7. FluorWPS: A Monte Carlo ray-tracing model to compute sun-induced chlorophyll fluorescence of three-dimensional canopy

    Science.gov (United States)

    A model to simulate radiative transfer (RT) of sun-induced chlorophyll fluorescence (SIF) of three-dimensional (3-D) canopy, FluorWPS, was proposed and evaluated. The inclusion of fluorescence excitation was implemented with the ‘weight reduction’ and ‘photon spread’ concepts based on Monte Carlo ra...

  8. Spent Nuclear Fuel (SNF) Project Design Verification and Validation Process

    International Nuclear Information System (INIS)

    OLGUIN, L.J.

    2000-01-01

    This document provides a description of design verification and validation activities implemented by the Spent Nuclear Fuel (SNF) Project. During the execution of early design verification, a management assessment (Bergman, 1999) and external assessments on configuration management (Augustenburg, 1999) and testing (Loscoe, 2000) were conducted and identified potential uncertainties in the verification process. This led the SNF Chief Engineer to implement corrective actions to improve process and design products. This included Design Verification Reports (DVRs) for each subproject, validation assessments for testing, and verification of the safety function of systems and components identified in the Safety Equipment List to ensure that the design outputs were compliant with the SNF Technical Requirements. Although some activities are still in progress, the results of the DVR and associated validation assessments indicate that Project requirements for design verification are being effectively implemented. These results have been documented in subproject-specific technical documents (Table 2). Identified punch-list items are being dispositioned by the Project. As these remaining items are closed, the technical reports (Table 2) will be revised and reissued to document the results of this work

  9. Structure and novel functional mechanism of Drosophila SNF in sex-lethal splicing.

    Directory of Open Access Journals (Sweden)

    Jicheng Hu

    Full Text Available Sans-fille (SNF is the Drosophila homologue of mammalian general splicing factors U1A and U2B'', and it is essential in Drosophila sex determination. We found that, besides its ability to bind U1 snRNA, SNF can also bind polyuridine RNA tracts flanking the male-specific exon of the master switch gene Sex-lethal (Sxl pre-mRNA specifically, similar to Sex-lethal protein (SXL. The polyuridine RNA binding enables SNF directly inhibit Sxl exon 3 splicing, as the dominant negative mutant SNF(1621 binds U1 snRNA but not polyuridine RNA. Unlike U1A, both RNA recognition motifs (RRMs of SNF can recognize polyuridine RNA tracts independently, even though SNF and U1A share very high sequence identity and overall structure similarity. As SNF RRM1 tends to self-associate on the opposite side of the RNA binding surface, it is possible for SNF to bridge the formation of super-complexes between two introns flanking Sxl exon 3 or between a intron and U1 snRNP, which serves the molecular basis for SNF to directly regulate Sxl splicing. Taken together, a new functional model for SNF in Drosophila sex determination is proposed. The key of the new model is that SXL and SNF function similarly in promoting Sxl male-specific exon skipping with SNF being an auxiliary or backup to SXL, and it is the combined dose of SXL and SNF governs Drosophila sex determination.

  10. Removal of spent fuel from the TVR reactor for reprocessing and proposals for the RA reactor spent fuel handling

    International Nuclear Information System (INIS)

    Volkov, E.B.; Konev, V.N.; Shvedov, O.V.; Bulkin, S.Yu; Sokolov, A.V.

    2002-01-01

    The 2,5 MW heavy-water moderated and cooled research reactor TVR was located at the Moscow Institute for Theoretical and Experimental Physics site. In 1990 the final batch of spent nuclear fuel (SNF) from the TVR reactor was transported for reprocessing to Production Association (PA) 'Mayak'. This transportation of the SNF was a part of TVR reactor decommissioning. The special technology and equipment was developed in order to fulfill the preparation of TVR SNF for transportation. The design of the TVR reactor and the fuel elements used are similar to the design and fuel elements of the RA reactor. Two different ways of RA spent fuel elements for transportation to reprocessing plant are considered: in aluminum barrels, and in additional cans. The experience and equipment used for the preparing TVR fuel elements for transportation can help the staff of RA reactor to find the optimal way for these technical operations. (author)

  11. Management of Hanford Site non-defense production reactor spent nuclear fuel, Hanford Site, Richland, Washington

    International Nuclear Information System (INIS)

    1997-03-01

    The US Department of Energy (DOE) needs to provide radiologically, and industrially safe and cost-effective management of the non-defense production reactor spent nuclear fuel (SNF) at the Hanford Site. The proposed action would place the Hanford Site's non-defense production reactor SNF in a radiologically- and industrially-safe, and passive storage condition pending final disposition. The proposed action would also reduce operational costs associated with storage of the non-defense production reactor SNF through consolidation of the SNF and through use of passive rather than active storage systems. Environmental, safety and health vulnerabilities associated with existing non-defense production reactor SNF storage facilities have been identified. DOE has determined that additional activities are required to consolidate non-defense production reactor SNF management activities at the Hanford Site, including cost-effective and safe interim storage, prior to final disposition, to enable deactivation of facilities where the SNF is now stored. Cost-effectiveness would be realized: through reduced operational costs associated with passive rather than active storage systems; removal of SNF from areas undergoing deactivation as part of the Hanford Site remediation effort; and eliminating the need to duplicate future transloading facilities at the 200 and 400 Areas. Radiologically- and industrially-safe storage would be enhanced through: (1) removal from aging facilities requiring substantial upgrades to continue safe storage; (2) utilization of passive rather than active storage systems for SNF; and (3) removal of SNF from some storage containers which have a limited remaining design life. No substantial increase in Hanford Site environmental impacts would be expected from the proposed action. Environmental impacts from postulated accident scenarios also were evaluated, and indicated that the risks associated with the proposed action would be small

  12. Hazard Evaluation for Storage of Spent Nuclear Fuel (SNF) Sludge at the Solid Waste Treatment Facility

    International Nuclear Information System (INIS)

    SCHULTZ, M.V.

    2000-01-01

    As part of the Spent Nuclear Fuel (SNF) storage basin clean-up project, sludge that has accumulated in the K Basins due to corrosion of damaged irradiated N Reactor will be loaded into containers and placed in interim storage. The Hanford Site Treatment Complex (T Plant) has been identified as the location where the sludge will be stored until final disposition of the material occurs. Long term storage of sludge from the K Basin fuel storage facilities requires identification and analysis of potential accidents involving sludge storage in T Plant. This report is prepared as the initial step in the safety assurance process described in DOE Order 5480.23, Nuclear Safety Analysis Reports and HNF-PRO-704, Hazards and Accident Analysis Process. This report documents the evaluation of potential hazards and off-normal events associated with sludge storage activities. This information will be used in subsequent safety analyses, design, and operations procedure development to ensure safe storage. The hazards evaluation for the storage of SNF sludge in T-Plant used the Hazards and Operability Analysis (HazOp) method. The hazard evaluation identified 42 potential hazardous conditions. No hazardous conditions involving hazardous/toxic chemical concerns were identified. Of the 42 items identified in the HazOp study, eight were determined to have potential for onsite worker consequences. No items with potential offsite consequences were identified in the HazOp study. Hazardous conditions with potential onsite worker or offsite consequences are candidates for quantitative consequence analysis. The hazardous conditions with potential onsite worker consequences were grouped into two event categories, Container failure due to overpressure - internal to T Plant, and Spill of multiple containers. The two event categories will be developed into accident scenarios that will be quantitatively analyzed to determine release consequences. A third category, Container failure due to

  13. Closing nuclear fuel cycle with fast reactors: problems and prospects

    Energy Technology Data Exchange (ETDEWEB)

    Shadrin, A.; Dvoeglazov, K.; Ivanov, V. [Bochvar Institute - VNIINM, Moscow (Russian Federation)

    2013-07-01

    The closed nuclear fuel cycle (CNFC) with fast reactors (FR) is the most promising way of nuclear energetics development because it prevents spent nuclear fuel (SNF) accumulation and minimizes radwaste volume due to minor actinides (MA) transmutation. CNFC with FR requires the elaboration of safety, environmentally acceptable and economically effective methods of treatment of SNF with high burn-up and low cooling time. The up-to-date industrially implemented SNF reprocessing technologies based on hydrometallurgical methods are not suitable for the reprocessing of SNF with high burn-up and low cooling time. The alternative dry methods (such as electrorefining in molten salts or fluoride technologies) applicable for such SNF reprocessing have not found implementation at industrial scale. So the cost of SNF reprocessing by means of dry technologies can hardly be estimated. Another problem of dry technologies is the recovery of fissionable materials pure enough for dense fuel fabrication. A combination of technical solutions performed with hydrometallurgical and dry technologies (pyro-technology) is proposed and it appears to be a promising way for the elaboration of economically, ecologically and socially accepted technology of FR SNF management. This paper deals with discussion of main principle of dry and aqueous operations combination that probably would provide safety and economic efficiency of the FR SNF reprocessing. (authors)

  14. Assessment results of the South Korea TRIGA SNF to be shipped to INEEL

    International Nuclear Information System (INIS)

    Cole, C.M.; Dirk, W.J.; Cottam, R.E.; Paik, S.T.

    1997-01-01

    This paper describes the Training, Research, Isotope, General Atomics (TRIGA) spent nuclear fuel (SNF) examination at the Seoul and the Taejon Research Reactor Facilities in South Korea. The examination was required before the SNF would be accepted for transportation and storage at the INEEL. The results of the aluminum and stainless steel clad TRIGA fuel examination have been summarized. A description of the examination team training, the examination work plan and examination equipment is also included. This paper also explains the technical basis for the examination and physical condition criteria used to determine what, if any, additional packaging would be required for transportation and for the receipt and storage of the fuel at the INEEL. This paper delineates the preparation activities prior to the fuel examinations and includes (1) collecting spent fuel data; (2) preparatory work by the Korean Atomic Energy Research Institute (KAERI) for fuel examination: (3) preparation of a radionuclide report, Radionuclide Mass Inventory, Activity, Decay Heat, and Dose Rate Parametric Data for TRIGA Spent Nuclear Fuels needed to provide input data for transportation and fuel acceptance at the Idaho National Engineering and Environmental Laboratory (INEEL); (4) gathering FRR Facility data; and (5) coordination between the INEEL and KAERI. Included, are the unanticipated conditions encountered in the unloading of fuel from the dry storage casks in Taejon in preparation for examination, a description of the damaged condition of the fuel removed from the casks, and the apparent cause of the damages. Lessons learned from all the activities are also addressed. A brief description of the preparatory work for the shipment of the spent fuel from Korea to INEEL is included

  15. Assessment results of the South Korea TRIGA SNF to be shipped to INEEL

    International Nuclear Information System (INIS)

    Cole, Charles M.; Dirk, Willam J.; Cottam, Russel E.; Paik, Sam T.

    1997-01-01

    This paper describes the Training, Research, Isotope, General Atomics (TRIGA) spent nuclear fuel (SNF) examination at the Seoul and the Taejon Research Reactor Facilities in South Korea. The examination was required before the SNF would be accepted for transportation and storage at the INEEL. The results of the aluminum and stainless steel clad TRIGA fuel examination have been summarized. A description of the examination team training, the examination work plan and examination equipment is also included. This paper also explains the technical basis for the examination and physical condition criteria used to determine what, if any, additional packaging (canning) would be required for transportation and for the receipt and storage of the fuel at the INEEL. This paper delineates the preparation activities prior to the fuel examinations and includes (1) collecting spent fuel data; (2) preparatory work by the Korean Atomic Energy Research Institute (KAERI) for fuel examination: (3) preparation of a radionuclide report, 'Radionuclide Mass Inventory, Activity, Decay Heat, and Dose Rate Parametric Data for TRIGA Spent Nuclear Fuels' needed to provide input data for transportation and fuel acceptance at the Idaho National Engineering and Environmental Laboratory (INEEL); (4) gathering FRR Facility data; (5) preparation of Appendix A; (6) and coordination between the INEEL and KAERI. Included, are the unanticipated conditions encountered in the unloading of fuel from the dry storage casks in Taejon in preparation for examination, a description of the damaged condition of the fuel removed from the casks, and the apparent cause of the damages. Lessons learned from all the activities are also addressed. A brief description of the preparatory work for the shipment of the spent fuel from Korea to INEEL is included. (author)

  16. Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

    International Nuclear Information System (INIS)

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D.

    2013-01-01

    General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2

  17. Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

    Energy Technology Data Exchange (ETDEWEB)

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D. [General Atomics 3550 General Atomics Court San Diego, CA 92130 (United States)

    2013-07-01

    General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2

  18. Management of Spent Nuclear Fuel from Nuclear Power Plant Reactor

    International Nuclear Information System (INIS)

    Wati, Nurokhim

    2008-01-01

    Management of spent nuclear fuel from Nuclear Power Plant (NPP) reactor had been studied to anticipate program of NPP operation in Indonesia. In this paper the quantity of generated spent nuclear fuel (SNF) is predicted based on the national electrical demand, power grade and type of reactor. Data was estimated using Pressurized Water Reactor (PWR) NPP type 1.000 MWe and the SNF management overview base on the experiences of some countries that have NPP. There are four strategy nuclear fuel cycle which can be developed i.e: direct disposal, reprocessing, DUPlC (Direct Use of Spent PWR Fuel In Candu) and wait and see. There are four alternative for SNF management i.e : storage at the reactor building (AR), away from reactor (AFR) using wet centralized storage, dry centralized storage AFR and prepare for reprocessing facility. For the Indonesian case, centralized facility of the wet type is recommended for PWR or BWR spent fuel. (author)

  19. The remote methods for radwaste and SNF control

    International Nuclear Information System (INIS)

    Ivanov, O; Stepanov, V; Danilovich, A; Potapov, V

    2017-01-01

    With the examples of developments carried out in the Kurchatov Institute and by the world leaders in the field the presentation considers the devices and methods to obtain remotely information on the distribution of radioactivity in radwaste and SNF. It describes the different types of light portable gamma cameras. The application of scanning spectrometric systems is considers also. The methods of recording UV radiation for detection of alpha contamination with the luminescence of air are presented. We discuss the scope and tasks that can be solved using remote and non-destructive methods. (paper)

  20. Spent Nuclear Fuel (SNF) Project Safety Basis Implementation Strategy

    International Nuclear Information System (INIS)

    TRAWINSKI, B.J.

    2000-01-01

    The objective of the Safety Basis Implementation is to ensure that implementation of activities is accomplished in order to support readiness to move spent fuel from K West Basin. Activities may be performed directly by the Safety Basis Implementation Team or they may be performed by other organizations and tracked by the Team. This strategy will focus on five key elements, (1) Administration of Safety Basis Implementation (general items), (2) Implementing documents, (3) Implementing equipment (including verification of operability), (4) Training, (5) SNF Project Technical Requirements (STRS) database system

  1. Subunits of the Snf1 kinase heterotrimer show interdependence for association and activity.

    Science.gov (United States)

    Elbing, Karin; Rubenstein, Eric M; McCartney, Rhonda R; Schmidt, Martin C

    2006-09-08

    The Snf1 kinase and its mammalian orthologue, the AMP-activated protein kinase (AMPK), function as heterotrimers composed of a catalytic alpha-subunit and two non-catalytic subunits, beta and gamma. The beta-subunit is thought to hold the complex together and control subcellular localization whereas the gamma-subunit plays a regulatory role by binding to and blocking the function of an auto-inhibitory domain (AID) present in the alpha-subunit. In addition, catalytic activity requires phosphorylation by a distinct upstream kinase. In yeast, any one of three Snf1-activating kinases, Sak1, Tos3, or Elm1, can fulfill this role. We have previously shown that Sak1 is the only Snf1-activating kinase that forms a stable complex with Snf1. Here we show that the formation of the Sak1.Snf1 complex requires the beta- and gamma-subunits in vivo. However, formation of the Sak1.Snf1 complex is not necessary for glucose-regulated phosphorylation of the Snf1 activation loop. Snf1 kinase purified from cells lacking the beta-subunits do not contain any gamma-subunit, indicating that the Snf1 kinase does not form a stable alphagamma dimer in vivo. In vitro kinase assays using purified full-length and truncated Snf1 proteins demonstrate that the kinase domain, which lacks the AID, is significantly more active than the full-length Snf1 protein. Addition of purified beta- and gamma-subunits could stimulate the kinase activity of the full-length alpha-subunit but only when all three subunits were present, suggesting an interdependence of all three subunits for assembly of a functional complex.

  2. A Multi-function Cask for At-Reactor Storage of Short-Cooled Spent Fuel, Transport, and Disposal

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    2004-01-01

    The spent nuclear fuel (SNF) system in the United States was designed with the assumptions that SNF would be stored for several years in an at-reactor pool and then transported to reprocessing plants for recovery of fissile materials, that security would not be a major issue, and that the SNF burnups would be low. The system has evolved into a once-through fuel cycle with high-burnup SNF, long-term storage at the reactor sites, and major requirements for safeguards and security. An alternative system is proposed to better meet these current requirements. The SNF is placed in multi-function casks with the casks used for at-reactor storage, transport, and repository disposal. The cask is the handling package, provides radiation shielding, and protects the SNF against accidents and assault. SNF assemblies are handled only once to minimize accident risks, maximize security and safeguards by minimizing access to SNF, and reduce costs. To maximize physical protection, the cask body is constructed of a cermet (oxide particles embedded in steel, the same class of materials used in tank armor) and contains no cooling channels or other penetrations that allow access to the SNF. To minimize pool storage of SNF, the cask is designed to accept short-cooled SNF. To maximize the capability of the cask to reject decay heat and to limit SNF temperatures from short-cooled SNF, the cask uses (1) natural circulation of inert gas mixtures inside the cask to transfer heat from the SNF to the cask body and (2) an overpack with external natural-circulation, liquid-cooled fins to transfer heat from the cask body to the atmosphere. This approach utilizes the entire cask body area for heat transfer to maximize heat removal rates-without any penetrations through the cask body that would reduce the physical protection capabilities of the cask body. After the SNF has cooled, the cooling overpack is removed. At the repository, the cask is placed in a corrosion-resistant overpack before disposal

  3. Warm pre-stress experiments on highly irradiated reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Landron, C.; Ait-Bachir, M.; Moinereau, D.; Molinie, E.; Garbay, E.

    2015-01-01

    In the aim to justify in-service integrity of reactor pressure vessel beyond 40 years, experimental warm pre-stress (WPS) tests were performed on irradiated materials representative of RPV steels corresponding to 40 operating years. Different types of WPS loading path have been considered to cover typical postulated accidental transients. These results confirmed the beneficial effect of WPS on the cleavage fracture resistance of the irradiated materials. No fracture occurred during the cooling phase of the loading path and the fracture toughness values are higher than that measured with conventional isothermal tests. The analyses of the experiments, conducted using either simplified engineering models or more refined fracture models based on local approach to cleavage fracture, are in agreement with the experimental results. (authors)

  4. Spent nuclear fuel management system of research reactor KINR NAS of Ukraine

    International Nuclear Information System (INIS)

    Slisenko, V.I.; Makarovskij, V.N.; Mazina, N.I.; D'yakov, A.G.; Andronov, O.B.

    2007-01-01

    The purpose of work - development of optimum, technically simple and reliable system of SNF management of research nuclear reactor taking into account received statistics and experience of long-term operation of reactor. The objective of this work - to place existing system of SNF management in conformity with modern normative requirements on the basis of 'ALARA' principle and taking into account the specificity of the object of application

  5. SNF sludge treatment system preliminary project execution plan

    International Nuclear Information System (INIS)

    Flament, T.A.

    1998-01-01

    The Fluor Daniel Hanford, Inc. (FDH) Project Director for the Spent Nuclear Fuel (SNF) Project has requested Numatec Hanford Company (NHC) to define how Hanford would manage a new subproject to provide a process system to receive and chemically treat radioactive sludge currently stored in the 100 K Area fuel retention basins. The subproject, named the Sludge Treatment System (STS) Subproject, provides and operates facilities and equipment to chemically process K Basin sludge to meet Tank Waste Remediation System (TWRS) requirements. This document sets forth the NHC management approach for the STS Subproject and will comply with the requirements of the SNF Project Management Plan (HNF-SD-SNFPMP-011). This version of this document is intended to apply to the initial phase of the subproject and to evolve through subsequent revision to include all design, fabrication, and construction conducted on the project and the necessary management and engineering functions within the scope of the subproject. As Project Manager, NHC will perform those activities necessary to complete the STS Subproject within approved cost and schedule baselines and turn over to FDH facilities, systems, and documentation necessary for operation of the STS

  6. Measurements of Fundamental Fluid Physics of SNF Storage Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Condie, Keith Glenn; Mc Creery, Glenn Ernest; McEligot, Donald Marinus

    2001-09-01

    With the University of Idaho, Ohio State University and Clarksean Associates, this research program has the long-term goal to develop reliable predictive techniques for the energy, mass and momentum transfer plus chemical reactions in drying / passivation (surface oxidation) operations in the transfer and storage of spent nuclear fuel (SNF) from wet to dry storage. Such techniques are needed to assist in design of future transfer and storage systems, prediction of the performance of existing and proposed systems and safety (re)evaluation of systems as necessary at later dates. Many fuel element geometries and configurations are accommodated in the storage of spent nuclear fuel. Consequently, there is no one generic fuel element / assembly, storage basket or canister and, therefore, no single generic fuel storage configuration. One can, however, identify generic flow phenomena or processes which may be present during drying or passivation in SNF canisters. The objective of the INEEL tasks was to obtain fundamental measurements of these flow processes in appropriate parameter ranges.

  7. Sp1 and CREB regulate basal transcription of the human SNF2L gene

    International Nuclear Information System (INIS)

    Xia Yu; Jiang Baichun; Zou Yongxin; Gao Guimin; Shang Linshan; Chen Bingxi; Liu Qiji; Gong Yaoqin

    2008-01-01

    Imitation Switch (ISWI) is a member of the SWI2/SNF2 superfamily of ATP-dependent chromatin remodelers, which are involved in multiple nuclear functions, including transcriptional regulation, replication, and chromatin assembly. Mammalian genomes encode two ISWI orthologs, SNF2H and SNF2L. In order to clarify the molecular mechanisms governing the expression of human SNF2L gene, we functionally examined the transcriptional regulation of human SNF2L promoter. Reporter gene assays demonstrated that the minimal SNF2L promoter was located between positions -152 to -86 relative to the transcription start site. In this region we have identified a cAMP-response element (CRE) located at -99 to -92 and a Sp1-binding site at -145 to -135 that play a critical role in regulating basal activity of human SNF2L gene, which were proven by deletion and mutation of specific binding sites, EMSA, and down-regulating Sp1 and CREB via RNAi. This study provides the first insight into the mechanisms that control basal expression of human SNF2L gene

  8. Trehalose-6-phosphate synthesis controls yeast gluconeogenesis downstream and independent of SNF1.

    Science.gov (United States)

    Deroover, Sofie; Ghillebert, Ruben; Broeckx, Tom; Winderickx, Joris; Rolland, Filip

    2016-06-01

    Trehalose-6-P (T6P), an intermediate of trehalose biosynthesis, was identified as an important regulator of yeast sugar metabolism and signaling. tps1Δ mutants, deficient in T6P synthesis (TPS), are unable to grow on rapidly fermentable medium with uncontrolled influx in glycolysis, depletion of ATP and accumulation of sugar phosphates. However, the exact molecular mechanisms involved are not fully understood. We show that SNF1 deletion restores the tps1Δ growth defect on glucose, suggesting that lack of TPS hampers inactivation of SNF1 or SNF1-regulated processes. In addition to alternative, non-fermentable carbon metabolism, SNF1 controls two major processes: respiration and gluconeogenesis. The tps1Δ defect appears to be specifically associated with deficient inhibition of gluconeogenesis, indicating more downstream effects. Consistently, Snf1 dephosphorylation and inactivation on glucose medium are not affected, as confirmed with an in vivo Snf1 activity reporter. Detailed analysis shows that gluconeogenic Pck1 and Fbp1 expression, protein levels and activity are not repressed upon glucose addition to tps1Δ cells, suggesting a link between the metabolic defect and persistent gluconeogenesis. While SNF1 is essential for induction of gluconeogenesis, T6P/TPS is required for inactivation of gluconeogenesis in the presence of glucose, downstream and independent of SNF1 activity and the Cat8 and Sip4 transcription factors. © FEMS 2016. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com.

  9. K Basins Spent Nuclear Fuel (SNF) Project Safety Analysis Report for Packaging (SARP) approval plan

    International Nuclear Information System (INIS)

    1995-01-01

    This document delineates the plan for preparation, review, and approval of the K Basins Spent Nuclear Fuel (SNF) Packaging Design Criteria (PDC) document and the on-site Safety Analysis Report for Packaging (SARP). The packaging addressed in these documents is used to transport SNF in a Multi- canister Overpack (MCO) configuration

  10. Geochemical Interactions in failed Co-Disposal Waste Packages for N Reactor and Ft. St. Vrain Spent Fuel and the Melt and Dilute Waste Form

    International Nuclear Information System (INIS)

    Arthur, S.E.; McNeish, J.

    2002-01-01

    The objective of this scientific analysis is to calculate the long-term geochemical behavior in a failed co-disposal waste package (WP) containing U. S. Department of Energy (DOE) spent nuclear fuel (SNF) and high level waste (HLW) glass. This analysis was prepared according to a Technical Work Plan (BSC 2002). Specifically the scope of these calculations is to determine: (1) The geochemical characteristics of the fluids inside the WP after breach, including the corrosion/dissolution of the initial WP configuration; (2) The transport of radionuclides of concern to performance assessment out of the degraded WP by infiltrating water; and (3) The range of parameter variation for additional laboratory and numerical evaluations. This analysis is limited to three SNF groups, uranium (U)/thorium (Th) carbide SNF (Group 5), U metal SNF (Group 7), and aluminum(Al)-based fuels (Group 9). Group 5 is represented by Ft. St. Vrain (FSV) U/Th carbide SNF, Group 7 is represented by N-Reactor U metal SNF, and Group 9 is represented by the Melt and Dilute (MandD) waste form developed from Al-based SNF. The DOE (2001a, Appendix A) describes all of these fuels. Table 1 shows the groups of DOE SNF, the representative SNF for each group, and the metric tons of heavy metal (MTHM) of SNF in each group

  11. Human error prediction and countermeasures based on CREAM in spent nuclear fuel (SNF) transportation

    International Nuclear Information System (INIS)

    Kim, Jae San

    2007-02-01

    Since the 1980s, in order to secure the storage capacity of spent nuclear fuel (SNF) at NPPs, SNF assemblies have been transported on-site from one unit to another unit nearby. However in the future the amount of the spent fuel will approach capacity in the areas used, and some of these SNFs will have to be transported to an off-site spent fuel repository. Most SNF materials used at NPPs will be transported by general cargo ships from abroad, and these SNFs will be stored in an interim storage facility. In the process of transporting SNF, human interactions will involve inspecting and preparing the cask and spent fuel, loading the cask onto the vehicle or ship, transferring the cask as well as storage or monitoring the cask. The transportation of SNF involves a number of activities that depend on reliable human performance. In the case of the transport of a cask, human errors may include spent fuel bundle misidentification or cask transport accidents among others. Reviews of accident events when transporting the Radioactive Material (RAM) throughout the world indicate that human error is the major causes for more than 65% of significant events. For the safety of SNF transportation, it is very important to predict human error and to deduce a method that minimizes the human error. This study examines the human factor effects on the safety of transporting spent nuclear fuel (SNF). It predicts and identifies the possible human errors in the SNF transport process (loading, transfer and storage of the SNF). After evaluating the human error mode in each transport process, countermeasures to minimize the human error are deduced. The human errors in SNF transportation were analyzed using Hollnagel's Cognitive Reliability and Error Analysis Method (CREAM). After determining the important factors for each process, countermeasures to minimize human error are provided in three parts: System design, Operational environment, and Human ability

  12. Co-evolution of SNF spliceosomal proteins with their RNA targets in trans-splicing nematodes.

    Science.gov (United States)

    Strange, Rex Meade; Russelburg, L Peyton; Delaney, Kimberly J

    2016-08-01

    Although the mechanism of pre-mRNA splicing has been well characterized, the evolution of spliceosomal proteins is poorly understood. The U1A/U2B″/SNF family (hereafter referred to as the SNF family) of RNA binding spliceosomal proteins participates in both the U1 and U2 small interacting nuclear ribonucleoproteins (snRNPs). The highly constrained nature of this system has inhibited an analysis of co-evolutionary trends between the proteins and their RNA binding targets. Here we report accelerated sequence evolution in the SNF protein family in Phylum Nematoda, which has allowed an analysis of protein:RNA co-evolution. In a comparison of SNF genes from ecdysozoan species, we found a correlation between trans-splicing species (nematodes) and increased phylogenetic branch lengths of the SNF protein family, with respect to their sister clade Arthropoda. In particular, we found that nematodes (~70-80 % of pre-mRNAs are trans-spliced) have experienced higher rates of SNF sequence evolution than arthropods (predominantly cis-spliced) at both the nucleotide and amino acid levels. Interestingly, this increased evolutionary rate correlates with the reliance on trans-splicing by nematodes, which would alter the role of the SNF family of spliceosomal proteins. We mapped amino acid substitutions to functionally important regions of the SNF protein, specifically to sites that are predicted to disrupt protein:RNA and protein:protein interactions. Finally, we investigated SNF's RNA targets: the U1 and U2 snRNAs. Both are more divergent in nematodes than arthropods, suggesting the RNAs have co-evolved with SNF in order to maintain the necessarily high affinity interaction that has been characterized in other species.

  13. 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...

  14. SNF project's MCO compliance assessment with DOE ''general design criteria,'' order 6430.1A and ''SNF project MCO additional NRC requirements,'' HNF-SD-SNF-DB-005

    International Nuclear Information System (INIS)

    GOLDMANN, L.H.

    1999-01-01

    This document is presented to demonstrate the MCOs compliance to the major design criteria invoked on the MCO. This document is broken down into a section for the MCO's evaluation against DOE Order 6430.1A General Design Criteria sixteen divisions and then the evaluation of the MCO against HNF-SD-SNF-DB-005 ''Spent Nuclear Fuel Project Multi-Canister Overpack Additional NRC Requirements.'' The compliance assessment is presented as a matrix in tabular form. The MCO is the primary container for the K-basin's spent nuclear fuel as it leaves the basin pools and through to the 40 year interim storage at the Canister Storage Building (CSB). The MCO and its components interface with; the K basins, shipping cask and transportation system, Cold Vacuum Drying facility individual process bays and equipment, and CSB facility including the MCO handling machine (MHM), the storage tubes, and the MCO work stations where sampling, welding, and inspection of the MCO is performed. As the MCO is the primary boundary for handling, process, and storage, its main goals are to minimize the spread of its radiological contents to the outside of the MCO and provide for nuclear criticality control. The MCO contains personnel radiation shielding only on its upper end, in the form of a shield plug, where the process interfaces are located. Shielding beyond the shield plug is the responsibility of the using facilities. The design of the MCO and its components is depicted in drawings H-2-828040 through H-2-828075. Not every drawing number in the sequence is used. The first drawing number, H-2-828040, is the drawing index for the MCO. The design performance specification for the MCO is HW-S-0426, and was reviewed and approved by the interfacing design authorities, the safety, regulatory, and operations groups, and the local DOE office. The current revision for the design performance specification is revision 5. The designs of the MCO have been reviewed and approved in a similar way and the reports

  15. Reconstruction of the yeast Snf1 kinase regulatory network reveals its role as a global energy regulator

    Science.gov (United States)

    Usaite, Renata; Jewett, Michael C; Oliveira, Ana Paula; Yates, John R; Olsson, Lisbeth; Nielsen, Jens

    2009-01-01

    Highly conserved among eukaryotic cells, the AMP-activated kinase (AMPK) is a central regulator of carbon metabolism. To map the complete network of interactions around AMPK in yeast (Snf1) and to evaluate the role of its regulatory subunit Snf4, we measured global mRNA, protein and metabolite levels in wild type, Δsnf1, Δsnf4, and Δsnfsnf4 knockout strains. Using four newly developed computational tools, including novel DOGMA sub-network analysis, we showed the benefits of three-level ome-data integration to uncover the global Snf1 kinase role in yeast. We for the first time identified Snf1's global regulation on gene and protein expression levels, and showed that yeast Snf1 has a far more extensive function in controlling energy metabolism than reported earlier. Additionally, we identified complementary roles of Snf1 and Snf4. Similar to the function of AMPK in humans, our findings showed that Snf1 is a low-energy checkpoint and that yeast can be used more extensively as a model system for studying the molecular mechanisms underlying the global regulation of AMPK in mammals, failure of which leads to metabolic diseases. PMID:19888214

  16. Available reprocessing and recycling services for research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tozser, Sandor; Marshall, Frances M.; Adelfang, Pablo; Bradley, Edward [International Atomic Energy Agency, Vienna (Austria); Budu, Madalina Elena [SOSNY Research and Development Company, Moscow (Russian Federation); Chiguer, Mustapha [AREVA, Paris La Defense (France)

    2016-03-15

    International activities in the back end of the research reactor (RR) fuel cycle have so far been dominated by the programmes of acceptance of highly-enriched uranium (HEU) spent nuclear fuel (SNF) by the country where it was originally enriched. In the future inventories of LEU SNF will continue to be created and the back end solution of RR SNF remains a critical issue. The IAEA, based on the experience gained during the decade of international cooperation in supporting the objectives of the HEU take-back programmes, drew up a report presenting available reprocessing and recycling services for RR SNF. This paper gives an overview of the report, which will address all aspects of reprocessing and recycling services for RR SNF.

  17. SNF Interim Storage Canister Corrosion and Surface Environment Investigations

    International Nuclear Information System (INIS)

    Bryan, Charles R.; Enos, David G.

    2015-01-01

    This progress report describes work being done at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. In order for SCC to occur, three criteria must be met. A corrosive environment must be present on the canister surface, the metal must susceptible to SCC, and sufficient tensile stress to support SCC must be present through the entire thickness of the canister wall. SNL is currently evaluating the potential for each of these criteria to be met.

  18. SNF Interim Storage Canister Corrosion and Surface Environment Investigations

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Enos, David G. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    This progress report describes work being done at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. In order for SCC to occur, three criteria must be met. A corrosive environment must be present on the canister surface, the metal must susceptible to SCC, and sufficient tensile stress to support SCC must be present through the entire thickness of the canister wall. SNL is currently evaluating the potential for each of these criteria to be met.

  19. Potential dispositioning flowsheets for ICPP SNF and wastes

    Energy Technology Data Exchange (ETDEWEB)

    Olson, A.L. [ed.; Anderson, P.A.; Bendixsen, C.L. [and others

    1995-11-01

    The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Laboratory (INEL), has reprocessed irradiated nuclear fuels for the US Department of Energy (DOE) since 1953. This activity resulted mainly in the recovery of uranium and the management of the resulting wastes. The acidic radioactive high-level liquid waste was routinely stored in stainless steel tanks and then calcined to form a dry granular solid. The calcine is stored in stainless steel bins that are housed in underground concrete vaults. In April 1992, the DOE discontinued the practice of reprocessing irradiated nuclear fuels. This decision has left a legacy of 1.8 million gallons of radioactive liquid wastes (1.5 million gallons of radioactive sodium-bearing liquid wastes and 0.3 million gallons of high-level liquid waste), 3800 cubic meters of calcine waste, and 289 metric tons of heavy metal within unprocessed spent nuclear fuel (SNF) left in inventory at the ICPP. The nation`s radioactive waste policy has been established by the Nuclear Waste Policy Act (NWPA), which requires the final disposal of SNF and radioactive waste in accordance with US Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) standards. In accordance with these regulations and other legal agreements between the State of Idaho and the DOE, the DOE must, among other requirements, (1) complete a final Environmental Impact Statement by April 30, 1995, (2) evaluate and test sodium-bearing waste pre-treatment technologies, (3) select the sodium-bearing and calcine waste pre-treatment technology, if necessary, by June 1, 1995, and (4) select a technology for converting calcined waste into an appropriate disposal form by June 1, 1995.

  20. Potential dispositioning flowsheets for ICPP SNF and wastes

    International Nuclear Information System (INIS)

    Olson, A.L.; Anderson, P.A.; Bendixsen, C.L.

    1995-11-01

    The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Laboratory (INEL), has reprocessed irradiated nuclear fuels for the US Department of Energy (DOE) since 1953. This activity resulted mainly in the recovery of uranium and the management of the resulting wastes. The acidic radioactive high-level liquid waste was routinely stored in stainless steel tanks and then calcined to form a dry granular solid. The calcine is stored in stainless steel bins that are housed in underground concrete vaults. In April 1992, the DOE discontinued the practice of reprocessing irradiated nuclear fuels. This decision has left a legacy of 1.8 million gallons of radioactive liquid wastes (1.5 million gallons of radioactive sodium-bearing liquid wastes and 0.3 million gallons of high-level liquid waste), 3800 cubic meters of calcine waste, and 289 metric tons of heavy metal within unprocessed spent nuclear fuel (SNF) left in inventory at the ICPP. The nation's radioactive waste policy has been established by the Nuclear Waste Policy Act (NWPA), which requires the final disposal of SNF and radioactive waste in accordance with US Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) standards. In accordance with these regulations and other legal agreements between the State of Idaho and the DOE, the DOE must, among other requirements, (1) complete a final Environmental Impact Statement by April 30, 1995, (2) evaluate and test sodium-bearing waste pre-treatment technologies, (3) select the sodium-bearing and calcine waste pre-treatment technology, if necessary, by June 1, 1995, and (4) select a technology for converting calcined waste into an appropriate disposal form by June 1, 1995

  1. Main Principles of the Perspective System of SNF Management in Russia - 13333

    International Nuclear Information System (INIS)

    Baryshnikov, Mikhail

    2013-01-01

    For the last several years the System of the Spent Nuclear Fuel management in Russia was seriously changed. The paper describes the main principles of the changes and the bases of the Perspective System of SNF Management in Russia. Among such the bases there are the theses with the interesting names like 'total knowledge', 'pollutant pays' and 'pay and forget'. There is also a brief description of the modern Russian SNF Management Infrastructure. And an outline of the whole System. The System which is - in case of Russia - is quite necessary to adjust SNF accumulation and to utilize the nuclear heritage. (authors)

  2. 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.)

  3. Evaluation of Test Methodologies for Dissolution and Corrosion of Al-SNF

    International Nuclear Information System (INIS)

    Wiersma, B.J.; Mickalonis, J.I.; Louthan, M.R.

    1998-09-01

    The performance of aluminum-based spent nuclear fuel (Al-SNF) in the repository will differ from that of the commercial nuclear fuels and the high level waste glasses. The program consists of evaluating three test methods

  4. Legal precedents regarding use and defensibility of risk assessment in Federal transportation of SNF and HLW

    International Nuclear Information System (INIS)

    Bentz, E.J. Jr.; Bentz, C.B.; O'Hora, T.D.; Chen, S.Y.

    1997-01-01

    Risk assessment has become an increasingly important and essential tool in support of Federal decision-making regarding the handling, storage, disposal, and transportation of spent nuclear fuel (SNF) and high-level radioactive waste (HLW). This paper analyzes the current statutory and regulatory framework and related legal precedents with regard to SNF and HLW transportation. The authors identify key scientific and technical issues regarding the use and defensibility of risk assessment in Federal decision-making regarding anticipated shipments

  5. Security preparation for receipt of SNF from the FRR to the INEEL

    International Nuclear Information System (INIS)

    Dahlquist, R.L.

    1997-01-01

    This paper reports the key security related activities associated with the FRR shipment. Starting with transportation of the SNF in the country of origin to the final destination at the INEEL. Methodology for compliance will be addressed. The graded approach and a three step system will be explained. This paper will be used as part of the planning to support the FRR Project for returning the Asia and European SNF back to the US

  6. Enhanced amino acid utilization sustains growth of cells lacking Snf1/AMPK

    DEFF Research Database (Denmark)

    Nicastro, Raffaele; Tripodi, Farida; Guzzi, Cinzia

    2015-01-01

    when grown with glucose excess. We show that loss of Snf1 in cells growing in 2% glucose induces an extensive transcriptional reprogramming, enhances glycolytic activity, fatty acid accumulation and reliance on amino acid utilization for growth. Strikingly, we demonstrate that Snf1/AMPK-deficient cells...... remodel their metabolism fueling mitochondria and show glucose and amino acids addiction, a typical hallmark of cancer cells....

  7. Role of Snf3 in glucose homeostasis of Saccharomyces cerevisiae (review)

    DEFF Research Database (Denmark)

    Kielland-Brandt, Morten

    signal pathways in directions opposite to those caused by extracellular nutrients (6,7), a phenomenon predicted to contribute to intracellular nutrient homeostasis. Although significant, the influence of intracellular leucine on signaling from Ssy1 is relatively modest (6), whereas the conditions...... with enhanced intracellular glucose concentrations (7) caused a strong decrease in signaling from Snf3, suggesting an important role of Snf3 in intracellular glucose homeostasis. Strategies for studies of this role will be discussed....

  8. Life-cycle cost analysis for Foreign Research Reactor, Spent Nuclear Fuel disposal

    International Nuclear Information System (INIS)

    Parks, P.B.; Geddes, R.L.; Jackson, W.N.; McDonell, W.R.; Dupont, M.E.; McWhorter, D.L.; Liutkus, A.S.

    1994-01-01

    DOE-EM-37 requested a life-cycle cost analysis for disposal of the Foreign Research Reactor-Spent Nuclear Fuel (FRR-SNF). The analysis was to address life-cycle and unit costs for a range of FRR-SNF elements from those currently available (6,000 elements) to the (then) bounding case (15,000 elements). Five alternative disposition strategies were devised for the FRR-SNF elements. Life-cycle costs were computed for each strategy. In addition, the five strategies were evaluated in terms of six societal and technical goals. This report summarizes the study that was originally documented to DOE-EM

  9. Complexon Solutions in Freon for Decontamination of Solids and SNF Treatment

    International Nuclear Information System (INIS)

    Kamachev, V.; Shadrin, A.; Murzin, A.

    2008-01-01

    Full text of publication follows: The possibility of using complexon solutions in supercritical and compressed carbon dioxide for decontamination of solid surfaces and for spent nuclear fuel (SNF) treatment was demonstrated in the works of Japanese, Russian and American researchers. The obtained data showed that the use of complexon solutions in carbon dioxide sharply decreases the volume of secondary radioactive wastes because it can be easily evaporated, purified and recycled. Moreover, high penetrability of carbon dioxide allows decontamination of surfaces with complex shape. However, one of the disadvantages of carbon dioxide is its high working pressure (10-20 MPa for supercritical CO 2 and 7 MPa for compressed CO 2 ). Moreover, in case of SNF treatment, carbon dioxide solvent will be contaminated with 14 C, which in the course of SNF dissolution in CO 2 containing TBP*HNO 3 adduct stage will be oxidized into CO 2 . These main disadvantages can be eliminated by using complexon solutions in ozone-friendly Freon HFC-134a for decontamination and SNF treatment. Our experimental data for real contaminated materials showed that the decontamination factor for complexon solutions in liquid Freon HFC-134a at 1,2 MPa and 25 deg. C is close to that attained in carbon dioxide. Moreover, the possibility of SNF treatment in Freon HFC-134a was demonstrated in trials using real SNF and its imitators. (authors)

  10. Design analysis of various transportation package options for BN-350 SNF in terms of nuclear radiation safety in planning for long-terms dry storage

    International Nuclear Information System (INIS)

    Aisabekov, A.Z.; Mukenova, S.A.; Tur, E.S.; Tsyngaev, V.M.

    2004-01-01

    Full text: This effort is performed under the BN-350 reactor facility decommissioning project. One of the project tasks - spent nuclear fuel handling - includes the following: fuel packaging into sealed canisters, transportation of the canisters in multi-seat metallo-concrete containers and placement of the containers for a long-term dry storage. The goal of this effort is to computationally validate nuclear and radiation safety of the SNF containers placed for storage both under normal storage conditions and probable accident situations. The basic unit structure and design configurations are presented: assemblies, canisters, transportation containers. The major factors influencing nuclear and radiation safety are presented: fuel burn-up, enrichment, fabrication tolerance, types of fuel assemblies, configuration of assemblies in the canister and canisters in the container, background of assemblies placed in the reactor and cooling pool. Conditions under which the SNF containers will be stored are described and probable accident situations are listed. Proceeding from the conservatism principle, selection of the assemblies posing the greatest nuclear hazard is validated. A neutron effective multiplication factor is calculated for the SNF containers under the normal storage conditions and for the case of emergency. The effective multiplication factor is shown to be within a standard value of 0.95 in any situation. Based on the experimental data on assembly and canister dose rates, canisters posing the highest radiation threat are selected. Activities of sources and gamma-radiation spectral composition are calculated. Distribution of the dose rate outside the containers both under the normal storage conditions and accident situations are calculated. The results obtained are analyzed

  11. Fuel Cycle of Reactor SVBR-100

    Energy Technology Data Exchange (ETDEWEB)

    Zrodnikov, A.V.; Toshinsky, G.I.; Komlev, O.G. [FSUE State Scientific Center Institute for Physics and Power Engineering, 1, Bondarenko sq., Obninsk, Kaluga rg., 249033 (Russian Federation)

    2009-06-15

    Modular fast reactor with lead-bismuth heavy liquid-metal coolant in 100 MWe class (SVBR 100) is referred to the IV Generation reactors and shall operate in a closed nuclear fuel cycle (NFC) without consumption of natural uranium. Usually it is considered that launch of fast reactors (FR) is realized using mixed uranium-plutonium fuel. However, such launch of FRs is not economically effective because of the current costs of natural uranium and uranium enrichment servicing. This is conditioned by the fact that the quantity of reprocessing the spent nuclear fuel (SNF) of thermal reactors (TR) calculated for a ton of plutonium that determines the expenditures for construction and operation of the corresponding enterprise is very large due to low content of plutonium in the TR SNF. The economical effectiveness of FRs will be reduced as the enterprises on reprocessing the TR SNF have to be built prior to FRs have been implemented in the nuclear power (NP). Moreover, the pace of putting the FRs in the NP will be constrained by the quantity of the TR SNF. The report grounds an alternative strategy of FRs implementation into the NP, which is considered to be more economically effective. That is conditioned by the fact that in the nearest future use of the mastered uranium oxide fuel for FRs and operation in the open fuel cycle with postponed reprocessing will be most economically expedient. Changeover to the mixed uranium-plutonium fuel and closed NFC will be economically effective when the cost of natural uranium is increased and the expenditures for construction of enterprises on SNF reprocessing, re-fabrication of new fuel with plutonium and their operating becomes lower than the corresponding costs of natural uranium, uranium enrichment servicing, expenditures for fabrication of fresh uranium fuel and long temporary storage of the SNF. As when operating in the open NFC, FRs use much more natural uranium as compared with TRs, and at a planned high pace of NP development

  12. Management and inspection of integrity of spent fuel from IRT MEPhI research reactor

    International Nuclear Information System (INIS)

    Aden, V.G.; Bulkin, S.Y.; Sokolov, A.V.; Bushuev, A.V.; Redkin, A.F.; Portnov, A.A.

    2002-01-01

    The information on wet storage and dry storage of the spent nuclear fuel (SNF) of the IRT MEPhI reactor and experience from SNF shipment for reprocessing are presented. The procedure and a facility for nondestructive inspection of local power density fields and the burnup of fuel assemblies based on studying the γ-activity of some fission products generated in U 235 and procedure for inspection of the fuel element cladding leak tightness are described. (author)

  13. Investigation of the warm prestress effect by X-ray and microfractographic measurements; Roentgenografische und mikrofraktografische Untersuchungen zum Warmvorbeanspruchungs (WPS) - Effekt

    Energy Technology Data Exchange (ETDEWEB)

    Blumenauer, H.; Eichler, B.; Krempe, M.; Ude, J. [Magdeburg Univ. (Germany). Inst. fuer Werkstofftechnik und Werkstoffpruefung

    1998-11-01

    The work reported was to investigate the changes caused at the crack tip of a specimen with incipient crack and to assess their effects with regard to component fracture. The steels selected for testing are the pressure vessel steels 10MnMoNi5-5 and 17MoV8-4, and the experiments were made with CT-25 specimens. The conclusion drawn from the results obtained is that the WPS effect is due to a stronger energy dissipation in the prestressed area ahead of the crack tip, assisted by intrinsic stress-induced crack closing. (orig./CB) [Deutsch] Das Ziel dieser Arbeit ist es, die durch eine Vorbeanspruchung angerissener Proben hervorgerufenen Veraenderungen an der Rissspitze und ihre Auswirkungen auf das Bruchverhalten zu untersuchen. Die Untersuchungen wurden an den Druckbehaelterstaehlen 10MnMoNi5-5 und 17MoV8-4 durchgefuehrt. Die Versuche wurden an CT-25-Proben durchgefuehrt. Die Autoren kommen zum Schluss, dass der WPS-Effekt mit einer staerkeren Energiedissipation in der vorgeschaedigten Prozesszone vor der Rissspitze, unterstuetzt durch eigenspannungsbedingtes Rissschliessen, erklaert werden kann. (orig./MM)

  14. Amino acid residues involved in ligand preference of the Snf3 transporter-like sensor in Saccharomyces cerevisiae

    DEFF Research Database (Denmark)

    Dietvorst, J.; Karhumaa, Kaisa; Kielland-Brandt, Morten

    2010-01-01

    /preferences of Snf3. The ability of cells to sense sugars in vivo was monitored by following the degradation of the Mth1 protein, :ill earl., event ill the signal pathway. Our study reveals that Snf3. ill addition to glucose. also senses fructose and mannose, as well as the glucose analogues 2-deoxyglucose, 3-O......-methylglucoside and 6-deoxyglucose. The signalling proficiency of a non-phosphorylatable analogue strongly supports the notion that sensing through Snf3 does not require sugar phosphorylation. Sequence comparisons of Snf3 to glucose transporters indicated amino acid residues possibly involved in sensing of sugars other...... than glucose. By site-specific mutagenesis of the structural gene, roles of specific residues in Snf3 could he established. Change of isoleucine-374 to valine ill transmembrane segment 7 of Snf3 partially abolished sensing of fructose mannose. while mutagenesis causing it change of phenylalanine-462 (4...

  15. Characterization Program Management Plan for Hanford K Basin Spent Nuclear Fuel (SNF) (OCRWM)

    International Nuclear Information System (INIS)

    BAKER, R.B.; TRIMBLE, D.J.

    2000-01-01

    The management plan developed to characterize the K Basin spent nuclear fuel (SNF) and sludge was originally developed for Westinghouse Hanford Company and Pacific Northwest National Laboratory to work together on a program to provide characterization data to support removal, conditioning, and subsequent dry storage of the SNF stored at the Hanford K Basins. The plan also addressed necessary characterization for the removal, transport, and storage of the sludge from the Hanford K Basins. This plan was revised in 1999 (i.e., Revision 2) to incorporate actions necessary to respond to the deficiencies revealed as the result of Quality Assurance surveillances and audits in 1999 with respect to the fuel characterization activities. Revision 3 to this Program Management Plan responds to a Worker Assessment resolution determined in Fical Year 2000. This revision includes an update to current organizational structures and other revisions needed to keep this management plan consistent with the current project scope. The plan continues to address both the SNF and the sludge accumulated at K Basins. Most activities for the characterization of the SNF have been completed. Data validation, Office of Civilian Radioactive Waste Management (OCRWM) document reviews, and OCRWM data qualification are the remaining SNF characterization activities. The transport and storage of K Basin sludge are affected by recent path forward revisions. These revisions require additional laboratory analyses of the sludge to complete the acquisition of required supporting engineering data. Hence, this revision of the management plan provides the overall work control for these remaining SNF and sludge characterization activities given the current organizational structure of the SNF Project

  16. Preparation for the Recovery of Spent Nuclear Fuel (SNF) at Andreeva Bay, North West Russia - 13309

    International Nuclear Information System (INIS)

    Field, D.; McAtamney, N.

    2013-01-01

    Andreeva Bay is located near Murmansk in the Russian Federation close to the Norwegian border. The ex-naval site was used to de-fuel nuclear-powered submarines and icebreakers during the Cold War. Approximately 22,000 fuel assemblies remain in three Dry Storage Units (DSUs) which means that Andreeva Bay has one of the largest stockpiles of highly enriched spent nuclear fuel (SNF) in the world. The high contamination and deteriorating condition of the SNF canisters has made improvements to the management of the SNF a high priority for the international community for safety, security and environmental reasons. International Donors have, since 2002, provided support to projects at Andreeva concerned with improving the management of the SNF. This long-term programme of work has been coordinated between the International Donors and responsible bodies within the Russian Federation. Options for the safe and secure management of SNF at Andreeva Bay were considered in 2004 and developed by a number of Russian Institutes with international participation. This consisted of site investigations, surveys and studies to understand the technical challenges. A principal agreement was reached that the SNF would be removed from the site altogether and transported to Russia's reprocessing facility at Mayak in the Urals. The analytical studies provided the information necessary to develop the construction plan for the site. Following design and regulatory processes, stakeholders endorsed the technical solution in April 2007. This detailed the processes, facilities and equipment required to safely remove the SNF and identified other site services and support facilities required on the site. Implementation of this strategy is now well underway with the facilities in various states of construction. Physical works have been performed to address the most urgent tasks including weather protection over one of the DSUs, installation of shielding over the cells, provision of radiation

  17. Preparation for the Recovery of Spent Nuclear Fuel (SNF) at Andreeva Bay, North West Russia - 13309

    Energy Technology Data Exchange (ETDEWEB)

    Field, D.; McAtamney, N. [Nuvia Limited (United Kingdom)

    2013-07-01

    Andreeva Bay is located near Murmansk in the Russian Federation close to the Norwegian border. The ex-naval site was used to de-fuel nuclear-powered submarines and icebreakers during the Cold War. Approximately 22,000 fuel assemblies remain in three Dry Storage Units (DSUs) which means that Andreeva Bay has one of the largest stockpiles of highly enriched spent nuclear fuel (SNF) in the world. The high contamination and deteriorating condition of the SNF canisters has made improvements to the management of the SNF a high priority for the international community for safety, security and environmental reasons. International Donors have, since 2002, provided support to projects at Andreeva concerned with improving the management of the SNF. This long-term programme of work has been coordinated between the International Donors and responsible bodies within the Russian Federation. Options for the safe and secure management of SNF at Andreeva Bay were considered in 2004 and developed by a number of Russian Institutes with international participation. This consisted of site investigations, surveys and studies to understand the technical challenges. A principal agreement was reached that the SNF would be removed from the site altogether and transported to Russia's reprocessing facility at Mayak in the Urals. The analytical studies provided the information necessary to develop the construction plan for the site. Following design and regulatory processes, stakeholders endorsed the technical solution in April 2007. This detailed the processes, facilities and equipment required to safely remove the SNF and identified other site services and support facilities required on the site. Implementation of this strategy is now well underway with the facilities in various states of construction. Physical works have been performed to address the most urgent tasks including weather protection over one of the DSUs, installation of shielding over the cells, provision of radiation

  18. Wongabel Rhabdovirus Accessory Protein U3 Targets the SWI/SNF Chromatin Remodeling Complex

    Science.gov (United States)

    Joubert, D. Albert; Rodriguez-Andres, Julio; Monaghan, Paul; Cummins, Michelle; McKinstry, William J.; Paradkar, Prasad N.; Moseley, Gregory W.

    2014-01-01

    ABSTRACT Wongabel virus (WONV) is an arthropod-borne rhabdovirus that infects birds. It is one of the growing array of rhabdoviruses with complex genomes that encode multiple accessory proteins of unknown function. In addition to the five canonical rhabdovirus structural protein genes (N, P, M, G, and L), the 13.2-kb negative-sense single-stranded RNA (ssRNA) WONV genome contains five uncharacterized accessory genes, one overlapping the N gene (Nx or U4), three located between the P and M genes (U1 to U3), and a fifth one overlapping the G gene (Gx or U5). Here we show that WONV U3 is expressed during infection in insect and mammalian cells and is required for efficient viral replication. A yeast two-hybrid screen against a mosquito cell cDNA library identified that WONV U3 interacts with the 83-amino-acid (aa) C-terminal domain of SNF5, a component of the SWI/SNF chromatin remodeling complex. The interaction was confirmed by affinity chromatography, and nuclear colocalization was established by confocal microscopy. Gene expression studies showed that SNF5 transcripts are upregulated during infection of mosquito cells with WONV, as well as West Nile virus (Flaviviridae) and bovine ephemeral fever virus (Rhabdoviridae), and that SNF5 knockdown results in increased WONV replication. WONV U3 also inhibits SNF5-regulated expression of the cytokine gene CSF1. The data suggest that WONV U3 targets the SWI/SNF complex to block the host response to infection. IMPORTANCE The rhabdoviruses comprise a large family of RNA viruses infecting plants, vertebrates, and invertebrates. In addition to the major structural proteins (N, P, M, G, and L), many rhabdoviruses encode a diverse array of accessory proteins of largely unknown function. Understanding the role of these proteins may reveal much about host-pathogen interactions in infected cells. Here we examine accessory protein U3 of Wongabel virus, an arthropod-borne rhabdovirus that infects birds. We show that U3 enters the

  19. Extracellular Matrix-Regulated Gene Expression RequiresCooperation of SWI/SNF and Transcription Factors

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Ren; Spencer, Virginia A.; Bissell, Mina J.

    2006-05-25

    Extracellular cues play crucial roles in the transcriptional regulation of tissue-specific genes, but whether and how these signals lead to chromatin remodeling is not understood and subject to debate. Using chromatin immunoprecipitation (ChIP) assays and mammary-specific genes as models, we show here that extracellular matrix (ECM) molecules and prolactin cooperate to induce histone acetylation and binding of transcription factors and the SWI/SNF complex to the {beta}- and ?-casein promoters. Introduction of a dominant negative Brg1, an ATPase subunit of SWI/SNF complex, significantly reduced both {beta}- and ?-casein expression, suggesting that SWI/SNF-dependent chromatin remodeling is required for transcription of mammary-specific genes. ChIP analyses demonstrated that the ATPase activity of SWI/SNF is necessary for recruitment of RNA transcriptional machinery, but not for binding of transcription factors or for histone acetylation. Coimmunoprecipitation analyses showed that the SWI/SNF complex is associated with STAT5, C/EBP{beta}, and glucocorticoid receptor (GR). Thus, ECM- and prolactin-regulated transcription of the mammary-specific casein genes requires the concerted action of chromatin remodeling enzymes and transcription factors.

  20. SNF5 is an essential executor of epigenetic regulation during differentiation.

    Science.gov (United States)

    You, Jueng Soo; De Carvalho, Daniel D; Dai, Chao; Liu, Minmin; Pandiyan, Kurinji; Zhou, Xianghong J; Liang, Gangning; Jones, Peter A

    2013-04-01

    Nucleosome occupancy controls the accessibility of the transcription machinery to DNA regulatory regions and serves an instructive role for gene expression. Chromatin remodelers, such as the BAF complexes, are responsible for establishing nucleosome occupancy patterns, which are key to epigenetic regulation along with DNA methylation and histone modifications. Some reports have assessed the roles of the BAF complex subunits and stemness in murine embryonic stem cells. However, the details of the relationships between remodelers and transcription factors in altering chromatin configuration, which ultimately affects gene expression during cell differentiation, remain unclear. Here for the first time we demonstrate that SNF5, a core subunit of the BAF complex, negatively regulates OCT4 levels in pluripotent cells and is essential for cell survival during differentiation. SNF5 is responsible for generating nucleosome-depleted regions (NDRs) at the regulatory sites of OCT4 repressed target genes such as PAX6 and NEUROG1, which are crucial for cell fate determination. Concurrently, SNF5 closes the NDRs at the regulatory regions of OCT4-activated target genes such as OCT4 itself and NANOG. Furthermore, using loss- and gain-of-function experiments followed by extensive genome-wide analyses including gene expression microarrays and ChIP-sequencing, we highlight that SNF5 plays dual roles during differentiation by antagonizing the expression of genes that were either activated or repressed by OCT4, respectively. Together, we demonstrate that SNF5 executes the switch between pluripotency and differentiation.

  1. SNF5 is an essential executor of epigenetic regulation during differentiation.

    Directory of Open Access Journals (Sweden)

    Jueng Soo You

    2013-04-01

    Full Text Available Nucleosome occupancy controls the accessibility of the transcription machinery to DNA regulatory regions and serves an instructive role for gene expression. Chromatin remodelers, such as the BAF complexes, are responsible for establishing nucleosome occupancy patterns, which are key to epigenetic regulation along with DNA methylation and histone modifications. Some reports have assessed the roles of the BAF complex subunits and stemness in murine embryonic stem cells. However, the details of the relationships between remodelers and transcription factors in altering chromatin configuration, which ultimately affects gene expression during cell differentiation, remain unclear. Here for the first time we demonstrate that SNF5, a core subunit of the BAF complex, negatively regulates OCT4 levels in pluripotent cells and is essential for cell survival during differentiation. SNF5 is responsible for generating nucleosome-depleted regions (NDRs at the regulatory sites of OCT4 repressed target genes such as PAX6 and NEUROG1, which are crucial for cell fate determination. Concurrently, SNF5 closes the NDRs at the regulatory regions of OCT4-activated target genes such as OCT4 itself and NANOG. Furthermore, using loss- and gain-of-function experiments followed by extensive genome-wide analyses including gene expression microarrays and ChIP-sequencing, we highlight that SNF5 plays dual roles during differentiation by antagonizing the expression of genes that were either activated or repressed by OCT4, respectively. Together, we demonstrate that SNF5 executes the switch between pluripotency and differentiation.

  2. Identification of multiple distinct Snf2 subfamilies with conserved structural motifs.

    Science.gov (United States)

    Flaus, Andrew; Martin, David M A; Barton, Geoffrey J; Owen-Hughes, Tom

    2006-01-01

    The Snf2 family of helicase-related proteins includes the catalytic subunits of ATP-dependent chromatin remodelling complexes found in all eukaryotes. These act to regulate the structure and dynamic properties of chromatin and so influence a broad range of nuclear processes. We have exploited progress in genome sequencing to assemble a comprehensive catalogue of over 1300 Snf2 family members. Multiple sequence alignment of the helicase-related regions enables 24 distinct subfamilies to be identified, a considerable expansion over earlier surveys. Where information is known, there is a good correlation between biological or biochemical function and these assignments, suggesting Snf2 family motor domains are tuned for specific tasks. Scanning of complete genomes reveals all eukaryotes contain members of multiple subfamilies, whereas they are less common and not ubiquitous in eubacteria or archaea. The large sample of Snf2 proteins enables additional distinguishing conserved sequence blocks within the helicase-like motor to be identified. The establishment of a phylogeny for Snf2 proteins provides an opportunity to make informed assignments of function, and the identification of conserved motifs provides a framework for understanding the mechanisms by which these proteins function.

  3. Mechanisms of regulation of SNF1/AMPK/SnRK1 protein kinases

    Science.gov (United States)

    Crozet, Pierre; Margalha, Leonor; Confraria, Ana; Rodrigues, Américo; Martinho, Cláudia; Adamo, Mattia; Elias, Carlos A.; Baena-González, Elena

    2014-01-01

    The SNF1 (sucrose non-fermenting 1)-related protein kinases 1 (SnRKs1) are the plant orthologs of the budding yeast SNF1 and mammalian AMPK (AMP-activated protein kinase). These evolutionarily conserved kinases are metabolic sensors that undergo activation in response to declining energy levels. Upon activation, SNF1/AMPK/SnRK1 kinases trigger a vast transcriptional and metabolic reprograming that restores energy homeostasis and promotes tolerance to adverse conditions, partly through an induction of catabolic processes and a general repression of anabolism. These kinases typically function as a heterotrimeric complex composed of two regulatory subunits, β and γ, and an α-catalytic subunit, which requires phosphorylation of a conserved activation loop residue for activity. Additionally, SNF1/AMPK/SnRK1 kinases are controlled by multiple mechanisms that have an impact on kinase activity, stability, and/or subcellular localization. Here we will review current knowledge on the regulation of SNF1/AMPK/SnRK1 by upstream components, post-translational modifications, various metabolites, hormones, and others, in an attempt to highlight both the commonalities of these essential eukaryotic kinases and the divergences that have evolved to cope with the particularities of each one of these systems. PMID:24904600

  4. White Paper: Multi-purpose canister (MPC) for DOE-owned spent nuclear fuel (SNF)

    International Nuclear Information System (INIS)

    Knecht, D.A.

    1994-04-01

    The paper examines the issue, What are the advantages, disadvantages, and other considerations for using the MPC concept as part of the strategy for interim storage and disposal of DOE-owned SNF? The paper is based in part on the results of an evaluation made for the DOE National Spent Fuel Program by the Waste Form Barrier/Canister Team, which is composed of knowledgeable DOE and DOE-contractor personnel. The paper reviews the MPC and DOE SNF status, provides criteria and other considerations applicable to the issue, and presents an evaluation, conclusions, and recommendations. The primary conclusion is that while most of DOE SNF is not currently sufficiently characterized to be sealed into an MPC, the advantages of standardized packages in handling, reduced radiation exposure, and improved human factors should be considered in DOE SNF program planning. While the design of MPCs for DOE SNF are likely premature at this time, the use of canisters should be considered which are consistent with interim storage options and the MPC design envelope

  5. DAF-16 employs the chromatin remodeller SWI/SNF to promote stress resistance and longevity.

    Science.gov (United States)

    Riedel, Christian G; Dowen, Robert H; Lourenco, Guinevere F; Kirienko, Natalia V; Heimbucher, Thomas; West, Jason A; Bowman, Sarah K; Kingston, Robert E; Dillin, Andrew; Asara, John M; Ruvkun, Gary

    2013-05-01

    Organisms are constantly challenged by stresses and privations and require adaptive responses for their survival. The forkhead box O (FOXO) transcription factor DAF-16 (hereafter referred to as DAF-16/FOXO) is a central nexus in these responses, but despite its importance little is known about how it regulates its target genes. Proteomic identification of DAF-16/FOXO-binding partners in Caenorhabditis elegans and their subsequent functional evaluation by RNA interference revealed several candidate DAF-16/FOXO cofactors, most notably the chromatin remodeller SWI/SNF. DAF-16/FOXO and SWI/SNF form a complex and globally co-localize at DAF-16/FOXO target promoters. We show that specifically for gene activation, DAF-16/FOXO depends on SWI/SNF, facilitating SWI/SNF recruitment to target promoters, to activate transcription by presumed remodelling of local chromatin. For the animal, this translates into an essential role for SWI/SNF in DAF-16/FOXO-mediated processes, in particular dauer formation, stress resistance and the promotion of longevity. Thus, we give insight into the mechanisms of DAF-16/FOXO-mediated transcriptional regulation and establish a critical link between ATP-dependent chromatin remodelling and lifespan regulation.

  6. DAF-16/FOXO employs the chromatin remodeller SWI/SNF to promote stress resistance and longevity

    Science.gov (United States)

    Riedel, Christian G.; Dowen, Robert H.; Lourenco, Guinevere F.; Kirienko, Natalia V.; Heimbucher, Thomas; West, Jason A.; Bowman, Sarah K.; Kingston, Robert E.; Dillin, Andrew; Asara, John M.; Ruvkun, Gary

    2013-01-01

    Organisms are constantly challenged by stresses and privations and require adaptive responses for their survival. The transcription factor DAF-16/FOXO is central nexus in these responses, but despite its importance little is known about how it regulates its target genes. Proteomic identification of DAF-16/FOXO binding partners in Caenorhabditis elegans and their subsequent functional evaluation by RNA interference (RNAi) revealed several candidate DAF-16/FOXO cofactors, most notably the chromatin remodeller SWI/SNF. DAF-16/FOXO and SWI/SNF form a complex and globally colocalize at DAF-16/FOXO target promoters. We show that specifically for gene-activation, DAF-16/FOXO depends on SWI/SNF, facilitating SWI/SNF recruitment to target promoters, in order to activate transcription by presumed remodelling of local chromatin. For the animal, this translates into an essential role of SWI/SNF for DAF-16/FOXO-mediated processes, i.e. dauer formation, stress resistance, and the promotion of longevity. Thus we give insight into the mechanisms of DAF-16/FOXO-mediated transcriptional regulation and establish a critical link between ATP-dependent chromatin remodelling and lifespan regulation. PMID:23604319

  7. Conditions With High Intracellular Glucose Inhibit Sensing Through Glucose Sensor Snf3 in Saccharomyces cerevisiae

    DEFF Research Database (Denmark)

    Karhumaa, Kaisa; Wu, B.Q.; Kielland-Brandt, Morten

    2010-01-01

    as for amino acids. An alternating-access model of the function of transporter-like sensors has been previously suggested based on amino acid sensing, where intracellular ligand inhibits binding of extracellular ligand. Here we studied the effect of intracellular glucose on sensing of extracellular glucose...... through the transporter-like sensor Snf3 in yeast. Sensing through Snf3 was determined by measuring degradation of Mth1 protein. High intracellular glucose concentrations were achieved by using yeast strains lacking monohexose transporters which were grown on maltose. The apparent affinity...... of extracellular glucose to Snf3 was measured for cells grown in non-fermentative medium or on maltose. The apparent affinity for glucose was lowest when the intracellular glucose concentration was high. The results conform to an alternating-access model for transporter-like sensors. J. Cell. Biochem. 110: 920...

  8. Functions and requirements for K Basin SNF characterization shipping

    International Nuclear Information System (INIS)

    Bergmann, D.W.

    1994-01-01

    This document details the plan for the shipping of fuel samples from the K Basins to the 300 Area for characterization. The fuel characterization project will evaluate the Hanford defense production fuel (N-Reactor and Single Pass Reactor) to support interim storage, transportation and final disposition. A limited number of fuel samples will be transported to a laboratory for analysis. It is currently estimated that 20 shipments of fuel per year for approximately 3 years (could be as long as 5 years) will be transported to the laboratory for analysis. Based on the NRC certificate of compliance each shipment is limited to 500 equivalent grams of 235 U. In practical terms this will limit shipments to three outer elements or two assemblies of any type of N-Reactor or SPR fuel. Case by case determination of broken fuel will be made based on the type of fuel and maximum potential fissile content

  9. Regulatory Experiences from Effective Step-wise Implementation of the SNF Disposal in Finland

    International Nuclear Information System (INIS)

    Hämäläinen, K.

    2016-01-01

    Finland is one of the foremost countries in the world in developing a disposal solution for spent nuclear fuel (SNF). The Construction License Application (CLA) for the Olkiluoto SNF encapsulation and disposal facility was submitted by Posiva, the implementer, to the authorities at the end of 2012 and the Government is expected to decide about the license during autumn 2015. In 1983 the Government made a strategy decision on the objectives and target time schedule for the research, development and technical planning of nuclear waste management. Decision included the milestones for site selection, submittal of construction license and start of disposal operations.

  10. Mutations affecting components of the SWI/SNF complex cause Coffin-Siris syndrome.

    Science.gov (United States)

    Tsurusaki, Yoshinori; Okamoto, Nobuhiko; Ohashi, Hirofumi; Kosho, Tomoki; Imai, Yoko; Hibi-Ko, Yumiko; Kaname, Tadashi; Naritomi, Kenji; Kawame, Hiroshi; Wakui, Keiko; Fukushima, Yoshimitsu; Homma, Tomomi; Kato, Mitsuhiro; Hiraki, Yoko; Yamagata, Takanori; Yano, Shoji; Mizuno, Seiji; Sakazume, Satoru; Ishii, Takuma; Nagai, Toshiro; Shiina, Masaaki; Ogata, Kazuhiro; Ohta, Tohru; Niikawa, Norio; Miyatake, Satoko; Okada, Ippei; Mizuguchi, Takeshi; Doi, Hiroshi; Saitsu, Hirotomo; Miyake, Noriko; Matsumoto, Naomichi

    2012-03-18

    By exome sequencing, we found de novo SMARCB1 mutations in two of five individuals with typical Coffin-Siris syndrome (CSS), a rare autosomal dominant anomaly syndrome. As SMARCB1 encodes a subunit of the SWItch/Sucrose NonFermenting (SWI/SNF) complex, we screened 15 other genes encoding subunits of this complex in 23 individuals with CSS. Twenty affected individuals (87%) each had a germline mutation in one of six SWI/SNF subunit genes, including SMARCB1, SMARCA4, SMARCA2, SMARCE1, ARID1A and ARID1B.

  11. Snf2 family gene distribution in higher plant genomes reveals DRD1 expansion and diversification in the tomato genome.

    Science.gov (United States)

    Bargsten, Joachim W; Folta, Adam; Mlynárová, Ludmila; Nap, Jan-Peter

    2013-01-01

    As part of large protein complexes, Snf2 family ATPases are responsible for energy supply during chromatin remodeling, but the precise mechanism of action of many of these proteins is largely unknown. They influence many processes in plants, such as the response to environmental stress. This analysis is the first comprehensive study of Snf2 family ATPases in plants. We here present a comparative analysis of 1159 candidate plant Snf2 genes in 33 complete and annotated plant genomes, including two green algae. The number of Snf2 ATPases shows considerable variation across plant genomes (17-63 genes). The DRD1, Rad5/16 and Snf2 subfamily members occur most often. Detailed analysis of the plant-specific DRD1 subfamily in related plant genomes shows the occurrence of a complex series of evolutionary events. Notably tomato carries unexpected gene expansions of DRD1 gene members. Most of these genes are expressed in tomato, although at low levels and with distinct tissue or organ specificity. In contrast, the Snf2 subfamily genes tend to be expressed constitutively in tomato. The results underpin and extend the Snf2 subfamily classification, which could help to determine the various functional roles of Snf2 ATPases and to target environmental stress tolerance and yield in future breeding.

  12. Reconstruction of the yeast Snf1 kinase regulatory network reveals its role as a global energy regulator

    DEFF Research Database (Denmark)

    Usaite, Renata; Jewett, Michael Christopher; Soberano de Oliveira, Ana Paula

    2009-01-01

    Highly conserved among eukaryotic cells, the AMP-activated kinase (AMPK) is a central regulator of carbon metabolism. To map the complete network of interactions around AMPK in yeast (Snf1) and to evaluate the role of its regulatory subunit Snf4, we measured global mRNA, protein and metabolite...

  13. Snf2 family gene distribution in higher plant genomes reveals DRD1 expansion and diversification in the tomato genome.

    Directory of Open Access Journals (Sweden)

    Joachim W Bargsten

    Full Text Available As part of large protein complexes, Snf2 family ATPases are responsible for energy supply during chromatin remodeling, but the precise mechanism of action of many of these proteins is largely unknown. They influence many processes in plants, such as the response to environmental stress. This analysis is the first comprehensive study of Snf2 family ATPases in plants. We here present a comparative analysis of 1159 candidate plant Snf2 genes in 33 complete and annotated plant genomes, including two green algae. The number of Snf2 ATPases shows considerable variation across plant genomes (17-63 genes. The DRD1, Rad5/16 and Snf2 subfamily members occur most often. Detailed analysis of the plant-specific DRD1 subfamily in related plant genomes shows the occurrence of a complex series of evolutionary events. Notably tomato carries unexpected gene expansions of DRD1 gene members. Most of these genes are expressed in tomato, although at low levels and with distinct tissue or organ specificity. In contrast, the Snf2 subfamily genes tend to be expressed constitutively in tomato. The results underpin and extend the Snf2 subfamily classification, which could help to determine the various functional roles of Snf2 ATPases and to target environmental stress tolerance and yield in future breeding.

  14. Differential Roles of the Glycogen-Binding Domains of β Subunits in Regulation of the Snf1 Kinase Complex▿

    Science.gov (United States)

    Mangat, Simmanjeet; Chandrashekarappa, Dakshayini; McCartney, Rhonda R.; Elbing, Karin; Schmidt, Martin C.

    2010-01-01

    Members of the AMP-activated protein kinase family, including the Snf1 kinase of Saccharomyces cerevisiae, are activated under conditions of nutrient stress. AMP-activated protein kinases are heterotrimeric complexes composed of a catalytic α subunit and regulatory β and γ subunits. In this study, the role of the β subunits in the regulation of Snf1 activity was examined. Yeasts express three isoforms of the AMP-activated protein kinase consisting of Snf1 (α), Snf4 (γ), and one of three alternative β subunits, either Sip1, Sip2, or Gal83. The Gal83 isoform of the Snf1 complex is the most abundant and was analyzed in the greatest detail. All three β subunits contain a conserved domain referred to as the glycogen-binding domain. The deletion of this domain from Gal83 results in a deregulation of the Snf1 kinase, as judged by a constitutive activity independent of glucose availability. In contrast, the deletion of this homologous domain from the Sip1 and Sip2 subunits had little effect on Snf1 kinase regulation. Therefore, the different Snf1 kinase isoforms are regulated through distinct mechanisms, which may contribute to their specialized roles in different stress response pathways. In addition, the β subunits are subjected to phosphorylation. The responsible kinases were identified as being Snf1 and casein kinase II. The significance of the phosphorylation is unclear since the deletion of the region containing the phosphorylation sites in Gal83 had little effect on the regulation of Snf1 in response to glucose limitation. PMID:19897735

  15. Differential roles of the glycogen-binding domains of beta subunits in regulation of the Snf1 kinase complex.

    Science.gov (United States)

    Mangat, Simmanjeet; Chandrashekarappa, Dakshayini; McCartney, Rhonda R; Elbing, Karin; Schmidt, Martin C

    2010-01-01

    Members of the AMP-activated protein kinase family, including the Snf1 kinase of Saccharomyces cerevisiae, are activated under conditions of nutrient stress. AMP-activated protein kinases are heterotrimeric complexes composed of a catalytic alpha subunit and regulatory beta and gamma subunits. In this study, the role of the beta subunits in the regulation of Snf1 activity was examined. Yeasts express three isoforms of the AMP-activated protein kinase consisting of Snf1 (alpha), Snf4 (gamma), and one of three alternative beta subunits, either Sip1, Sip2, or Gal83. The Gal83 isoform of the Snf1 complex is the most abundant and was analyzed in the greatest detail. All three beta subunits contain a conserved domain referred to as the glycogen-binding domain. The deletion of this domain from Gal83 results in a deregulation of the Snf1 kinase, as judged by a constitutive activity independent of glucose availability. In contrast, the deletion of this homologous domain from the Sip1 and Sip2 subunits had little effect on Snf1 kinase regulation. Therefore, the different Snf1 kinase isoforms are regulated through distinct mechanisms, which may contribute to their specialized roles in different stress response pathways. In addition, the beta subunits are subjected to phosphorylation. The responsible kinases were identified as being Snf1 and casein kinase II. The significance of the phosphorylation is unclear since the deletion of the region containing the phosphorylation sites in Gal83 had little effect on the regulation of Snf1 in response to glucose limitation.

  16. Criticality Potential of Waste Packages Containing DOE SNF Affected by Igneous Intrusion

    International Nuclear Information System (INIS)

    D.S. Kimball; C.E. Sanders

    2006-01-01

    The Department of Energy (DOE) is currently preparing an application to submit to the U.S. Nuclear Regulatory Commission for a construction authorization for a monitored geologic repository. The repository will contain spent nuclear fuel (SNF) and defense high-level waste (DHLW) in waste packages placed in underground tunnels, or drifts. The primary objective of this paper is to perform a criticality analysis for waste packages containing DOE SNF affected by a disruptive igneous intrusion event in the emplacement drifts. The waste packages feature one DOE SNF canister placed in the center and surrounded by five High-Level Waste (HLW) glass canisters. The effective neutron multiplication factor (k eff ) is determined for potential configurations of the waste package during and after an intrusive igneous event. Due to the complexity of the potential scenarios following an igneous intrusion, finding conservative and bounding configurations with respect to criticality requires some additional considerations. In particular, the geometry of a slumped and damaged waste package must be examined, drift conditions must be modeled over a range of parameters, and the chemical degradation of DOE SNF and waste package materials must be considered for the expected high temperatures. The secondary intent of this calculation is to present a method for selecting conservative and bounding configurations for a wide range of end conditions

  17. Evaluation of Radionuclide Release from Aluminum-Based SNF in Basin Storage

    International Nuclear Information System (INIS)

    Sindelar, R.L.; Burke, S.D.; Howell, J.P.

    1998-09-01

    This report provides an evaluation of the release rates of radionuclides from breached A1-SNF assemblies and evaluates the effect of direct storage of breached fuel at a conservative upper bound reference condition on the SRS basin water activity levels

  18. Duo_2-Steel cermet manufacturing technology for PWR Spent Nuclear Fuel (SNF) casks

    International Nuclear Information System (INIS)

    Siti Alimah; Budiarto

    2005-01-01

    Assessment of DUO_2-Steel cermet manufacturing technology for PWR SNF casks has been done. DUO_2-Steel cermet consisting of DUO_2 particulates and other particulates, embedded in a steel matrix. Cermet SNF casks have the potential for superior performance compared with casks constructed of other materials. The addition of DUO_2 ceramic particulates can increase SNF cask capacity, improve of repository performance and disposal of excess depleted uranium as potential waste. Two sets of cermet manufacturing technologies are casting and powder metallurgy. Three casting methods are infusion casting, traditional casting and centrifugal casting. While for powder metallurgy methods there are traditional method and new method. DUO_2-Steel cermet have traditionally been produced by powder metallurgy methods. The production of a cask, however, presents special requirements: the manufacture of an annular object with weights up to 100 tons, and methods are being not to manufacture a cermet of this size and geometry. A new powder metallurgy method, is a method for manufacturing cermet for PWR SNF cask. This powder metallurgy techniques have potentials low costs and provides greater freedom In the design of the cermet cask by allowing variable cermet properties. (author)

  19. Storage of non-defense production reactor spent nuclear fuel at the Department of Energy's Hanford Site

    International Nuclear Information System (INIS)

    Carlson, A.B.

    1998-01-01

    In 1992, the US Department of Energy (DOE) established a program at the Hanford Site for management of DOE-owned spent nuclear fuel (SNF) until final disposition. Currently, the DOE-owned SNF Program is developing and implementing plans to assure existing storage, achieve interim storage, and prepare DOE-owned SNF for final disposition. Program requirements for management of the SNF are delineated in the DOE-owned SNF Program Plan.(DOE 1995a) and the DOE Spent Fuel Program's Requirements Document (DOE 1994a). Major program requirements are driven by the following: commitments established in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 94-1 Implementation Plan (DOE 1995b); corrective action plans for resolving vulnerabilities identified in the DOE Spent Fuel Working Group's Report on Health, Safety, and Environmental Vulnerabilities for Reactor Irradiated Nuclear Materials (DOE 1993); the settlement agreement between the US Department of Navy, the US Department of Energy, and the State of Idaho on the record of decision (ROD) from the DOE Programmatic SNF Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Environmental Impact Statement (DOE Programmatic SNF EIS) (Idaho, 1995)

  20. Marine reactor pressure vessels dumped in the Kara Sea

    International Nuclear Information System (INIS)

    Mount, M.E.

    1997-01-01

    Between 1965 and 1988, 16 marine reactors from seven Russian submarines and the icebreaker Lenin, each of which suffered some form of reactor accident, were dumped in a variety of containments, using a number of sealing methods, at five sites in the Kara Sea. All reactors were dumped at sites that varied in depth from 12 to 300 m and six contained their spent nuclear fuel (SNF). This paper examines the breakdown of the reactor pressure vessel (RPV) barriers due to corrosion, with specific emphasis on those RPVs containing SNF. Included are discussions of the structural aspects of the steam generating installations and their associated RPVs, a summary of the disposal operations, assumptions on corrosion rates of structural and filler materials, and an estimate of the structural integrity of the RPVs at the present time (1996) and in the year 2015

  1. Human Error Prediction and Countermeasures based on CREAM in Loading and Storage Phase of Spent Nuclear Fuel (SNF)

    International Nuclear Information System (INIS)

    Kim, Jae San; Kim, Min Su; Jo, Seong Youn

    2007-01-01

    With the steady demands for nuclear power energy in Korea, the amount of accumulated SNF has inevitably increased year by year. Thus far, SNF has been on-site transported from one unit to a nearby unit or an on-site dry storage facility. In the near future, as the amount of SNF generated approaches the capacity of these facilities, a percentage of it will be transported to another SNF storage facility. In the process of transporting SNF, human interactions involve inspecting and preparing the cask and spent fuel, loading the cask, transferring the cask and storage or monitoring the cask, etc. So, human actions play a significant role in SNF transportation. In analyzing incidents that have occurred during transport operations, several recent studies have indicated that 'human error' is a primary cause. Therefore, the objectives of this study are to predict and identify possible human errors during the loading and storage of SNF. Furthermore, after evaluating human error for each process, countermeasures to minimize human error are deduced

  2. Snf1 Phosphorylates Adenylate Cyclase and Negatively Regulates Protein Kinase A-dependent Transcription in Saccharomyces cerevisiae.

    Science.gov (United States)

    Nicastro, Raffaele; Tripodi, Farida; Gaggini, Marco; Castoldi, Andrea; Reghellin, Veronica; Nonnis, Simona; Tedeschi, Gabriella; Coccetti, Paola

    2015-10-09

    In eukaryotes, nutrient availability and metabolism are coordinated by sensing mechanisms and signaling pathways, which influence a broad set of cellular functions such as transcription and metabolic pathways to match environmental conditions. In yeast, PKA is activated in the presence of high glucose concentrations, favoring fast nutrient utilization, shutting down stress responses, and boosting growth. On the contrary, Snf1/AMPK is activated in the presence of low glucose or alternative carbon sources, thus promoting an energy saving program through transcriptional activation and phosphorylation of metabolic enzymes. The PKA and Snf1/AMPK pathways share common downstream targets. Moreover, PKA has been reported to negatively influence the activation of Snf1/AMPK. We report a new cross-talk mechanism with a Snf1-dependent regulation of the PKA pathway. We show that Snf1 and adenylate cyclase (Cyr1) interact in a nutrient-independent manner. Moreover, we identify Cyr1 as a Snf1 substrate and show that Snf1 activation state influences Cyr1 phosphorylation pattern, cAMP intracellular levels, and PKA-dependent transcription. © 2015 by The American Society for Biochemistry and Molecular Biology, Inc.

  3. A method for selection of spent nuclear fuel (SNF) transportation route considering socioeconomic cost based on contingent valuation method (CVM)

    International Nuclear Information System (INIS)

    Kim, Young Sik

    2008-02-01

    A transportation of SNF may cause an additional radiation exposure to human beings. It means that the radiological risk should be estimated and managed quantitatively for the public who live near the shipments route. Before the SNF transportation is performed, the route selection is concluded based on the radiological risk estimated with RADTRAN code in existing method generally. It means the existing method for route selection is based only on the radiological health risk but there are not only the impacts related to the radiological health risk but also the socioeconomic impacts related to the cost. In this study, a new method and its numerical formula for route selection on transporting SNF is proposed based on cost estimation because there are several costs in transporting SNF. The total cost consists of radiological health cost, transportation cost, and socioeconomic cost. Each cost is defined properly to the characteristics of SNF transportation and many coefficients and variables describing the meaning of each cost are obtained or estimated through many surveys. Especially to get the socioeconomic cost, contingent valuation method (CVM) is used with a questionnaire. The socioeconomic cost estimation is the most important part of the total cost originated from transporting SNF because it is a very dominant cost in the total cost. The route selection regarding SNF transportation can be supported with the proposed method reasonably and unnecessary or exhausting controversies about the shipments could be avoided

  4. Available reprocessing and recycling services for research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tozser, Sandor Miklos; Adelfang, Pablo; Bradley, Ed [International Atomic Energy Agency, Vienna (Austria); Budu, Madalina [SOSNY Research and Development Company, Moscow (Russian Federation); Chiguer, Mustapha [AREVA, Paris (France)

    2015-05-15

    International activities in the back-end of the research reactor (RR) fuel cycle have so far been dominated by the programmes of acceptance of highly-enriched uranium (HEU) spent nuclear fuel (SNF) by the country where it was originally enriched. These programmes will soon have achieved their goals and the SNF take-back programmes will cease. However, the needs of the nuclear community dictate that the majority of the research reactors continue to operate using low enriched uranium (LEU) fuel in order to meet the varied mission objectives. As a result, inventories of LEU SNF will continue to be created and the back-end solution of RR SNF remains a critical issue. In view of this fact, the IAEA, based on the experience gained during the decade of international cooperation in supporting the objectives of the HEU take-back programmes, will draw up a report presenting available reprocessing and recycling services for research reactor spent nuclear fuel. This paper gives an overview of the guiding document which will address all aspects of Reprocessing and Recycling Services for RR SNF, including an overview of solutions, decision making support, service suppliers, conditions (prerequisites, options, etc.), services offered by the managerial and logistics support providers with a focus on available transport packages and applicable transport modes.

  5. Glucose de-repression by yeast AMP-activated protein kinase SNF1 is controlled via at least two independent steps.

    Science.gov (United States)

    García-Salcedo, Raúl; Lubitz, Timo; Beltran, Gemma; Elbing, Karin; Tian, Ye; Frey, Simone; Wolkenhauer, Olaf; Krantz, Marcus; Klipp, Edda; Hohmann, Stefan

    2014-04-01

    The AMP-activated protein kinase, AMPK, controls energy homeostasis in eukaryotic cells but little is known about the mechanisms governing the dynamics of its activation/deactivation. The yeast AMPK, SNF1, is activated in response to glucose depletion and mediates glucose de-repression by inactivating the transcriptional repressor Mig1. Here we show that overexpression of the Snf1-activating kinase Sak1 results, in the presence of glucose, in constitutive Snf1 activation without alleviating glucose repression. Co-overexpression of the regulatory subunit Reg1 of the Glc-Reg1 phosphatase complex partly restores glucose regulation of Snf1. We generated a set of 24 kinetic mathematical models based on dynamic data of Snf1 pathway activation and deactivation. The models that reproduced our experimental observations best featured (a) glucose regulation of both Snf1 phosphorylation and dephosphorylation, (b) determination of the Mig1 phosphorylation status in the absence of glucose by Snf1 activity only and (c) a regulatory step directing active Snf1 to Mig1 under glucose limitation. Hence it appears that glucose de-repression via Snf1-Mig1 is regulated by glucose via at least two independent steps: the control of activation of the Snf1 kinase and directing active Snf1 to inactivating its target Mig1. © 2014 FEBS.

  6. Safety assessment for a potential SNF repository and its implication to the proliferation resistance nuclear fuel cycle

    International Nuclear Information System (INIS)

    Hwang, Y.; Jeong, M.S.; Seo, C.S.

    2007-01-01

    KAERI is developing the pyro-process technology to minimize the burden on permanent disposal of spent nuclear fuel. In addition, KAERI has developed the Korean Reference System for potential spent nuclear fuel disposal since 1997. The deep geologic disposal system is composed of a multi-barrier system in a crystalline rock to dispose of 36,000 MT of spent nuclear fuel (SNF) from a CANDU and a PWR. Quite recently, introduction of advanced nuclear fuel cycles such as pyro-processing is a big issue to solve the everlasting disposal problem and to assure the sustainable supply of fuel for reactors. To compare the effect of direct disposal of SNF with that of the high level waste disposal for waste generated from the advanced nuclear fuel cycles, the total system performance assessment for two different schemes is developed; one for direct disposal of SNF and the other for the introduction of the pyro-processing and direct disposal CANDU spent nuclear fuel. The safety indicators to assess the environmental friendliness of the disposal option are annual individual doses, toxicities and risks. Even though many scientists use the toxicity to understand the environmental friendliness of the disposal, scientifically the annual individual doses or risks are meaningful indicators for it. The major mechanisms to determine the doses and risks for direct disposal are as follows: (1) Dissolution mechanisms of uranium dioxides which control the dissolution of most nuclides such as TRU's and most parts of fission products. (2) Instant release fraction of highly soluble nuclides such as I-129, C-135, Tc-99, and others. (3) Retardation and dilution effect of natural and engineered barriers. (4) Dilution effect in the biosphere. The dominant nuclide is I-129 which follows both congruent and instantaneous release modes. Since its long half life associated with the instantaneous release I-129 is dominant well beyond one million. The impact of the TRU's is negligible until the significant

  7. Thermal properties and phase transition in the fluoride, (NH4)3SnF7

    International Nuclear Information System (INIS)

    Kartashev, A.V.; Gorev, M.V.; Bogdanov, E.V.; Flerov, I.N.; Laptash, N.M.

    2016-01-01

    Calorimetric, dilatometric and differential thermal analysis studies were performed on (NH 4 ) 3 SnF 7 for a wide range of temperatures and pressures. Large entropy (δS 0 =22 J/mol K) and elastic deformation (δ(ΔV/V) 0 =0.89%) jumps have proven that the Pa-3↔Pm-3m phase transition is a strong first order structural transformation. A total entropy change of ΔS 0 =32.5 J/mol K is characteristic for the order–disorder phase transition, and is equal to the sum of entropy changes in the related material, (NH 4 ) 3 TiF 7 , undergoing transformation between the two cubic phases through the intermediate phases. Hydrostatic pressure decreases the stability of the high temperature Pm-3m phase in (NH 4 ) 3 SnF 7 , contrary to (NH 4 ) 3 TiF 7 , characterised by a negative baric coefficient. The effect of experimental conditions on the chemical stability of (NH 4 ) 3 SnF 7 was observed. - Graphical abstract: Strong first order structural transformation Pa-3↔Pm-3m in (NH 4 ) 3 SnF 7 is associated with very large total entropy change of ΔS 0 =32.5 J/mol K characteristic for the ordering processes and equal to the sum of entropy changes in the related (NH 4 ) 3 TiF 7 undergoing transformation between the same two cubic phases through the intermediate phases. - Highlights: • (NH 4 ) 3 SnF 7 undergoes strong first order Pa-3↔Pm-3m phase transition. • Anomalous behaviour of ΔC p and ΔV/V exists far below phase transition temperature. • Structural distortions are accompanied by huge total entropy change ΔS≈Rln50. • High pressure strongly increases the stability of Pa-3 phase in (NH 4 ) 3 SnF 7 . • Entropy of the Pa-3↔Pm-3m phase transition does not depend on pressure.

  8. Inhibitors and facilitators of willingness to participate (WTP) in an HIV vaccine trial: construction and initial validation of the Inhibitors and Facilitators of Willingness to Participate Scale (WPS) among women at risk for HIV infection.

    Science.gov (United States)

    Fincham, Dylan; Kagee, Ashraf; Swartz, Leslie

    2010-04-01

    A psychometric scale assessing inhibitors and facilitators of willingness to participate (WTP) in an HIV vaccine trial has not yet been developed. This study aimed to construct and derive the exploratory factor structure of such a scale. The 35-item Inhibitors and Facilitators of Willingness to Participate Scale (WPS) was developed and administered to a convenience sample of 264 Black females between the ages of 16 and 49 years living in an urban-informal settlement near Cape Town. The subscales of the WPS demonstrated good internal consistency with Cronbach's alpha coefficients ranging between 0.69 and 0.82. A principal components exploratory factor analysis revealed the presence of five latent factors. The factors, which accounted for 45.93% of the variance in WTP, were (1) personal costs, (2) safety and convenience, (3) stigmatisation, (4) personal gains and (5) social approval and trust. Against the backdrop of the study limitations, these results provide initial support for the reliability and construct validity of the WPS among the most eligible trial participants in the Western Cape of South Africa.

  9. Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying (CVD) Facility Operations Manual

    Energy Technology Data Exchange (ETDEWEB)

    IRWIN, J.J.

    2000-02-03

    This document provides the Operations Manual for the Cold Vacuum Drying Facility (CVDF). The Manual was developed in conjunction with HNF-SD-SNF-SAR-002, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of the Processing Systems (Garvin 1998) and, the HNF-SD-SNF-DRD-002, 1997, Cold Vacuum Drying Facility Design Requirements, Rev. 3a. The Operations Manual contains general descriptions of all the process, safety and facility systems in the CVDF, a general CVD operations sequence, and has been developed for the spent nuclear fuel project (SNFP) Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved.

  10. Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying (CVD) Facility Operations Manual

    International Nuclear Information System (INIS)

    IRWIN, J.J.

    2000-01-01

    This document provides the Operations Manual for the Cold Vacuum Drying Facility (CVDF). The Manual was developed in conjunction with HNF-SD-SNF-SAR-002, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of the Processing Systems (Garvin 1998) and, the HNF-SD-SNF-DRD-002, 1997, Cold Vacuum Drying Facility Design Requirements, Rev. 3a. The Operations Manual contains general descriptions of all the process, safety and facility systems in the CVDF, a general CVD operations sequence, and has been developed for the spent nuclear fuel project (SNFP) Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved

  11. Development of the ENVI simulator to estimate Korean SNF flow and its cost - 16060

    International Nuclear Information System (INIS)

    Hwang, Yongsoo; Miller, Ian

    2009-01-01

    This paper describes an integrated model developed by the Korean Atomic Energy Research Institute (KAERI) to simulate options for managing spent nuclear fuel (SNF) in South Korea. A companion paper (Hwang and Miller, 2009) describes a performance assessment model to address the long-term safety of alternative geological disposal options for different waste streams. The model addresses alternative concepts for storage, transportation, and processing of SNF of different types (Candu, PWR), leading up to permanent disposal in geological repositories. It uses the GoldSim software to simulate the logistics of the associated activities, including the associated capital and operating costs. The model's results allow direct comparison of alternative waste management concepts, and predict the sizes and timings of different facilities required. Future versions of the model will also address the uncertainties associated with the different system components in order to provide risk-based assessments. (authors)

  12. 327 SNF fuel return to K-Basin quality process plan

    International Nuclear Information System (INIS)

    Ham, J.E.

    1998-01-01

    The B and W Hanford Company's (BWHC) 327 Facility, in the 300 Area of the Hanford Site, contains Spent Nuclear Fuel (SNF) single fuel element canisters (SFEC) and fuel remnant canisters (FRC) which are to be returned to K-Basin. Seven shipments of up to six fuel canisters will be loaded into the CNS 1-13G Cask and transported to 105-KE

  13. Uranium Oxide Rate Summary for the Spent Nuclear Fuel (SNF) Project (OCRWM)

    Energy Technology Data Exchange (ETDEWEB)

    PAJUNEN, A.L.

    2000-09-20

    The purpose of this document is to summarize the uranium oxidation reaction rate information developed by the Hanford Spent Nuclear Fuel (SNF) Project and describe the basis for selecting reaction rate correlations used in system design. The selection basis considers the conditions of practical interest to the fuel removal processes and the reaction rate application during design studies. Since the reaction rate correlations are potentially used over a range of conditions, depending of the type of evaluation being performed, a method for transitioning between oxidation reactions is also documented. The document scope is limited to uranium oxidation reactions of primary interest to the SNF Project processes. The reactions influencing fuel removal processes, and supporting accident analyses, are: uranium-water vapor, uranium-liquid water, uranium-moist air, and uranium-dry air. The correlation selection basis will consider input from all available sources that indicate the oxidation rate of uranium fuel, including the literature data, confirmatory experimental studies, and fuel element observations. Trimble (2000) summarizes literature data and the results of laboratory scale experimental studies. This document combines the information in Trimble (2000) with larger scale reaction observations to describe uranium oxidation rate correlations applicable to conditions of interest to the SNF Project.

  14. A SWI/SNF Chromatin Remodelling Protein Controls Cytokinin Production through the Regulation of Chromatin Architecture

    KAUST Repository

    Jé gu, Teddy; Domenichini, Sé verine; Blein, Thomas; Ariel, Federico; Christ, Auré lie; Kim, SoonKap; Crespi, Martin; Boutet-Mercey, Sté phanie; Mouille, Gré gory; Bourge, Mickaë l; Hirt, Heribert; Bergounioux, Catherine; Raynaud, Cé cile; Benhamed, Moussa

    2015-01-01

    Chromatin architecture determines transcriptional accessibility to DNA and consequently gene expression levels in response to developmental and environmental stimuli. Recently, chromatin remodelers such as SWI/SNF complexes have been recognized as key regulators of chromatin architecture. To gain insight into the function of these complexes during root development, we have analyzed Arabidopsis knock-down lines for one sub-unit of SWI/SNF complexes: BAF60. Here, we show that BAF60 is a positive regulator of root development and cell cycle progression in the root meristem via its ability to down-regulate cytokinin production. By opposing both the deposition of active histone marks and the formation of a chromatin regulatory loop, BAF60 negatively regulates two crucial target genes for cytokinin biosynthesis (IPT3 and IPT7) and one cell cycle inhibitor (KRP7). Our results demonstrate that SWI/SNF complexes containing BAF60 are key factors governing the equilibrium between formation and dissociation of a chromatin loop controlling phytohormone production and cell cycle progression.

  15. A SWI/SNF Chromatin Remodelling Protein Controls Cytokinin Production through the Regulation of Chromatin Architecture

    KAUST Repository

    Jégu, Teddy

    2015-10-12

    Chromatin architecture determines transcriptional accessibility to DNA and consequently gene expression levels in response to developmental and environmental stimuli. Recently, chromatin remodelers such as SWI/SNF complexes have been recognized as key regulators of chromatin architecture. To gain insight into the function of these complexes during root development, we have analyzed Arabidopsis knock-down lines for one sub-unit of SWI/SNF complexes: BAF60. Here, we show that BAF60 is a positive regulator of root development and cell cycle progression in the root meristem via its ability to down-regulate cytokinin production. By opposing both the deposition of active histone marks and the formation of a chromatin regulatory loop, BAF60 negatively regulates two crucial target genes for cytokinin biosynthesis (IPT3 and IPT7) and one cell cycle inhibitor (KRP7). Our results demonstrate that SWI/SNF complexes containing BAF60 are key factors governing the equilibrium between formation and dissociation of a chromatin loop controlling phytohormone production and cell cycle progression.

  16. Uranium Oxide Rate Summary for the Spent Nuclear Fuel (SNF) Project (OCRWM)

    International Nuclear Information System (INIS)

    PAJUNEN, A.L.

    2000-01-01

    The purpose of this document is to summarize the uranium oxidation reaction rate information developed by the Hanford Spent Nuclear Fuel (SNF) Project and describe the basis for selecting reaction rate correlations used in system design. The selection basis considers the conditions of practical interest to the fuel removal processes and the reaction rate application during design studies. Since the reaction rate correlations are potentially used over a range of conditions, depending of the type of evaluation being performed, a method for transitioning between oxidation reactions is also documented. The document scope is limited to uranium oxidation reactions of primary interest to the SNF Project processes. The reactions influencing fuel removal processes, and supporting accident analyses, are: uranium-water vapor, uranium-liquid water, uranium-moist air, and uranium-dry air. The correlation selection basis will consider input from all available sources that indicate the oxidation rate of uranium fuel, including the literature data, confirmatory experimental studies, and fuel element observations. Trimble (2000) summarizes literature data and the results of laboratory scale experimental studies. This document combines the information in Trimble (2000) with larger scale reaction observations to describe uranium oxidation rate correlations applicable to conditions of interest to the SNF Project

  17. Molten salt reactor related research in Switzerland

    International Nuclear Information System (INIS)

    Krepel, Jiri; Hombourger, Boris; Fiorina, Carlo

    2015-01-01

    Switzerland represented by the Paul Scherrer Institute (PSI) is a member of the Generation IV International Forum (GIF). In the past, the research at PSI focused mainly on HTR, SFR, and GFR. Currently, a research program was established also for Molten Salt Reactors (MSR). Safety is the key point and main interest of the MSR research at the Nuclear Energy and Safety (NES) department of PSI. However, it cannot be evaluated without knowing the system design, fuel chemistry, salt thermal-hydraulics features, safety and fuel cycle approach, and the relevant material and chemical limits. Accordingly, sufficient knowledge should be acquired in the other individual fields before the safety can be evaluated. The MSR research at NES may be divided into four working packages (WP): WP1: MSR core design and fuel cycle, WP2: MSR fuel behavior at nominal and accidental conditions, WP3: MSR thermal-hydraulics and decay heat removal system, WP4: MSR safety, fuel stream, and relevant limits. The WPs are proposed so that there are research topics which can be independently studied within each of them. The work plan of the four WPs is based on several ongoing or past national and international projects relevant to MSR, where NES/PSI participates. At the current stage, the program focuses on several specific and design independent studies. The safety is the key point and main long-term interest of the MSR research at NES. (author)

  18. Design Of Dry Cask Storage For Serpong Multipurpose Reactor Spent Nuclear Fuel

    Directory of Open Access Journals (Sweden)

    Dyah Sulistyani Rahayu

    2018-03-01

    Full Text Available DESIGN OF DRY CASK STORAGE FOR SERPONG MULTI PURPOSE REACTOR SPENT NUCLEAR FUEL. The spent nuclear fuel (SNF from Serpong Multipurpose Reactor, after 100 days storing in the reactor pond, is transferred to water pool interim storage for spent fuel (ISFSF. At present there are a remaining of 245 elements of SNF on the ISSF,198 element of which have been re-exported to the USA. The dry-cask storage allows the SNF, which has already been cooled in the ISSF, to lower its radiation exposure and heat decayat a very low level. Design of the dry cask storage for SNF has been done. Dual purpose of unventilated vertical dry cask was selected among other choices of metal cask, horizontal concrete modules, and modular vaults by taking into account of technical and economical advantages. The designed structure of cask consists of SNF rack canister, inner steel liner, concrete shielding of cask, and outer steel liner. To avoid bimetallic corrosion, the construction material for canister and inner steel liner follows the same material construction of fuel cladding, i.e. the alloy of AlMg2. The construction material of outer steel liner is copper to facilitate the heat transfer from the cask to the atmosphere. The total decay heat is transferred from SNF elements bundle to the atmosphere by a serial of heat transfer resistance for canister wall, inner steel liner, concrete shielding, and outer steel liner respectedly. The rack canister optimum capacity of 34 fuel elements was designed by geometric similarity method basedon SNF position arrangement of 7 x 6 triangular pitch array of fuel elements for prohibiting criticality by spontaneous neutron. The SNF elements are stored vertically on the rack canister.  The thickness of concrete wall shielding was calculated by trial and error to give air temperature of 30 oC and radiation dose on the wall surface of outer liner of 200 mrem/h. The SNF elements bundles originate from the existing racks of wet storage, i

  19. A Green Approach to SNF Reprocessing: Are Common Household Reagents the Answer?

    International Nuclear Information System (INIS)

    Peper, Shane M.; McNamara, Bruce K.; O'Hara, Matthew J.; Douglas, Matthew

    2008-01-01

    It has been discovered that UO2, the principal component of spent nuclear fuel (SNF), can efficiently be dissolved at room temperature using a combination of common household reagents, namely hydrogen peroxide, baking soda, and ammonia. This rather serendipitous discovery opens up the possibility, for the first time, of considering a non-acidic process for recycling U from SNF. Albeit at the early stages of development, our unconventional dissolution approach possesses many attractive features that could make it a reality in the future. With dissolution byproducts of water and oxygen, our approach poses a minimal threat to the environment. Moreover, the use of common household reagents to afford actinide oxide dissolution suggests a certain degree of economic favorability. With the use of a ''closed'' digestion vessel as a reaction chamber, our approach has substantial versatility with the option of using either aqueous or gaseous reactant feeds or a combination of both. Our approach distinguishes itself from all existing reprocessing technologies in two important ways. First and foremost, it is an alkaline rather than an acidic process, using mild non-corrosive chemicals under ambient conditions to effect actinide separations. Secondly, it does not dissolve the entire SNF matrix, but rather selectively solubilizes U and other light actinides for subsequent separation, resulting in potentially faster head-end dissolution and fewer downstream separation steps. From a safeguards perspective, the use of oxidizing alkaline solutions to effect actinide separations also potentially offers a degree of inherent proliferation resistance, by allowing the U to be selectively removed from the remaining dissolver solution while keeping Pu grouped with the other minor actinides and fission products. This paper will describe the design and general experimental setup of a 'closed' digestion vessel for performing uranium oxide dissolutions under alkaline conditions using gaseous

  20. A Green Approach to SNF Reprocessing: Are Common Household Reagents the Answer?

    Energy Technology Data Exchange (ETDEWEB)

    Peper, Shane M.; McNamara, Bruce K.; O' Hara, Matthew J.; Douglas, Matthew

    2008-04-03

    It has been discovered that UO2, the principal component of spent nuclear fuel (SNF), can efficiently be dissolved at room temperature using a combination of common household reagents, namely hydrogen peroxide, baking soda, and ammonia. This rather serendipitous discovery opens up the possibility, for the first time, of considering a non-acidic process for recycling U from SNF. Albeit at the early stages of development, our unconventional dissolution approach possesses many attractive features that could make it a reality in the future. With dissolution byproducts of water and oxygen, our approach poses a minimal threat to the environment. Moreover, the use of common household reagents to afford actinide oxide dissolution suggests a certain degree of economic favorability. With the use of a “closed” digestion vessel as a reaction chamber, our approach has substantial versatility with the option of using either aqueous or gaseous reactant feeds or a combination of both. Our approach distinguishes itself from all existing reprocessing technologies in two important ways. First and foremost, it is an alkaline rather than an acidic process, using mild non-corrosive chemicals under ambient conditions to effect actinide separations. Secondly, it does not dissolve the entire SNF matrix, but rather selectively solubilizes U and other light actinides for subsequent separation, resulting in potentially faster head-end dissolution and fewer downstream separation steps. From a safeguards perspective, the use of oxidizing alkaline solutions to effect actinide separations also potentially offers a degree of inherent proliferation resistance, by allowing the U to be selectively removed from the remaining dissolver solution while keeping Pu grouped with the other minor actinides and fission products. This paper will describe the design and general experimental setup of a “closed” digestion vessel for performing uranium oxide dissolutions under alkaline conditions using

  1. Downregulation of SWI/SNF chromatin remodeling factor subunits modulates cisplatin cytotoxicity

    International Nuclear Information System (INIS)

    Kothandapani, Anbarasi; Gopalakrishnan, Kathirvel; Kahali, Bhaskar; Reisman, David; Patrick, Steve M.

    2012-01-01

    Chromatin remodeling complex SWI/SNF plays important roles in many cellular processes including transcription, proliferation, differentiation and DNA repair. In this report, we investigated the role of SWI/SNF catalytic subunits Brg1 and Brm in the cellular response to cisplatin in lung cancer and head/neck cancer cells. Stable knockdown of Brg1 and Brm enhanced cellular sensitivity to cisplatin. Repair kinetics of cisplatin DNA adducts revealed that downregulation of Brg1 and Brm impeded the repair of both intrastrand adducts and interstrand crosslinks (ICLs). Cisplatin ICL-induced DNA double strand break repair was also decreased in Brg1 and Brm depleted cells. Altered checkpoint activation with enhanced apoptosis as well as impaired chromatin relaxation was observed in Brg1 and Brm deficient cells. Downregulation of Brg1 and Brm did not affect the recruitment of DNA damage recognition factor XPC to cisplatin DNA lesions, but affected ERCC1 recruitment, which is involved in the later stages of DNA repair. Based on these results, we propose that SWI/SNF chromatin remodeling complex modulates cisplatin cytotoxicity by facilitating efficient repair of the cisplatin DNA lesions. -- Highlights: ► Stable knockdown of Brg1 and Brm enhances cellular sensitivity to cisplatin. ► Downregulation of Brg1 and Brm impedes the repair of cisplatin intrastrand adducts and interstrand crosslinks. ► Brg1 and Brm deficiency results in impaired chromatin relaxation, altered checkpoint activation as well as enhanced apoptosis. ► Downregulation of Brg1 and Brm affects recruitment of ERCC1, but not XPC to cisplatin DNA lesions.

  2. The SNF2-family member Fun30 promotes gene silencing in heterochromatic loci.

    Directory of Open Access Journals (Sweden)

    Ana Neves-Costa

    2009-12-01

    Full Text Available Chromatin regulates many key processes in the nucleus by controlling access to the underlying DNA. SNF2-like factors are ATP-driven enzymes that play key roles in the dynamics of chromatin by remodelling nucleosomes and other nucleoprotein complexes. Even simple eukaryotes such as yeast contain members of several subfamilies of SNF2-like factors. The FUN30/ETL1 subfamily of SNF2 remodellers is conserved from yeasts to humans, but is poorly characterized. We show that the deletion of FUN30 leads to sensitivity to the topoisomerase I poison camptothecin and to severe cell cycle progression defects when the Orc5 subunit is mutated. We demonstrate a role of FUN30 in promoting silencing in the heterochromatin-like mating type locus HMR, telomeres and the rDNA repeats. Chromatin immunoprecipitation experiments demonstrate that Fun30 binds at the boundary element of the silent HMR and within the silent HMR. Mapping of nucleosomes in vivo using micrococcal nuclease demonstrates that deletion of FUN30 leads to changes of the chromatin structure at the boundary element. A point mutation in the ATP-binding site abrogates the silencing function of Fun30 as well as its toxicity upon overexpression, indicating that the ATPase activity is essential for these roles of Fun30. We identify by amino acid sequence analysis a putative CUE motif as a feature of FUN30/ETL1 factors and show that this motif assists Fun30 activity. Our work suggests that Fun30 is directly involved in silencing by regulating the chromatin structure within or around silent loci.

  3. Plan for Characterization of K Basin Spent Nuclear Fuel (SNF) and Sludge (OCRWM)

    International Nuclear Information System (INIS)

    TRIMBLE, D.J.

    2000-01-01

    This is an update of the plan for the characterization of spent nuclear fuel (SNF) and sludge stored in the Hanford K West and K East Basins. The purpose of the characterization program is to provide fuel and sludge data in support of the SNF Project in the effort to remove the fuel from the K Basins and place it into dry storage. Characterization of the K Basin fuel and sludge was initiated in 1994 and has been guided by the characterization plans (Abrefah 1994, Lawrence 1995a, Lawrence 1995b) and the characterization program management plan (PMP) (Lawrence 1995c, Lawrence 1998, Trimble 1999). The fuel characterization was completed in 1999. Summaries of these activities were documented by Lawrence (1999) and Suyama (1999). Lawrence (1999) is a summary report providing a road map to the detailed documentation of the fuel characterization. Suyama (1999) provides a basis for the limited characterization sample size as it relates to supporting design limits and the operational safety envelope for the SNF Project. The continuing sludge characterization is guided by a data quality objective (DQO) (Makenas 2000) and a sampling and analysis plan (SAP) (Baker, Welsh and Makenas 2000) The original intent of the characterization program was ''to provide bounding behavior for the fuel'' (Lawrence 1995a). To accomplish this objective, a fuel characterization program was planned that would provide data to augment data from the literature. The program included in-situ examinations of the stored fuel and laboratory testing of individual elements and small samples of fuel (Lawrence 1995a). Some of the planned tests were scaled down or canceled due to the changing needs of the SNF Project. The fundamental technical basis for the process that will be used to place the K Basin fuel into dry storage was established by several key calculations. These calculations characterized nominal and bounding behavior of fuel in Multi-Canister Overpacks (MCOs) during processing and storage

  4. Technical Basis - Spent Nuclear Fuels (SNF) Project Radiation and Contamination Trending Program

    International Nuclear Information System (INIS)

    ELGIN, J.C.

    2000-01-01

    This report documents the technical basis for the Spent Nuclear Fuel (SNF) Program radiation and contamination trending program. The program consists of standardized radiation and contamination surveys of the KE Basin, radiation surveys of the KW basin, radiation surveys of the Cold Vacuum Drying Facility (CVD), and radiation surveys of the Canister Storage Building (CSB) with the associated tracking. This report also discusses the remainder of radiological areas within the SNFP that do not have standardized trending programs and the basis for not having this program in those areas

  5. Confirmation of an ARID2 defect in SWI/SNF-related intellectual disability.

    Science.gov (United States)

    Van Paemel, Ruben; De Bruyne, Pauline; van der Straaten, Saskia; D'hondt, Marleen; Fränkel, Urlien; Dheedene, Annelies; Menten, Björn; Callewaert, Bert

    2017-11-01

    We present a 4-year-old girl with delayed neuromotor development, short stature of prenatal onset, and specific behavioral and craniofacial features harboring an intragenic deletion in the ARID2 gene. The phenotype confirmed the major features of the recently described ARID2-related intellectual disability syndrome. However, our patient showed overlapping features with Nicolaides-Baraitser syndrome and Coffin-Siris syndrome, providing further arguments to reclassify these disorders as "SWI/SNF-related intellectual disability syndromes." © 2017 Wiley Periodicals, Inc.

  6. Radionuclide Inventories for DOE SNF Waste Stream and Uranium/Thorium Carbide Fuels

    International Nuclear Information System (INIS)

    K.L. Goluoglu

    2000-01-01

    The objective of this calculation is to generate radionuclide inventories for the Department of Energy (DOE) spent nuclear fuel (SNF) waste stream destined for disposal at the potential repository at Yucca Mountain. The scope of this calculation is limited to the calculation of two radionuclide inventories; one for all uranium/thorium carbide fuels in the waste stream and one for the entire waste stream. These inventories will provide input in future screening calculations to be performed by Performance Assessment to determine important radionuclides

  7. Design of dry cask storage for Serpong multi purpose reactor spent nuclear fuel

    International Nuclear Information System (INIS)

    Dyah Sulistyani Rahayu; Yuli Purwanto; Zainus Salimin

    2018-01-01

    The spent nuclear fuel (SNF) from Serpong Multipurpose Reactor, after 100 days storing in the reactor pond, is transferred to water pool interim storage for spent fuel (ISFSF). At present there are a remaining of 245 elements of SNF on the ISSF, 198 element of which have been re-exported to the USA. The dry-cask storage allows the SNF, which has already been cooled in the ISSF, to lower its radiation exposure and heat decay at a very low level. Design of the dry cask storage for SNF has been done. Dual purpose of unventilated vertical dry cask was selected among other choices of metal cask, horizontal concrete modules, and modular vaults by taking into account of technical and economical advantages. The designed structure of cask consists of SNF rack canister, inner steel liner, concrete shielding of cask, and outer steel liner. To avoid bimetallic corrosion, the construction material for canister and inner steel liner follows the same material construction of fuel cladding, i.e. the alloy of AlMg 2 . The construction material of outer steel liner is copper to facilitate the heat transfer from the cask to the atmosphere. The total decay heat is transferred from SNF elements bundle to the atmosphere by a serial of heat transfer resistance for canister wall, inner steel liner, concrete shielding, and outer steel liner respectedly. The rack canister optimum capacity of 34 fuel elements was designed by geometric similarity method based on SNF position arrangement of 7 x 6 triangular pitch array of fuel elements for prohibiting criticality by spontaneous neutron. The SNF elements are stored vertically on the rack canister. The thickness of concrete wall shielding was calculated by trial and error to give air temperature of 30 °C and radiation dose on the wall surface of outer liner of 200 mrem/h. The SNF elements bundles originate from the existing racks of wet storage, i.e. rack canister no 3, 8 and 10. The value of I 0 from the rack no 3, 8 and 10 are 434.307; 446

  8. A novel Snf2 protein maintains trans-generational regulatory states established by paramutation in maize.

    Directory of Open Access Journals (Sweden)

    Christopher J Hale

    2007-10-01

    Full Text Available Paramutations represent heritable epigenetic alterations that cause departures from Mendelian inheritance. While the mechanism responsible is largely unknown, recent results in both mouse and maize suggest paramutations are correlated with RNA molecules capable of affecting changes in gene expression patterns. In maize, multiple required to maintain repression (rmr loci stabilize these paramutant states. Here we show rmr1 encodes a novel Snf2 protein that affects both small RNA accumulation and cytosine methylation of a proximal transposon fragment at the Pl1-Rhoades allele. However, these cytosine methylation differences do not define the various epigenetic states associated with paramutations. Pedigree analyses also show RMR1 does not mediate the allelic interactions that typically establish paramutations. Strikingly, our mutant analyses show that Pl1-Rhoades RNA transcript levels are altered independently of transcription rates, implicating a post-transcriptional level of RMR1 action. These results suggest the RNA component of maize paramutation maintains small heterochromatic-like domains that can affect, via the activity of a Snf2 protein, the stability of nascent transcripts from adjacent genes by way of a cotranscriptional repression process. These findings highlight a mechanism by which alleles of endogenous loci can acquire novel expression patterns that are meiotically transmissible.

  9. Progress in realization of the state policy in RW and SNF Management in the Russian Federation

    International Nuclear Information System (INIS)

    Borzunov, Andrey I.

    1999-01-01

    The basic infrastructure at the majority of the enterprises for management of radioactive waste (RW) and spent nuclear fuel (SNF) built in Russia in the 1960s and 1970s are now morally and technically obsolete and require reconstruction. As stated in this presentation, the most complicated problem is the shortage of financial resources, and International support is very important. The presentation is organised in sections discussing (1) the problem, (2) basic aspects of the State policy in this field, (3) the federal institutions in charge, (4) the principles upon which the State policy is grounded, (5) the main objectives of the RW and SNF management in Russia, (6) the federal programme: Radioactive wastes and spent nuclear materials management, their disposal and burial for the period 1996-2005, (7) plans for impending solution of the problems of the Northern and Pacific regions of Russia, (8) some top priority work of Minatom, (9) measures planned at the Russian power plants, (10) some basic results so far, (11) international co-operation

  10. Coffin-Siris syndrome is a SWI/SNF complex disorder.

    Science.gov (United States)

    Tsurusaki, Y; Okamoto, N; Ohashi, H; Mizuno, S; Matsumoto, N; Makita, Y; Fukuda, M; Isidor, B; Perrier, J; Aggarwal, S; Dalal, A B; Al-Kindy, A; Liebelt, J; Mowat, D; Nakashima, M; Saitsu, H; Miyake, N; Matsumoto, N

    2014-06-01

    Coffin-Siris syndrome (CSS) is a congenital disorder characterized by intellectual disability, growth deficiency, microcephaly, coarse facial features, and hypoplastic or absent fifth fingernails and/or toenails. We previously reported that five genes are mutated in CSS, all of which encode subunits of the switch/sucrose non-fermenting (SWI/SNF) ATP-dependent chromatin-remodeling complex: SMARCB1, SMARCA4, SMARCE1, ARID1A, and ARID1B. In this study, we examined 49 newly recruited CSS-suspected patients, and re-examined three patients who did not show any mutations (using high-resolution melting analysis) in the previous study, by whole-exome sequencing or targeted resequencing. We found that SMARCB1, SMARCA4, or ARID1B were mutated in 20 patients. By examining available parental samples, we ascertained that 17 occurred de novo. All mutations in SMARCB1 and SMARCA4 were non-truncating (missense or in-frame deletion) whereas those in ARID1B were all truncating (nonsense or frameshift deletion/insertion) in this study as in our previous study. Our data further support that CSS is a SWI/SNF complex disorder. © 2013 John Wiley & Sons A/S. Published by John Wiley & Sons Ltd.

  11. Progress in realization of the state policy in RW and SNF Management in the Russian Federation

    Energy Technology Data Exchange (ETDEWEB)

    Borzunov, Andrey I

    1999-07-01

    The basic infrastructure at the majority of the enterprises for management of radioactive waste (RW) and spent nuclear fuel (SNF) built in Russia in the 1960s and 1970s are now morally and technically obsolete and require reconstruction. As stated in this presentation, the most complicatedproblem is the shortage of financial resources, and International support is very important. The presentation is organised in sections discussing (1) the problem, (2) basic aspects of the State policy in this field, (3) the federal institutions in charge, (4) the principles upon which the State policy is grounded, (5) the main objectives of the RW and SNF management in Russia, (6) the federal programme: Radioactive wastes and spent nuclear materials management, their disposal and burial for the period 1996-2005, (7) plans for impending solution of the problems of the Northern and Pacific regions of Russia, (8) some top priority work of Minatom, (9) measures planned at the Russian power plants, (10) some basic results so far, (11) international co-operation.

  12. Fanconi anemia protein, FANCA, associates with BRG1, a component of the human SWI/SNF complex.

    Science.gov (United States)

    Otsuki, T; Furukawa, Y; Ikeda, K; Endo, H; Yamashita, T; Shinohara, A; Iwamatsu, A; Ozawa, K; Liu, J M

    2001-11-01

    Fanconi anemia (FA) is a genetic disorder that predisposes to hematopoietic failure, birth defects and cancer. We identified an interaction between the FA protein, FANCA and brm-related gene 1 (BRG1) product. BRG1 is a subunit of the SWI/SNF complex, which remodels chromatin structure through a DNA-dependent ATPase activity. FANCA was demonstrated to associate with the endogenous SWI/SNF complex. We also found a significant increase in the molecular chaperone, glucose-regulated protein 94 (GRP94) among BRG1-associated factors isolated from a FANCA-mutant cell line, which was not seen in either a normal control cell line or the mutant line complemented by wild-type FANCA. Despite this specific difference, FANCA did not appear to be absolutely required for in vitro chromatin remodeling. Finally, we demonstrated co-localization in the nucleus between transfected FANCA and BRG1. The physiological action of FANCA on the SWI/SNF complex remains to be clarified, but our work suggests that FANCA may recruit the SWI/SNF complex to target genes, thereby enabling coupled nuclear functions such as transcription and DNA repair.

  13. Spent Nuclear Fuel project stage and store K basin SNF in canister storage building functions and requirements. Revision 1

    International Nuclear Information System (INIS)

    Womack, J.C.

    1995-01-01

    This document establishes the functions and requirements baseline for the implementation of the Canister Storage Building Subproject. The mission allocated to the Canister Storage Building Subproject is to provide safe, environmentally sound staging and storage of K Basin SNF until a decision on the final disposition is reached and implemented

  14. Chromatin-remodeling SWI/SNF complex regulates coenzyme Q6 synthesis and a metabolic shift to respiration in yeast.

    Science.gov (United States)

    Awad, Agape M; Venkataramanan, Srivats; Nag, Anish; Galivanche, Anoop Raj; Bradley, Michelle C; Neves, Lauren T; Douglass, Stephen; Clarke, Catherine F; Johnson, Tracy L

    2017-09-08

    Despite its relatively streamlined genome, there are many important examples of regulated RNA splicing in Saccharomyces cerevisiae Here, we report a role for the chromatin remodeler SWI/SNF in respiration, partially via the regulation of splicing. We find that a nutrient-dependent decrease in Snf2 leads to an increase in splicing of the PTC7 transcript. The spliced PTC7 transcript encodes a mitochondrial phosphatase regulator of biosynthesis of coenzyme Q 6 (ubiquinone or CoQ 6 ) and a mitochondrial redox-active lipid essential for electron and proton transport in respiration. Increased splicing of PTC7 increases CoQ 6 levels. The increase in PTC7 splicing occurs at least in part due to down-regulation of ribosomal protein gene expression, leading to the redistribution of spliceosomes from this abundant class of intron-containing RNAs to otherwise poorly spliced transcripts. In contrast, a protein encoded by the nonspliced isoform of PTC7 represses CoQ 6 biosynthesis. Taken together, these findings uncover a link between Snf2 expression and the splicing of PTC7 and establish a previously unknown role for the SWI/SNF complex in the transition of yeast cells from fermentative to respiratory modes of metabolism. © 2017 by The American Society for Biochemistry and Molecular Biology, Inc.

  15. Standardization of Fat:SNF ratio of milk and addition of sprouted wheat fada (semolina) for the manufacture of halvasan.

    Science.gov (United States)

    Chaudhary, Apurva H; Patel, H G; Prajapati, P S; Prajapati, J P

    2015-04-01

    Traditional Indian Dairy Products such as Halvasan are manufactured in India using an age old practice. For manufacture of such products industrially, a standard formulation is required. Halvasan is a region specific, very popular heat desiccated milk product but has not been studied scientifically. Fat and Solids-not-fat (SNF) plays an important role in physico-chemical, sensory, textural characteristics and also the shelf life of any milk sweet. Hence for process standardization of Halvasan manufacture, different levels of Fat:SNF ratios i.e. 0.44, 0.55, 0.66 and 0.77 of milk were studied so that an optimum level yielding best organoleptic characteristics in final product can be selected. The product was made from milk standardized to these ratios of Fat:SNF and the product was manufactured as per the method tentatively employed on the basis of characterization of market samples of the product in laboratory. Based on the sensory results obtained, a Fat:SNF ratio of 0.66 for the milk has been selected. In the similar way, for standardizing the rate of addition of fada (semolina); 30, 40, 50 and 60 g fada (semolina) per kg of milk were added and based on the sensory observations, the level of fada (semolina) addition @50 gm/kg of milk was adjudged the best for Halvasan manufacture and hence selected.

  16. Advantages of Production of New Fissionable Nuclides for the Nuclear Power Industry in Hybrid Fusion-Fission Reactors

    Science.gov (United States)

    Tsibulskiy, V. F.; Andrianova, E. A.; Davidenko, V. D.; Rodionova, E. V.; Tsibulskiy, S. V.

    2017-12-01

    A concept of a large-scale nuclear power engineering system equipped with fusion and fission reactors is presented. The reactors have a joint fuel cycle, which imposes the lowest risk of the radiation impact on the environment. The formation of such a system is considered within the framework of the evolution of the current nuclear power industry with the dominance of thermal reactors, gradual transition to the thorium fuel cycle, and integration into the system of the hybrid fusion-fission reactors for breeding nuclear fuel for fission reactors. Such evolution of the nuclear power engineering system will allow preservation of the existing structure with the dominance of thermal reactors, enable the reprocessing of the spent nuclear fuel (SNF) with low burnup, and prevent the dangerous accumulation of minor actinides. The proposed structure of the nuclear power engineering system minimizes the risk of radioactive contamination of the environment and the SNF reprocessing facilities, decreasing it by more than one order of magnitude in comparison with the proposed scheme of closing the uranium-plutonium fuel cycle based on the reprocessing of SNF with high burnup from fast reactors.

  17. The Arabidopsis SWI/SNF protein BAF60 mediates seedling growth control by modulating DNA accessibility

    KAUST Repository

    Jégu, Teddy

    2017-06-15

    Plant adaptive responses to changing environments involve complex molecular interplays between intrinsic and external signals. Whilst much is known on the signaling components mediating diurnal, light, and temperature controls on plant development, their influence on chromatin-based transcriptional controls remains poorly explored.In this study we show that a SWI/SNF chromatin remodeler subunit, BAF60, represses seedling growth by modulating DNA accessibility of hypocotyl cell size regulatory genes. BAF60 binds nucleosome-free regions of multiple G box-containing genes, opposing in cis the promoting effect of the photomorphogenic and thermomorphogenic regulator Phytochrome Interacting Factor 4 (PIF4) on hypocotyl elongation. Furthermore, BAF60 expression level is regulated in response to light and daily rhythms.These results unveil a short path between a chromatin remodeler and a signaling component to fine-tune plant morphogenesis in response to environmental conditions.

  18. Spent Nuclear Fuel (SNF) Cold Vacuum Drying (CVD) Facility Operations Manual; FINAL

    International Nuclear Information System (INIS)

    IRWIN, J.J.

    1999-01-01

    This document provides the Operations Manual for the Cold Vacuum Drying Facility (CVDF). The Manual was developed in conjunction with HNF-553, Spent Nuclear Fuel Project Final Safety Analysis Report Annex B-Cold Vacuum Drying Facility. The HNF-SD-SNF-DRD-002, 1999, Cold Vacuum Drying Facility Design Requirements, Rev. 4, and the CVDF Final Design Report. The Operations Manual contains general descriptions of all the process, safety and facility systems in the CVDF, a general CVD operations sequence and references to the CVDF System Design Descriptions (SDDs). This manual has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved

  19. FY17 Status Report: Research on Stress Corrosion Cracking of SNF Interim Storage Canisters.

    Energy Technology Data Exchange (ETDEWEB)

    Schindelholz, Eric John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Alexander, Christopher L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-09-01

    This progress report describes work done in FY17 at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. Work in FY17 refined our understanding of the chemical and physical environment on canister surfaces, and evaluated the relationship between chemical and physical environment and the form and extent of corrosion that occurs. The SNL corrosion work focused predominantly on pitting corrosion, a necessary precursor for SCC, and process of pit-to-crack transition; it has been carried out in collaboration with university partners. SNL is collaborating with several university partners to investigate SCC crack growth experimentally, providing guidance for design and interpretation of experiments.

  20. Spent Nuclear Fuel (SNF) Cold Vacuum Drying (CVD) Facility Operations Manual

    Energy Technology Data Exchange (ETDEWEB)

    IRWIN, J.J.

    1999-07-02

    This document provides the Operations Manual for the Cold Vacuum Drying Facility (CVDF). The Manual was developed in conjunction with HNF-553, Spent Nuclear Fuel Project Final Safety Analysis Report Annex B--Cold Vacuum Drying Facility. The HNF-SD-SNF-DRD-002, 1999, Cold Vacuum Drying Facility Design Requirements, Rev. 4, and the CVDF Final Design Report. The Operations Manual contains general descriptions of all the process, safety and facility systems in the CVDF, a general CVD operations sequence and references to the CVDF System Design Descriptions (SDDs). This manual has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved.

  1. The Arabidopsis SWI/SNF protein BAF60 mediates seedling growth control by modulating DNA accessibility

    KAUST Repository

    Jé gu, Teddy; Veluchamy, Alaguraj; Ramirez Prado, Juan Sebastian; Rizzi-Paillet, Charley; Perez, Magalie; Lhomme, Anaï s; Latrasse, David; Coleno, Emeline; Vicaire, Serge; Legras, Sté phanie; Jost, Bernard; Rougé e, Martin; Barneche, Fredy; Bergounioux, Catherine; Crespi, Martin; Mahfouz, Magdy M.; Hirt, Heribert; Raynaud, Cé cile; Benhamed, Moussa

    2017-01-01

    Plant adaptive responses to changing environments involve complex molecular interplays between intrinsic and external signals. Whilst much is known on the signaling components mediating diurnal, light, and temperature controls on plant development, their influence on chromatin-based transcriptional controls remains poorly explored.In this study we show that a SWI/SNF chromatin remodeler subunit, BAF60, represses seedling growth by modulating DNA accessibility of hypocotyl cell size regulatory genes. BAF60 binds nucleosome-free regions of multiple G box-containing genes, opposing in cis the promoting effect of the photomorphogenic and thermomorphogenic regulator Phytochrome Interacting Factor 4 (PIF4) on hypocotyl elongation. Furthermore, BAF60 expression level is regulated in response to light and daily rhythms.These results unveil a short path between a chromatin remodeler and a signaling component to fine-tune plant morphogenesis in response to environmental conditions.

  2. Vertical Drop of the Naval SNF Long Waste Package On Unyielding Surface

    International Nuclear Information System (INIS)

    S. Mastilovic

    2006-01-01

    The purpose of this calculation is to determine the structural response of a Naval SNF (Spent Nuclear Fuel) Long Waste Package (WP) subjected to 2 m-vertical drop on unyielding surface (US). The scope of this document is limited to reporting the calculation results in terms of maximum stress intensities. This calculation is associated with the waste package design; calculation is performed by the Waste Package Design group. AP-3.12Q, Revision 0, ICN 0, Calculations, is used to perform the calculation and develop the document. The finite element calculation is performed by using the commercially available ANSYS Version (V) 5.4 finite element code. The result of this calculation is provided in terms of maximum stress intensities

  3. Candida albicans Swi/Snf and Mediator Complexes Differentially Regulate Mrr1-Induced MDR1 Expression and Fluconazole Resistance.

    Science.gov (United States)

    Liu, Zhongle; Myers, Lawrence C

    2017-11-01

    Long-term azole treatment of patients with chronic Candida albicans infections can lead to drug resistance. Gain-of-function (GOF) mutations in the transcription factor Mrr1 and the consequent transcriptional activation of MDR1 , a drug efflux coding gene, is a common pathway by which this human fungal pathogen acquires fluconazole resistance. This work elucidates the previously unknown downstream transcription mechanisms utilized by hyperactive Mrr1. We identified the Swi/Snf chromatin remodeling complex as a key coactivator for Mrr1, which is required to maintain basal and induced open chromatin, and Mrr1 occupancy, at the MDR1 promoter. Deletion of snf2 , the catalytic subunit of Swi/Snf, largely abrogates the increases in MDR1 expression and fluconazole MIC observed in MRR1 GOF mutant strains. Mediator positively and negatively regulates key Mrr1 target promoters. Deletion of the Mediator tail module med3 subunit reduces, but does not eliminate, the increased MDR1 expression and fluconazole MIC conferred by MRR1 GOF mutations. Eliminating the kinase activity of the Mediator Ssn3 subunit suppresses the decreased MDR1 expression and fluconazole MIC of the snf2 null mutation in MRR1 GOF strains. Ssn3 deletion also suppresses MDR1 promoter histone displacement defects in snf2 null mutants. The combination of this work with studies on other hyperactive zinc cluster transcription factors that confer azole resistance in fungal pathogens reveals a complex picture where the induction of drug efflux pump expression requires the coordination of multiple coactivators. The observed variations in transcription factor and target promoter dependence of this process may make the search for azole sensitivity-restoring small molecules more complicated. Copyright © 2017 American Society for Microbiology.

  4. DESIGN VERIFICATION REPORT SPENT NUCLEAR FUEL (SNF) PROJECT CANISTER STORAGE BUILDING (CSB)

    Energy Technology Data Exchange (ETDEWEB)

    BAZINET, G.D.

    2003-02-12

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. The original version of this document was prepared by Vista Engineering for the SNF Project. Revision 1 documented verification actions that were pending at the time the initial report was prepared. Revision 3 of this document incorporates MCO Cover Cap Assembly welding verification activities. Verification activities for the installed and operational SSCs have been completed.

  5. Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

    Energy Technology Data Exchange (ETDEWEB)

    BAZINET, G.D.

    2001-05-15

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. The original version of this document was prepared by Vista Engineering for the SNF Project. Revision 1 documented verification actions that were pending at the time the initial report was prepared. Verification activities for the installed and operational SSCs have been completed. Verification of future additions to the CSB related to the canister cover cap and welding fixture system and MCO Internal Gas Sampling equipment will be completed as appropriate for those components. The open items related to verification of those requirements are noted in section 3

  6. DESIGN VERIFICATION REPORT SPENT NUCLEAR FUEL (SNF) PROJECT CANISTER STORAGE BUILDING (CSB)

    International Nuclear Information System (INIS)

    BAZINET, G.D.

    2003-01-01

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. The original version of this document was prepared by Vista Engineering for the SNF Project. Revision 1 documented verification actions that were pending at the time the initial report was prepared. Revision 3 of this document incorporates MCO Cover Cap Assembly welding verification activities. Verification activities for the installed and operational SSCs have been completed

  7. Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

    International Nuclear Information System (INIS)

    BAZINET, G.D.

    2001-01-01

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. The original version of this document was prepared by Vista Engineering for the SNF Project. Revision 1 documented verification actions that were pending at the time the initial report was prepared. Verification activities for the installed and operational SSCs have been completed. Verification of future additions to the CSB related to the canister cover cap and welding fixture system and MCO Internal Gas Sampling equipment will be completed as appropriate for those components. The open items related to verification of those requirements are noted in section 3.1.5 and will be

  8. Transportation of failed or damaged foreign research reactor spent nuclear fuel

    International Nuclear Information System (INIS)

    Messick, C.E.; Mustin, T.P.; Massey, C.D.

    1998-01-01

    Since resuming the Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance Program in 1996, the Program has had to deal with difficult issues associated with the transportation of failed or damaged spent fuel. In several instances, problems with failed or damaged fuel have prevented the acceptance of the fuel at considerable cost to both the Department of Energy (DOE) and research reactor operators. In response to the problems faced by the Acceptance Program, DOE has undertaken significant steps to better define the spent fuel acceptance criteria. DOE has worked closely with the U.S. Nuclear Regulatory Commission to address failed or damaged research reactor spent fuel and to identify cask certificate issues which must be resolved by cask owners and foreign regulatory authorities. The specific issues associated with the transport of Materials Testing Reactor (MTR)-type FRR SNF will be discussed. The information presented will include U.S. Nuclear Regulatory Commission regulatory issues, cask certificate issues, technical constraints, and lessons learned. Specific information will also be provided on the latest efforts to revise DOE's Appendix B, Transport Package (Cask) Acceptance Criteria. The information presented in this paper will be important to foreign research reactor operators, shippers, and cask vendors, so that appropriate amendments to the Certificate of Compliance for spent fuel casks can be submitted in a timely manner to facilitate the safe and scheduled transport of FRR SNF

  9. The pathway by which the yeast protein kinase Snf1p controls acquisition of sodium tolerance is different from that mediating glucose regulation.

    Science.gov (United States)

    Ye, Tian; Elbing, Karin; Hohmann, Stefan

    2008-09-01

    It recently became apparent that the highly conserved Snf1p protein kinase plays roles in controlling different cellular processes in the yeast Saccharomyces cerevisiae, in addition to its well-known function in glucose repression/derepression. We have previously reported that Snf1p together with Gis4p controls ion homeostasis by regulating expression of ENA1, which encodes the Ena1p Na(+) extrusion system. In this study we found that Snf1p is rapidly phosphorylated when cells are exposed to NaCl and this phosphorylation is required for the role of Snf1p in Na(+) tolerance. In contrast to activation by low glucose levels, the salt-induced phosphorylation of Snf1p promoted neither phosphorylation nor nuclear export of the Mig1p repressor. The mechanism that prevents Mig1p phosphorylation by active Snf1p under salt stress does not involve either hexokinase PII or the Gis4p regulator. Instead, Snf1p may mediate upregulation of ENA1 expression via the repressor Nrg1p. Activation of Snf1p in response to glucose depletion requires any of the three upstream protein kinases Sak1p, Tos3p and Elm1p, with Sak1p playing the most prominent role. The same upstream kinases were required for salt-induced Snf1p phosphorylation, and also under these conditions Sak1p played the most prominent role. Unexpectedly, however, it appears that Elm1p plays a dual role in acquisition of salt tolerance by activating Snf1p and in a presently unknown parallel pathway. Together, these results indicate that under salt stress Snf1p takes part in a different pathway from that during glucose depletion and this role is performed together as well as in parallel with its upstream kinase Elm1p. Snf1p appears to be part of a wider functional network than previously anticipated and the full complexity of this network remains to be elucidated.

  10. Leaching of radionuclides from furfural-based polymers used to solidify reactor compartments and components disposed of in the Arctic Kara Sea

    International Nuclear Information System (INIS)

    Heiser, J.H.; Sivintsev, Y.; Alexandrov, V.P.; Dyer, R.S.

    1999-01-01

    Within the course of operating its nuclear navy, the former Soviet Union (FSU) disposed of reactor vessels and spent nuclear fuel (SNF) in three fjords on the east coast of Novaya Zemlya and in the open Kara Sea within the Novaya Zemlya Trough during the period 1965 to 1988. The dumping consisted of 16 reactors, six of which contained SNF and one special container that held ca. 60% of the damaged SNF and the screening assembly from the No. 2 reactor of the atomic icebreaker Lenin. At the time, the FSU considered dumping of decommissioned nuclear submarines with damaged cores in the bays of and near by the Novaya Zemlya archipelago in the Arctic Kara Sea to be acceptable. To provide an additional level of safety, a group of Russian scientists embarked upon a course of research to develop a solidification agent that would provide an ecologically safe barrier. The barrier material would prevent direct contact of seawater with the SNF and the resultant leaching and release of radionuclides. The solidification agent was to be introduced by flooding the reactors vessels and inner cavities. Once introduced the agent would harden and form an impermeable barrier. This report describes the sample preparation of several ''Furfurol'' compositions and their leach testing using cesium 137 as tracer

  11. Experimental and numerical investigations on the applicability of local approach methods on the warm prestress effect; Experimentelle und numerische Untersuchungen zur Anwendbarkeit von Local-Approach-Ansaetzen auf WPS-Effekt

    Energy Technology Data Exchange (ETDEWEB)

    Roos, E; Elsaesser, K [Stuttgart Univ. (Germany). Staatliche Materialpruefungsanstalt

    1998-11-01

    The three materials 10 MnMoNi 5 5, as a welded specimen, 22 NiMoCr 3 7 mod., and 17 MoV 8 4 mod. have been examined as a part of project work. The major known influencing variables with respect to the WPS effect are materials hardening, crack tip blunting, and the developing residual stress field. The project task was to investigate to what extent the finite element method will be able to describe these influencing variables. Using the damaging model of Rousselier it is possible to simulate the processes during warm prestressing at upper-shelf toughness while taking into account the crack growth. Using the parameter derived with the Beremin model with notched specimens and applying these to pre-cracked specimens will not yield accurate results describing the fracture toughness in the area affected by embrittlement. (orig./CB) [Deutsch] Im Rahmen des Forschungsprojektes wurden die drei Werkstoffe 10 MnMoNi 5 5 formgeschweisst, 22 NiMoCr 3 7 mod. und 17 MoV 8 4 mod. untersucht. Als hauptsaechliche Einflussgroessen fuer den WPS-Effekt gelten die Werkstoffverfestigung, die Abstumpfung der Rissspitze (Blunting), sowie das sich ausbildende Eigenspannungsfeld. Es wurde untersucht, inwieweit es mit Hilfe der Methode der Finiten Elemente moeglich ist, die einzelnen Einflussgroessen zu beschreiben. Mit Hilfe des Schaedigungsmodelles von Rousselier koennen die Vorgaenge beim Warmvorbelasten in der Zaehigkeitshochlage auch unter Beruecksichtigung von Risswachstum simuliert werden. Bei Verwendung der an gekerbten Proben ermittelten Parameter des Beremin-Modelles und deren Uebertragung auf angerissene Proben wird die Bruchzaehigkeit im Sproedbruchgebiet nicht zutreffend berechnet. (orig./MM)

  12. Research reactor spent nuclear fuel shipment from the Czech Republic to the Russian Federation

    International Nuclear Information System (INIS)

    Svoboda, K.; Broz, V.; Novosad, P.; Podlaha, J.; Svitak, F.

    2009-01-01

    In May 2004, the Global Threat Reduction Initiative agreement was signed by the governments of the United States and the Russian Federation. The goal of this initiative is to minimize, in cooperation with the International Atomic Energy Agency (IAEA) in Vienna, the existing threat of misuse of nuclear and radioactive materials for terrorist purposes, particularly highly enriched uranium (HEU), fresh and spent nuclear fuel (SNF), and plutonium, which have been stored in a number of countries. Within the framework of the initiative, HEU materials and SNF from research reactors of Russian origin will be transported back to the Russian Federation for reprocessing/liquidation. The program is designated as the Russian Research Reactor Fuel Return (RRRFR) Program and is similar to the U.S. Foreign Research Reactor Spent Nuclear Fuel Acceptance Program, which is underway for nuclear materials of United States origin. These RRRFR activities are carried out under the responsibilities of the respective ministries (i.e., U.S. Department of Energy (DOE) and Russian Federation Rosatom). The Czech Republic and the Nuclear Research Institute Rez, plc (NRI) joined Global Threat Reduction Initiative in 2004. During NRI's more than 50 years of existence, radioactive and nuclear materials had accumulated and had been safely stored on its grounds. In 1995, the Czech regulatory body , State Office for Nuclear Safety (SONS), instructed NRI that all ecological burdens from its past activities must be addressed and that the SNF from the research reactor LVR -15 had to be transported for reprocessing. At the end of November 2007, all these activities culminated with the unique shipment to the Russian Federation of 527 fuel assemblies of SNF type EK-10 (enrichment 10% U-235) and IRT-M (enrichment 36% and 80% U-235) and 657 irradiated fuel rods of EK-10 fuel, which were used in LVR-15 reactor. (authors)

  13. Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

    Energy Technology Data Exchange (ETDEWEB)

    BAZINET, G.D.

    2000-11-03

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. The original version of this document was prepared by Vista Engineering for the SNF Project. The purpose of this revision is to document completion of verification actions that were pending at the time the initial report was prepared. Verification activities for the installed and operational SSCs have been completed. Verification of future additions to the CSB related to the canister cover cap and welding fixture system and MCO Internal Gas Sampling equipment will be completed as appropriate for those components. The open items related to verification of those

  14. Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

    International Nuclear Information System (INIS)

    BAZINET, G.D.

    2000-01-01

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. The original version of this document was prepared by Vista Engineering for the SNF Project. The purpose of this revision is to document completion of verification actions that were pending at the time the initial report was prepared. Verification activities for the installed and operational SSCs have been completed. Verification of future additions to the CSB related to the canister cover cap and welding fixture system and MCO Internal Gas Sampling equipment will be completed as appropriate for those components. The open items related to verification of those requirements are noted

  15. ABI1 and PP2CA Phosphatases Are Negative Regulators of Snf1-Related Protein Kinase1 Signaling in Arabidopsis

    OpenAIRE

    Rodrigues, A.; Adamo, M.; Crozet, P.; Margalha, L.; Confraria, A.; Martinho, C.; Elias, A.; Rabissi, A.; Lumbreras, V.; Gonzalez-Guzman, M.; Antoni, R.; Rodriguez, P. L.; Baena-Gonzalez, E.

    2013-01-01

    Plant survival under environmental stress requires the integration of multiple signaling pathways into a coordinated response, but the molecular mechanisms underlying this integration are poorly understood. Stress-derived energy deprivation activates the Snf1-related protein kinases1 (SnRK1s), triggering a vast transcriptional and metabolic reprogramming that restores homeostasis and promotes tolerance to adverse conditions. Here, we show that two clade A type 2C protein phosphatases (PP2Cs),...

  16. Biallelic germline and somatic mutations in malignant mesothelioma: multiple mutations in transcription regulators including mSWI/SNF genes.

    Science.gov (United States)

    Yoshikawa, Yoshie; Sato, Ayuko; Tsujimura, Tohru; Otsuki, Taiichiro; Fukuoka, Kazuya; Hasegawa, Seiki; Nakano, Takashi; Hashimoto-Tamaoki, Tomoko

    2015-02-01

    We detected low levels of acetylation for histone H3 tail lysines in malignant mesothelioma (MM) cell lines resistant to histone deacetylase inhibitors. To identify the possible genetic causes related to the low histone acetylation levels, whole-exome sequencing was conducted with MM cell lines established from eight patients. A mono-allelic variant of BRD1 was common to two MM cell lines with very low acetylation levels. We identified 318 homozygous protein-damaging variants/mutations (18-78 variants/mutations per patient); annotation analysis showed enrichment of the molecules associated with mammalian SWI/SNF (mSWI/SNF) chromatin remodeling complexes and co-activators that facilitate initiation of transcription. In seven of the patients, we detected a combination of variants in histone modifiers or transcription factors/co-factors, in addition to variants in mSWI/SNF. Direct sequencing showed that homozygous mutations in SMARCA4, PBRM1 and ARID2 were somatic. In one patient, homozygous germline variants were observed for SMARCC1 and SETD2 in chr3p22.1-3p14.2. These exhibited extended germline homozygosity and were in regions containing somatic mutations, leading to a loss of BAP1 and PBRM1 expression in MM cell line. Most protein-damaging variants were heterozygous in normal tissues. Heterozygous germline variants were often converted into hemizygous variants by mono-allelic deletion, and were rarely homozygous because of acquired uniparental disomy. Our findings imply that MM might develop through the somatic inactivation of mSWI/SNF complex subunits and/or histone modifiers, including BAP1, in subjects that have rare germline variants of these transcription regulators and/or transcription factors/co-factors, and in regions prone to mono-allelic deletion during oncogenesis. © 2014 UICC.

  17. HIC1 interacts with a specific subunit of SWI/SNF complexes, ARID1A/BAF250A

    International Nuclear Information System (INIS)

    Van Rechem, Capucine; Boulay, Gaylor; Leprince, Dominique

    2009-01-01

    HIC1, a tumor suppressor gene epigenetically silenced in many human cancers encodes a transcriptional repressor involved in regulatory loops modulating p53-dependent and E2F1-dependent cell survival and stress responses. HIC1 is also implicated in growth control since it recruits BRG1, one of the two alternative ATPases (BRM or BRG1) of SWI/SNF chromatin-remodeling complexes to repress transcription of E2F1 in quiescent fibroblasts. Here, through yeast two-hybrid screening, we identify ARID1A/BAF250A, as a new HIC1 partner. ARID1A/BAF250A is one of the two mutually exclusive ARID1-containing subunits of SWI/SNF complexes which define subsets of complexes endowed with anti-proliferative properties. Co-immunoprecipitation assays in WI38 fibroblasts and in BRG1-/- SW13 cells showed that endogenous HIC1 and ARID1A proteins interact in a BRG1-dependent manner. Furthermore, we demonstrate that HIC1 does not interact with BRM. Finally, sequential chromatin immunoprecipitation (ChIP-reChIP) experiments demonstrated that HIC1 represses E2F1 through the recruitment of anti-proliferative SWI/SNF complexes containing ARID1A.

  18. Dynamic Recruitment of Functionally Distinct Swi/Snf Chromatin Remodeling Complexes Modulates Pdx1 Activity in Islet β Cells

    Directory of Open Access Journals (Sweden)

    Brian McKenna

    2015-03-01

    Full Text Available Pdx1 is a transcription factor of fundamental importance to pancreas formation and adult islet β cell function. However, little is known about the positive- and negative-acting coregulators recruited to mediate transcriptional control. Here, we isolated numerous Pdx1-interacting factors possessing a wide range of cellular functions linked with this protein, including, but not limited to, coregulators associated with transcriptional activation and repression, DNA damage response, and DNA replication. Because chromatin remodeling activities are essential to developmental lineage decisions and adult cell function, our analysis focused on investigating the influence of the Swi/Snf chromatin remodeler on Pdx1 action. The two mutually exclusive and indispensable Swi/Snf core ATPase subunits, Brg1 and Brm, distinctly affected target gene expression in β cells. Furthermore, physiological and pathophysiological conditions dynamically regulated Pdx1 binding to these Swi/Snf complexes in vivo. We discuss how context-dependent recruitment of coregulatory complexes by Pdx1 could impact pancreas cell development and adult islet β cell activity.

  19. The SWI/SNF BAF-A complex is essential for neural crest development.

    Science.gov (United States)

    Chandler, Ronald L; Magnuson, Terry

    2016-03-01

    Growing evidence indicates that chromatin remodeler mutations underlie the pathogenesis of human neurocristopathies or disorders that affect neural crest cells (NCCs). However, causal relationships among chromatin remodeler subunit mutations and NCC defects remain poorly understood. Here we show that homozygous loss of ARID1A-containing, SWI/SNF chromatin remodeling complexes (BAF-A) in NCCs results in embryonic lethality in mice, with mutant embryos succumbing to heart defects. Strikingly, monoallelic loss of ARID1A in NCCs led to craniofacial defects in adult mice, including shortened snouts and low set ears, and these defects were more pronounced following homozygous loss of ARID1A, with the ventral cranial bones being greatly reduced in size. Early NCC specification and expression of the BRG1 NCC target gene, PLEXINA2, occurred normally in the absence of ARID1A. Nonetheless, mutant embryos displayed incomplete conotruncal septation of the cardiac outflow tract and defects in the posterior pharyngeal arteries, culminating in persistent truncus arteriosus and agenesis of the ductus arteriosus. Consistent with this, migrating cardiac NCCs underwent apoptosis within the circumpharyngeal ridge. Our data support the notion that multiple, distinct chromatin remodeling complexes govern genetically separable events in NCC development and highlight a potential pathogenic role for NCCs in the human BAF complex disorder, Coffin-Siris Syndrome. Copyright © 2016 Elsevier Inc. All rights reserved.

  20. The role of BAF (mSWI/SNF) complexes in mammalian neural development.

    Science.gov (United States)

    Son, Esther Y; Crabtree, Gerald R

    2014-09-01

    The BAF (mammalian SWI/SNF) complexes are a family of multi-subunit ATP-dependent chromatin remodelers that use ATP hydrolysis to alter chromatin structure. Distinct BAF complex compositions are possible through combinatorial assembly of homologous subunit families and can serve non-redundant functions. In mammalian neural development, developmental stage-specific BAF assemblies are found in embryonic stem cells, neural progenitors and postmitotic neurons. In particular, the neural progenitor-specific BAF complexes are essential for controlling the kinetics and mode of neural progenitor cell division, while neuronal BAF function is necessary for the maturation of postmitotic neuronal phenotypes as well as long-term memory formation. The microRNA-mediated mechanism for transitioning from npBAF to nBAF complexes is instructive for the neuronal fate and can even convert fibroblasts into neurons. The high frequency of BAF subunit mutations in neurological disorders underscores the rate-determining role of BAF complexes in neural development, homeostasis, and plasticity. © 2014 Wiley Periodicals, Inc.

  1. SWI/SNF-like chromatin remodeling factor Fun30 supports point centromere function in S. cerevisiae.

    Directory of Open Access Journals (Sweden)

    Mickaël Durand-Dubief

    2012-09-01

    Full Text Available Budding yeast centromeres are sequence-defined point centromeres and are, unlike in many other organisms, not embedded in heterochromatin. Here we show that Fun30, a poorly understood SWI/SNF-like chromatin remodeling factor conserved in humans, promotes point centromere function through the formation of correct chromatin architecture at centromeres. Our determination of the genome-wide binding and nucleosome positioning properties of Fun30 shows that this enzyme is consistently enriched over centromeres and that a majority of CENs show Fun30-dependent changes in flanking nucleosome position and/or CEN core micrococcal nuclease accessibility. Fun30 deletion leads to defects in histone variant Htz1 occupancy genome-wide, including at and around most centromeres. FUN30 genetically interacts with CSE4, coding for the centromere-specific variant of histone H3, and counteracts the detrimental effect of transcription through centromeres on chromosome segregation and suppresses transcriptional noise over centromere CEN3. Previous work has shown a requirement for fission yeast and mammalian homologs of Fun30 in heterochromatin assembly. As centromeres in budding yeast are not embedded in heterochromatin, our findings indicate a direct role of Fun30 in centromere chromatin by promoting correct chromatin architecture.

  2. A Safety Case Approach for Deep Geologic Disposal of DOE HLW and DOE SNF in Bedded Salt - 13350

    Energy Technology Data Exchange (ETDEWEB)

    Sevougian, S. David [Advanced Nuclear Energy Programs Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); MacKinnon, Robert J. [Advanced Nuclear Energy Programs Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); Leigh, Christi D. [Defense Waste Management Programs Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); Hansen, Frank D. [Geoscience Research and Applications Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States)

    2013-07-01

    The primary objective of this study is to investigate the feasibility and utility of developing a defensible safety case for disposal of United States Department of Energy (U.S. DOE) high-level waste (HLW) and DOE spent nuclear fuel (SNF) in a conceptual deep geologic repository that is assumed to be located in a bedded salt formation of the Delaware Basin [1]. A safety case is a formal compilation of evidence, analyses, and arguments that substantiate and demonstrate the safety of a proposed or conceptual repository. We conclude that a strong initial safety case for potential licensing can be readily compiled by capitalizing on the extensive technical basis that exists from prior work on the Waste Isolation Pilot Plant (WIPP), other U.S. repository development programs, and the work published through international efforts in salt repository programs such as in Germany. The potential benefits of developing a safety case include leveraging previous investments in WIPP to reduce future new repository costs, enhancing the ability to effectively plan for a repository and its licensing, and possibly expediting a schedule for a repository. A safety case will provide the necessary structure for organizing and synthesizing existing salt repository science and identifying any issues and gaps pertaining to safe disposal of DOE HLW and DOE SNF in bedded salt. The safety case synthesis will help DOE to plan its future R and D activities for investigating salt disposal using a risk-informed approach that prioritizes test activities that include laboratory, field, and underground investigations. It should be emphasized that the DOE has not made any decisions regarding the disposition of DOE HLW and DOE SNF. Furthermore, the safety case discussed herein is not intended to either site a repository in the Delaware Basin or preclude siting in other media at other locations. Rather, this study simply presents an approach for accelerated development of a safety case for a potential

  3. A Safety Case Approach for Deep Geologic Disposal of DOE HLW and DOE SNF in Bedded Salt - 13350

    International Nuclear Information System (INIS)

    Sevougian, S. David; MacKinnon, Robert J.; Leigh, Christi D.; Hansen, Frank D.

    2013-01-01

    The primary objective of this study is to investigate the feasibility and utility of developing a defensible safety case for disposal of United States Department of Energy (U.S. DOE) high-level waste (HLW) and DOE spent nuclear fuel (SNF) in a conceptual deep geologic repository that is assumed to be located in a bedded salt formation of the Delaware Basin [1]. A safety case is a formal compilation of evidence, analyses, and arguments that substantiate and demonstrate the safety of a proposed or conceptual repository. We conclude that a strong initial safety case for potential licensing can be readily compiled by capitalizing on the extensive technical basis that exists from prior work on the Waste Isolation Pilot Plant (WIPP), other U.S. repository development programs, and the work published through international efforts in salt repository programs such as in Germany. The potential benefits of developing a safety case include leveraging previous investments in WIPP to reduce future new repository costs, enhancing the ability to effectively plan for a repository and its licensing, and possibly expediting a schedule for a repository. A safety case will provide the necessary structure for organizing and synthesizing existing salt repository science and identifying any issues and gaps pertaining to safe disposal of DOE HLW and DOE SNF in bedded salt. The safety case synthesis will help DOE to plan its future R and D activities for investigating salt disposal using a risk-informed approach that prioritizes test activities that include laboratory, field, and underground investigations. It should be emphasized that the DOE has not made any decisions regarding the disposition of DOE HLW and DOE SNF. Furthermore, the safety case discussed herein is not intended to either site a repository in the Delaware Basin or preclude siting in other media at other locations. Rather, this study simply presents an approach for accelerated development of a safety case for a potential

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-26

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

  5. Methodology of fuel cycles long-term safety assessment of SNF/HLW geological disposal

    International Nuclear Information System (INIS)

    Pritrsky, J.

    2008-02-01

    Methodology for the long-term safety assessment of nuclear fuel cycles is given in the presented doctoral thesis. The aim of work was to develop a geological repository model for disposal of spent nuclear fuel (SNF) and high level waste (HLW) using an appropriate software code able to calculate the influence of partitioning and transmutation in advanced fuel cycles. The first step in this process was specifying of indicators which can be used to quantify the radiological impact of each fuel cycle. Indicators such as annual effective dose and radiotoxicity of inventory have been quantitatively analysed to determine the potential risk and radiological consequences associated with production of SNF/HLW. Advanced fuel types bring a number of advantages in comparison to uranium oxide fuel UO 2 used worldwide nowadays in terms of safety improvement due to minor actinides transmutation and non-proliferation aspects as well. Within the scope of work, three different fuel cycles are compared from the point of view of long-term safety of deep geological repository. The first considered fuel cycle is the currently used open fuel cycle (UOX) which uses only U-FA (Uranium Fuel Assembly). The second assessed cycle is a closed fuel cycle (MOX) with MOX-FA (Mixed OXides Fuel Assembly) and the third considered one is a partially closed fuel cycle (IMF) with IMC-FA (Inert Matrix Combined Fuel Assembly). Description and input data of advanced fuel cycles have been gained by participation in the EC project RED-IMPACT. Results were calculated using code AMBER, which is a flexible software tool that allows building dynamic compartmental models to represent the migration and fate of contaminants in a system, for example in the surface and sub-surface environment. Contaminants in solid, liquid and gaseous phases can be considered. AMBER gives the user the flexibility to define any number of compartments; any number of contaminants and associated decays; deterministic, probabilistic and

  6. Methodology of fuel cycles long-term safety assessment of SNF/HLW geological disposal

    International Nuclear Information System (INIS)

    Pritrsky, J.

    2008-01-01

    Methodology for the long-term safety assessment of nuclear fuel cycles is given in the presented doctoral thesis. The aim of work was to develop a geological repository model for disposal of spent nuclear fuel (SNF) and high level waste (HLW) using an appropriate software code able to calculate the influence of partitioning and transmutation in advanced fuel cycles. The first step in this process was specifying of indicators which can be used to quantify the radiological impact of each fuel cycle. Indicators such as annual effective dose and radiotoxicity of inventory have been quantitatively analysed to determine the potential risk and radiological consequences associated with production of SNF/HLW. Advanced fuel types bring a number of advantages in comparison to uranium oxide fuel UO 2 used worldwide nowadays in terms of safety improvement due to minor actinides transmutation and non-proliferation aspects as well. Within the scope of work, three different fuel cycles are compared from the point of view of long-term safety of deep geological repository. The first considered fuel cycle is the currently used open fuel cycle (UOX) which uses only U-FA (Uranium Fuel Assembly). The second assessed cycle is a closed fuel cycle (MOX) with MOX-FA (Mixed OXides Fuel Assembly) and the third considered one is a partially closed fuel cycle (IMF) with IMC-FA (Inert Matrix Combined Fuel Assembly). Description and input data of advanced fuel cycles have been gained by participation in the EC project RED-IMPACT. Results were calculated using code AMBER, which is a flexible software tool that allows building dynamic compartmental models to represent the migration and fate of contaminants in a system, for example in the surface and sub-surface environment. Contaminants in solid, liquid and gaseous phases can be considered. AMBER gives the user the flexibility to define any number of compartments; any number of contaminants and associated decays; deterministic, probabilistic and

  7. Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

    Energy Technology Data Exchange (ETDEWEB)

    PICKETT, W.W.

    2000-09-22

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. Because this sub-project is still in the construction/start-up phase, all verification activities have not yet been performed (e.g., canister cover cap and welding fixture system verification, MCO Internal Gas Sampling equipment verification, and As-built verification.). The verification activities identified in this report that still are to be performed will be added to the start-up punchlist and tracked to closure.

  8. Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

    International Nuclear Information System (INIS)

    PICKETT, W.W.

    2000-01-01

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. Because this sub-project is still in the construction/start-up phase, all verification activities have not yet been performed (e.g., canister cover cap and welding fixture system verification, MCO Internal Gas Sampling equipment verification, and As-built verification.). The verification activities identified in this report that still are to be performed will be added to the start-up punchlist and tracked to closure

  9. A randomized trial of heart failure disease management in skilled nursing facilities (SNF Connect): Lessons learned.

    Science.gov (United States)

    Daddato, Andrea; Wald, Heidi L; Horney, Carolyn; Fairclough, Diane L; Leister, Erin C; Coors, Marilyn; Capell, Warren H; Boxer, Rebecca S

    2017-06-01

    Conducting clinical trials in skilled nursing facilities is particularly challenging. This manuscript describes facility and patient recruitment challenges and solutions for clinical research in skilled nursing facilities. Lessons learned from the SNF Connect Trial, a randomized trial of a heart failure disease management versus usual care for patients with heart failure receiving post-acute care in skilled nursing facilities, are discussed. Description of the trial design and barriers to facility and patient recruitment along with regulatory issues are presented. The recruitment of Denver-metro skilled nursing facilities was facilitated by key stakeholders of the skilled nursing facilities community. However, there were still a number of barriers to facility recruitment including leadership turnover, varying policies regarding research, fear of litigation and of an increased workload. Engagement of facilities was facilitated by their strong interest in reducing hospital readmissions, marketing potential to hospitals, and heart failure management education for their staff. Recruitment of patients proved difficult and there were few facilitators. Identified patient recruitment challenges included patients being unaware of their heart failure diagnosis, patients overwhelmed with their illness and care, and frequently there was no available proxy for cognitively impaired patients. Flexibility in changing the recruitment approach and targeting skilled nursing facilities with higher rates of admissions helped to overcome some barriers. Recruitment of skilled nursing facilities and patients in skilled nursing facilities for clinical trials is challenging. Strategies to attract both facilities and patients are warranted. These include aligning study goals with facility incentives and flexible recruitment protocols to work with patients in "transition crisis."

  10. SNF1-related protein kinases 2 are negatively regulated by a plant-specific calcium sensor.

    Science.gov (United States)

    Bucholc, Maria; Ciesielski, Arkadiusz; Goch, Grażyna; Anielska-Mazur, Anna; Kulik, Anna; Krzywińska, Ewa; Dobrowolska, Grażyna

    2011-02-04

    SNF1-related protein kinases 2 (SnRK2s) are plant-specific enzymes involved in environmental stress signaling and abscisic acid-regulated plant development. Here, we report that SnRK2s interact with and are regulated by a plant-specific calcium-binding protein. We screened a Nicotiana plumbaginifolia Matchmaker cDNA library for proteins interacting with Nicotiana tabacum osmotic stress-activated protein kinase (NtOSAK), a member of the SnRK2 family. A putative EF-hand calcium-binding protein was identified as a molecular partner of NtOSAK. To determine whether the identified protein interacts only with NtOSAK or with other SnRK2s as well, we studied the interaction of an Arabidopsis thaliana orthologue of the calcium-binding protein with selected Arabidopsis SnRK2s using a two-hybrid system. All kinases studied interacted with the protein. The interactions were confirmed by bimolecular fluorescence complementation assay, indicating that the binding occurs in planta, exclusively in the cytoplasm. Calcium binding properties of the protein were analyzed by fluorescence spectroscopy using Tb(3+) as a spectroscopic probe. The calcium binding constant, determined by the protein fluorescence titration, was 2.5 ± 0.9 × 10(5) M(-1). The CD spectrum indicated that the secondary structure of the protein changes significantly in the presence of calcium, suggesting its possible function as a calcium sensor in plant cells. In vitro studies revealed that the activity of SnRK2 kinases analyzed is inhibited in a calcium-dependent manner by the identified calcium sensor, which we named SCS (SnRK2-interacting calcium sensor). Our results suggest that SCS is involved in response to abscisic acid during seed germination most probably by negative regulation of SnRK2s activity.

  11. DESIGN OF A CONCRETE SLAB FOR STORAGE OF SNF AND HLW CASKS

    International Nuclear Information System (INIS)

    J. Bisset

    2005-01-01

    This calculation documents the design of the Spent Nuclear Fuel (SNF) and High-Level Waste (HLW) Cask storage slab for the Aging Area. The design is based on the weights of casks that may be stored on the slab, the weights of vehicles that may be used to move the casks, and the layout shown on the sketch for a 1000 Metric Ton of Heavy Metal (MTHM) storage pad on Attachment 2, Sht.1 of the calculation 170-C0C-C000-00100-000-00A (BSC 2004a). The analytical model used herein is based on the storage area for 8 vertical casks. To simplify the model, the storage area of the horizontal concrete modules and their related shield walls is not included. The heavy weights of the vertical storage casks and the tensile forces due to pullout at the anchorages will produce design moments and shear forces that will envelope those that would occur in the storage area of the horizontal modules. The design loadings will also include snow and live loads. In addition, the design will also reflect pertinent geotechnical data. This calculation will document the preliminary thickness and general reinforcing steel requirements for the slab. This calculation also documents the initial design of the cask anchorage. Other slab details are not developed in this calculation. They will be developed during the final design process. The calculation also does not include the evaluation of the effects of cask drop loads. These will be evaluated in this or another calculation when the exact cask geometry is known

  12. Computational modeling of Repeat1 region of INI1/hSNF5: An evolutionary link with ubiquitin

    Science.gov (United States)

    Bhutoria, Savita

    2016-01-01

    Abstract The structure of a protein can be very informative of its function. However, determining protein structures experimentally can often be very challenging. Computational methods have been used successfully in modeling structures with sufficient accuracy. Here we have used computational tools to predict the structure of an evolutionarily conserved and functionally significant domain of Integrase interactor (INI)1/hSNF5 protein. INI1 is a component of the chromatin remodeling SWI/SNF complex, a tumor suppressor and is involved in many protein‐protein interactions. It belongs to SNF5 family of proteins that contain two conserved repeat (Rpt) domains. Rpt1 domain of INI1 binds to HIV‐1 Integrase, and acts as a dominant negative mutant to inhibit viral replication. Rpt1 domain also interacts with oncogene c‐MYC and modulates its transcriptional activity. We carried out an ab initio modeling of a segment of INI1 protein containing the Rpt1 domain. The structural model suggested the presence of a compact and well defined ββαα topology as core structure in the Rpt1 domain of INI1. This topology in Rpt1 was similar to PFU domain of Phospholipase A2 Activating Protein, PLAA. Interestingly, PFU domain shares similarity with Ubiquitin and has ubiquitin binding activity. Because of the structural similarity between Rpt1 domain of INI1 and PFU domain of PLAA, we propose that Rpt1 domain of INI1 may participate in ubiquitin recognition or binding with ubiquitin or ubiquitin related proteins. This modeling study may shed light on the mode of interactions of Rpt1 domain of INI1 and is likely to facilitate future functional studies of INI1. PMID:27261671

  13. 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

  14. Mcm1p binding sites in ARG1 positively regulate Gcn4p binding and SWI/SNF recruitment

    OpenAIRE

    Yoon, Sungpil; Hinnebusch, Alan G.

    2009-01-01

    Transcription of the arginine biosynthetic gene ARG1 is activated by Gcn4p, a transcription factor induced by starvation for any amino acid. Previously we showed that Gcn4p binding stimulates the recruitment of Mcm1p and co-activator SWI/SNF to ARG1 in cells via Gcn4p induction through amino acid starvation. Here we report that Gcn4p binding is reduced by point mutations of the Mcm1p binding site and increased by overexpression of Mcm1p. This result suggests that Mcm1p plays a positive role i...

  15. Status of Progress Made Toward Safety Analysis and Technical Site Evaluations for DOE Managed HLW and SNF.

    Energy Technology Data Exchange (ETDEWEB)

    Sevougian, S. David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Stein, Emily [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gross, Michael B [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hammond, Glenn Edward [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Frederick, Jennifer M [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Mariner, Paul [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-11-01

    The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) is conducting research and development (R&D) on generic deep geologic disposal systems (i.e., repositories). This report describes specific activities in FY 2016 associated with the development of a Defense Waste Repository (DWR)a for the permanent disposal of a portion of the HLW and SNF derived from national defense and research and development (R&D) activities of the DOE.

  16. HIV-1 replication in cell lines harboring INI1/hSNF5 mutations.

    Science.gov (United States)

    Sorin, Masha; Yung, Eric; Wu, Xuhong; Kalpana, Ganjam V

    2006-08-31

    INI1/hSNF5 is a cellular protein that directly interacts with HIV-1 integrase (IN). It is specifically incorporated into HIV-1 virions. A dominant negative mutant derived from INI1 inhibits HIV-1 replication. Recent studies indicate that INI1 is associated with pre-integration and reverse transcription complexes that are formed upon viral entry into the target cells. INI1 also is a tumor suppressor, biallelically deleted/mutated in malignant rhabdoid tumors. We have utilized cell lines derived from the rhabdoid tumors, MON and STA-WT1, that harbor either null or truncating mutations of INI1 respectively, to assess the effect of INI1 on HIV-1 replication. We found that while HIV-1 virions produced in 293T cells efficiently transduced MON and STA-WT1 cells, HIV-1 particle production was severely reduced in both of these cells. Reintroduction of INI1 into MON and STA-WT1 significantly enhanced the particle production in both cell lines. HIV-1 particles produced in MON cells were reduced for infectivity, while those produced in STA-WT1 were not. Further analysis indicated the presence of INI1 in those virions produced from STA-WT1 but not from those produced from MON cells. HIV-1 produced in MON cells were defective for synthesis of early and late reverse transcription products in the target cells. Furthermore, virions produced in MON cells were defective for exogenous reverse transcriptase activity carried out using exogenous template, primer and substrate. Our results suggest that INI1-deficient cells exhibit reduced particle production that can be partly enhanced by re-introduction of INI1. Infectivity of HIV-1 produced in some but not all INI1 defective cells, is affected and this defect may correlate to the lack of INI1 and/or some other proteins in these virions. The block in early events of virion produced from MON cells appears to be at the stage of reverse transcription. These studies suggest that presence of INI1 or some other host factor in virions and

  17. HIV-1 replication in cell lines harboring INI1/hSNF5 mutations

    Directory of Open Access Journals (Sweden)

    Wu Xuhong

    2006-08-01

    Full Text Available Abstract Background INI1/hSNF5 is a cellular protein that directly interacts with HIV-1 integrase (IN. It is specifically incorporated into HIV-1 virions. A dominant negative mutant derived from INI1 inhibits HIV-1 replication. Recent studies indicate that INI1 is associated with pre-integration and reverse transcription complexes that are formed upon viral entry into the target cells. INI1 also is a tumor suppressor, biallelically deleted/mutated in malignant rhabdoid tumors. We have utilized cell lines derived from the rhabdoid tumors, MON and STA-WT1, that harbor either null or truncating mutations of INI1 respectively, to assess the effect of INI1 on HIV-1 replication. Results We found that while HIV-1 virions produced in 293T cells efficiently transduced MON and STA-WT1 cells, HIV-1 particle production was severely reduced in both of these cells. Reintroduction of INI1 into MON and STA-WT1 significantly enhanced the particle production in both cell lines. HIV-1 particles produced in MON cells were reduced for infectivity, while those produced in STA-WT1 were not. Further analysis indicated the presence of INI1 in those virions produced from STA-WT1 but not from those produced from MON cells. HIV-1 produced in MON cells were defective for synthesis of early and late reverse transcription products in the target cells. Furthermore, virions produced in MON cells were defective for exogenous reverse transcriptase activity carried out using exogenous template, primer and substrate. Conclusion Our results suggest that INI1-deficient cells exhibit reduced particle production that can be partly enhanced by re-introduction of INI1. Infectivity of HIV-1 produced in some but not all INI1 defective cells, is affected and this defect may correlate to the lack of INI1 and/or some other proteins in these virions. The block in early events of virion produced from MON cells appears to be at the stage of reverse transcription. These studies suggest that

  18. SNF3 as high affinity glucose sensor and its function in supporting the viability of Candida glabrata under glucose-limited environment

    Directory of Open Access Journals (Sweden)

    Tzu Shan eNg

    2015-12-01

    Full Text Available Candida glabrata is an emerging human fungal pathogen that has efficacious nutrient sensing and responsiveness ability. It can be seen through its ability to thrive in diverse range of nutrient limited-human anatomical sites. Therefore, nutrient sensing particularly glucose sensing is thought to be crucial in contributing to the development and fitness of the pathogen. This study aimed to elucidate the role of SNF3 (Sucrose Non Fermenting 3 as a glucose sensor and its possible role in contributing to the fitness and survivability of C. glabrata in glucose-limited environment. The SNF3 knockout strain was constructed and subjected to different glucose concentrations to evaluate its growth, biofilm formation, amphotericin B susceptibility, ex vivo survivability and effects on the transcriptional profiling of the sugar receptor repressor (SRR pathway-related genes. The SNF3Δ strain showed a retarded growth in low glucose environments (0.01% and 0.1% in both fermentation and respiration-preferred conditions but grew well in high glucose concentration environments (1% and 2%. It was also found to be more susceptible to amphotericin B in low glucose environment (0.1% and macrophage engulfment but showed no difference in the biofilm formation capability. The deletion of SNF3 also resulted in the down-regulation of about half of hexose transporters genes (4 out of 9. Overall, the deletion of SNF3 causes significant reduction in the ability of C. glabrata to sense limited surrounding glucose and consequently disrupts its competency to transport and perform the uptake of this critical nutrient. This study highlighted the role of SNF3 as a high affinity glucose sensor and its role in aiding the survivability of C. glabrata particularly in glucose limited environment.

  19. CRITICALITY CALCULATION FOR THE MOST REACTIVE DEGRADED CONFIGURATIONS OF THE FFTF SNF CODISPOSAL WP CONTAINING AN INTACT IDENT-69 CONTAINER

    International Nuclear Information System (INIS)

    D.R. Moscalu

    2002-01-01

    The objective of this calculation is to perform additional degraded mode criticality evaluations of the Department of Energy's (DOE) Fast Flux Test Facility (FFTF) Spent Nuclear Fuel (SNF) codisposed in a 5-Defense High-Level Waste (5-DHLW) Waste Package (WP). The scope of this calculation is limited to the most reactive degraded configurations of the codisposal WP with an almost intact Ident-69 container (breached and flooded but otherwise non-degraded) containing intact FFTF SNF pins. The configurations have been identified in a previous analysis (CRWMS M andO 1999a) and the present evaluations include additional relevant information that was left out of the original calculations. The additional information describes the exact distribution of fissile material in each container (DOE 2002a). The effects of the changes that have been included in the baseline design of the codisposal WP (CRWMS M andO 2000) are also investigated. The calculation determines the effective neutron multiplication factor (k eff ) for selected degraded mode internal configurations of the codisposal waste package. These calculations will support the demonstration of the technical viability of the design solution adopted for disposing of MOX (FFTF) spent nuclear fuel in the potential repository. This calculation is subject to the Quality Assurance Requirements and Description (QARD) (DOE 2002b) per the activity evaluation under work package number P6212310M2 in the technical work plan TWP-MGR-MD-0000101 (BSC 2002)

  20. The SWI/SNF chromatin-remodeling factors BAF60a, b, and c in nutrient signaling and metabolic control

    Directory of Open Access Journals (Sweden)

    Ruo-Ran Wang

    2017-07-01

    Full Text Available ABSTRACT Metabolic syndrome has become a global epidemic that adversely affects human health. Both genetic and environmental factors contribute to the pathogenesis of metabolic disorders; however, the mechanisms that integrate these cues to regulate metabolic physiology and the development of metabolic disorders remain incompletely defined. Emerging evidence suggests that SWI/SNF chromatin-remodeling complexes are critical for directing metabolic reprogramming and adaptation in response to nutritional and other physiological signals. The ATP-dependent SWI/SNF chromatin-remodeling complexes comprise up to 11 subunits, among which the BAF60 subunit serves as a key link between the core complexes and specific transcriptional factors. The BAF60 subunit has three members, BAF60a, b, and c. The distinct tissue distribution patterns and regulatory mechanisms of BAF60 proteins confer each isoform with specialized functions in different metabolic cell types. In this review, we summarize the emerging roles and mechanisms of BAF60 proteins in the regulation of nutrient sensing and energy metabolism under physiological and disease conditions.

  1. 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...

  2. Transport of spent nuclear fuel from the High Flux Beam Reactor

    International Nuclear Information System (INIS)

    Holland, Michael; Carelli, Joseph; Shelton, Thomas

    1997-01-01

    The shipment of more than 1000 elements of spent nuclear fuel (SNF) from the Department of Energy's Brookhaven National Laboratory (BNL) High Flux Beam Reactor (HFBR) to the Department's Savannah River Site (SRS) for long term interim storage required overcoming several significant obstacles. The project management team was comprised of DOE, BNL and NAC International personnel. This achievement involved coordinating the efforts of numerous government and contractor organizations such as the U.S. Coast Guard, the U.S. Nuclear Regulatory Commission, state and local governments, marine and motor carriers, and carrier inspectors. Unique experience was gained during development and execution of the project in the following areas: dry transfer of SNF to shipping casks; inter-modal transfers; logistics; cask licensing by the Nuclear Regulatory Commission (NRC); compliance with environmental regulations; transportation plan development, and stakeholder outreach and coordination

  3. The metal and concrete cask for SNF and its radiation protection quality control

    International Nuclear Information System (INIS)

    Shchigolev, N.D.; Golubev, O.M.

    2005-01-01

    The transportation and packing module on the basis of metal and concrete cask for the long-term storage and shipment of the spent nuclear fuel is developed in compliance with the requirements of the national standards and IAEA recommendations. Such wares designed for the NPP and submarine reactors fuel may be remade also for the research ones. A procedure and remote device for the radiation protection control of this cask equally its integrity checks after dynamic testing also is described. (author)

  4. Thermal Analysis of Cold Vacuum Drying (CVD) of Spent Nuclear Fuel (SNF)

    International Nuclear Information System (INIS)

    PIEPHO, M.G.

    2000-01-01

    The thermal analysis examined transient thermal and chemical behavior of the Multi-Canister Overpack (MCO) container for a broad range of cases that represent the Cold Vacuum Drying (CVD) processes. The cases were defined to consider both normal and off-normal operations at the CVD Facility for an MCO with N Reactor spent fuel. This analysis provides the basis for the MCO thermal behavior at the CVD Facility in support of the safety basis documentation

  5. National spent fuel program preliminary report RCRA characteristics of DOE-owned spent nuclear fuel DOE-SNF-REP-002. Revision 3

    International Nuclear Information System (INIS)

    1995-07-01

    This report presents information on the preliminary process knowledge to be used in characterizing all Department of Energy (DOE)-owned Spent Nuclear Fuel (SNF) types that potentially exhibit a Resource Conservation and Recovery Act (RCRA) characteristic. This report also includes the process knowledge, analyses, and rationale used to preliminarily exclude certain SNF types from RCRA regulation under 40 CFR section 261.4(a)(4), ''Identification and Listing of Hazardous Waste,'' as special nuclear and byproduct material. The evaluations and analyses detailed herein have been undertaken as a proactive approach. In the event that DOE-owned SNF is determined to be a RCRA solid waste, this report provides general direction for each site regarding further characterization efforts. The intent of this report is also to define the path forward to be taken for further evaluation of specific SNF types and a recommended position to be negotiated and established with regional and state regulators throughout the DOE Complex regarding the RCRA-related policy issues

  6. Spent Nuclear Fuel (SNF) Project Cask and MCO Helium Purge System Design Review Completion Report - Project A.5 and A.6

    International Nuclear Information System (INIS)

    ARD, K.E.

    2000-01-01

    This report documents the results of the design verification performed on the Cask and Multiple Canister Over-pack (MCO) Helium Purge System. The helium purge system is part of the Spent Nuclear Fuel (SNF) Project Cask Loadout System (CLS) at 100K area. The design verification employed the ''Independent Review Method'' in accordance with Administrative Procedure (AP) EN-6-027-01

  7. National spent fuel program preliminary report RCRA characteristics of DOE-owned spent nuclear fuel DOE-SNF-REP-002. Revision 3

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    This report presents information on the preliminary process knowledge to be used in characterizing all Department of Energy (DOE)-owned Spent Nuclear Fuel (SNF) types that potentially exhibit a Resource Conservation and Recovery Act (RCRA) characteristic. This report also includes the process knowledge, analyses, and rationale used to preliminarily exclude certain SNF types from RCRA regulation under 40 CFR {section}261.4(a)(4), ``Identification and Listing of Hazardous Waste,`` as special nuclear and byproduct material. The evaluations and analyses detailed herein have been undertaken as a proactive approach. In the event that DOE-owned SNF is determined to be a RCRA solid waste, this report provides general direction for each site regarding further characterization efforts. The intent of this report is also to define the path forward to be taken for further evaluation of specific SNF types and a recommended position to be negotiated and established with regional and state regulators throughout the DOE Complex regarding the RCRA-related policy issues.

  8. Rhabdoid and Undifferentiated Phenotype in Renal Cell Carcinoma: Analysis of 32 Cases Indicating a Distinctive Common Pathway of Dedifferentiation Frequently Associated With SWI/SNF Complex Deficiency.

    Science.gov (United States)

    Agaimy, Abbas; Cheng, Liang; Egevad, Lars; Feyerabend, Bernd; Hes, Ondřej; Keck, Bastian; Pizzolitto, Stefano; Sioletic, Stefano; Wullich, Bernd; Hartmann, Arndt

    2017-02-01

    Undifferentiated (anaplastic) and rhabdoid cell features are increasingly recognized as adverse prognostic findings in renal cell carcinoma (RCC), but their molecular pathogenesis has not been studied sufficiently. Recent studies identified alterations in the Switch Sucrose nonfermentable (SWI/SNF) chromatin remodeling complex as molecular mechanisms underlying dedifferentiation and rhabdoid features in carcinomas of different organs. We herein have analyzed 32 undifferentiated RCCs having in common an undifferentiated (anaplastic) phenotype, prominent rhabdoid features, or both, irrespective of the presence or absence of conventional RCC component. Cases were stained with 6 SWI/SNF pathway members (SMARCB1, SMARCA2, SMARCA4, ARID1A, SMARCC1, and SMARCC2) in addition to conventional RCC markers. Patients were 20 males and 12 females aged 32 to 85 years (mean, 59). A total of 22/27 patients with known stage presented with ≥pT3. A differentiated component varying from microscopic to major component was detected in 20/32 cases (16 clear cell and 2 cases each chromophobe and papillary RCC). The undifferentiated component varied from rhabdoid dyscohesive cells to large epithelioid to small monotonous anaplastic cells. Variable loss of at least 1 SWI/SNF complex subunit was noted in the undifferentiated/rhabdoid component of 21/32 cases (65%) compared with intact or reduced expression in the differentiated component. A total of 15/17 patients (88%) with follow-up died of metastatic disease (mostly within 1 y). Only 2 patients were disease free at last follow-up (1 and 6 y). No difference in survival, age distribution, or sex was observed between the SWI/SNF-deficient and the SWI/SNF-intact group. This is the first study exploring the role of SWI/SNF deficiency as a potential mechanism underlying undifferentiated and rhabdoid phenotype in RCC. Our results highlight the association between the aggressive rhabdoid phenotype and the SWI/SNF complex deficiency, consistent

  9. German research reactor back-end provisions

    International Nuclear Information System (INIS)

    Koester, Siegfried; Gruber, Gerhard

    2002-01-01

    Germany has several types of Research Reactors in operation. These reactors use fuel containing uranium of U.S. origin. Basically all the fuel which will be spent until May 2006 will be returned to the U.S. under existing contracts with the U.S. Department of Energy. The contracts are based on the U.S. FRR SNF (Foreign Research Reactor Spent Nuclear Fuel) Program which started in May 1996 and which will last for 10 years. In 1990, the German Federal Government started a program to long-term store (approx. 40 years) and finally dispose of spent fuel in Germany after the so-called U.S. fuel return window will be closed. In order to long-term store the fuel, a special container was designed which covers all different types of spent fuel from the Research Reactors. The container called 'CASTOR MTR 2' is basically licensed and is already in use for the spent fuel of Russian origin from the 'Research Reactor Rossendorf' in the eastern part of Germany. All that fuel is expected to be stored in the existing intermediate storage facility, the so-called BZA (Brennelemente Zwischenlager Ahaus). BZA already accomodates spent fuel from the former THTR-300 high temperature reactor. A final repository does not yet exist in Germany. Alternative provisions to close the back-end of the Research Reactor fuel cycle are reprocessing at COGEMA (France) or in Russian facilities, perspectively. Waste return in a form to be agreed will be mandatory, at least in France. (author)

  10. STRUCTURAL CALCULATIONS FOR THE LIFTING IN VERTICAL ORIENTATION OF 5-DHLW/DOE SNF SINGLE CRM WASTE PACKAGES

    International Nuclear Information System (INIS)

    S. Mastilovic

    1999-01-01

    The purpose of this activity is to determine the structural response of the extension of outer shell (which is referred to as skirt throughout this document) designs of both long and short design concepts of 5-Defense High-Level Waste (DHLW) Department of Energy (DOE) spent nuclear fuel (SNF) single corrosion resistant material (CRM) waste packages (WP), subjected to a gravitational load in the course of lifting in vertical orientation. The scope of this document is limited to reporting the calculation results in terms of stress intensity magnitudes. This activity is associated with the WP design; calculations are performed by the Waste Package Design group. AP-3.124, Revision 0, ICN 0, Calculations, is used to perform the calculation and develop the document

  11. Regional Geologic Evaluations for Disposal of HLW and SNF: The Pierre Shale of the Northern Great Plains

    Energy Technology Data Exchange (ETDEWEB)

    Perry, Frank Vinton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kelley, Richard E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-09-14

    The DOE Spent Fuel and Waste Technology (SWFT) R&D Campaign is supporting research on crystalline rock, shale (argillite) and salt as potential host rocks for disposal of HLW and SNF in a mined geologic repository. The distribution of these three potential repository host rocks is limited to specific regions of the US and to different geologic and hydrologic environments (Perry et al., 2014), many of which may be technically suitable as a site for mined geologic disposal. This report documents a regional geologic evaluation of the Pierre Shale, as an example of evaluating a potentially suitable shale for siting a geologic HLW repository. This report follows a similar report competed in 2016 on a regional evaluation of crystalline rock that focused on the Superior Province of the north-central US (Perry et al., 2016).

  12. Canisters and nonfuel components at commercial nuclear reactors

    International Nuclear Information System (INIS)

    Gibbard, K.; Disbrow, J.

    1994-01-01

    This paper discusses detailed data on canisters and nonfuel components (NFC) at US commercial nuclear power reactors. A wide variety of NFC have been reported on the Form RW-859, open-quotes Nuclear Fuel Dataclose quotes survey. They may have been integral with an assembly, noncanistered in baskets, destined for disposal as low-level radioactive waste, or stored in canisters. Similarly, data on the family of canistered spent nuclear fuel (SNF) in storage pools was compiled. Approximately 85 percent of the 40,194 pieces of nonfuel assembly (NFA) hardware reported were integral with an assembly. This represents data submitted by 95 of the 107 reactors in 10 generic assembly classes. In addition, a total of 286 canisters have been reported as being in storage pools as of December 31, 1992. However, an additional 264 open baskets were also reported to contain miscellaneous SNF and nonfuel materials, garbage and debris. All of these 286 canisters meet the dimensional envelope requirements specified for disposal for open-quotes standard fuelclose quotes under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (10 CFR 961); most of the baskets do not

  13. Solution NMR structure of the HLTF HIRAN domain: a conserved module in SWI2/SNF2 DNA damage tolerance proteins

    International Nuclear Information System (INIS)

    Korzhnev, Dmitry M.; Neculai, Dante; Dhe-Paganon, Sirano; Arrowsmith, Cheryl H.; Bezsonova, Irina

    2016-01-01

    HLTF is a SWI2/SNF2-family ATP-dependent chromatin remodeling enzyme that acts in the error-free branch of DNA damage tolerance (DDT), a cellular mechanism that enables replication of damaged DNA while leaving damage repair for a later time. Human HLTF and a closely related protein SHPRH, as well as their yeast homologue Rad5, are multi-functional enzymes that share E3 ubiquitin-ligase activity required for activation of the error-free DDT. HLTF and Rad5 also function as ATP-dependent dsDNA translocases and possess replication fork reversal activities. Thus, they can convert Y-shaped replication forks into X-shaped Holliday junction structures that allow error-free replication over DNA lesions. The fork reversal activity of HLTF is dependent on 3′-ssDNA-end binding activity of its N-terminal HIRAN domain. Here we present the solution NMR structure of the human HLTF HIRAN domain, an OB-like fold module found in organisms from bacteria (as a stand-alone domain) to plants, fungi and metazoan (in combination with SWI2/SNF2 helicase-like domain). The obtained structure of free HLTF HIRAN is similar to recently reported structures of its DNA bound form, while the NMR analysis also reveals that the DNA binding site of the free domain exhibits conformational heterogeneity. Sequence comparison of N-terminal regions of HLTF, SHPRH and Rad5 aided by knowledge of the HLTF HIRAN structure suggests that the SHPRH N-terminus also includes an uncharacterized structured module, exhibiting weak sequence similarity with HIRAN regions of HLTF and Rad5, and potentially playing a similar functional role.

  14. Solution NMR structure of the HLTF HIRAN domain: a conserved module in SWI2/SNF2 DNA damage tolerance proteins

    Energy Technology Data Exchange (ETDEWEB)

    Korzhnev, Dmitry M. [University of Connecticut Health, Department of Molecular Biology and Biophysics (United States); Neculai, Dante [Zhejiang University, School of Medicine (China); Dhe-Paganon, Sirano [Dana-Farber Cancer Institute, Department of Cancer Biology (United States); Arrowsmith, Cheryl H. [University of Toronto, Structural Genomics Consortium (Canada); Bezsonova, Irina, E-mail: bezsonova@uchc.edu [University of Connecticut Health, Department of Molecular Biology and Biophysics (United States)

    2016-11-15

    HLTF is a SWI2/SNF2-family ATP-dependent chromatin remodeling enzyme that acts in the error-free branch of DNA damage tolerance (DDT), a cellular mechanism that enables replication of damaged DNA while leaving damage repair for a later time. Human HLTF and a closely related protein SHPRH, as well as their yeast homologue Rad5, are multi-functional enzymes that share E3 ubiquitin-ligase activity required for activation of the error-free DDT. HLTF and Rad5 also function as ATP-dependent dsDNA translocases and possess replication fork reversal activities. Thus, they can convert Y-shaped replication forks into X-shaped Holliday junction structures that allow error-free replication over DNA lesions. The fork reversal activity of HLTF is dependent on 3′-ssDNA-end binding activity of its N-terminal HIRAN domain. Here we present the solution NMR structure of the human HLTF HIRAN domain, an OB-like fold module found in organisms from bacteria (as a stand-alone domain) to plants, fungi and metazoan (in combination with SWI2/SNF2 helicase-like domain). The obtained structure of free HLTF HIRAN is similar to recently reported structures of its DNA bound form, while the NMR analysis also reveals that the DNA binding site of the free domain exhibits conformational heterogeneity. Sequence comparison of N-terminal regions of HLTF, SHPRH and Rad5 aided by knowledge of the HLTF HIRAN structure suggests that the SHPRH N-terminus also includes an uncharacterized structured module, exhibiting weak sequence similarity with HIRAN regions of HLTF and Rad5, and potentially playing a similar functional role.

  15. Life-cycle cost implications of a system using bare SNF transfer

    International Nuclear Information System (INIS)

    Rose, M.

    1995-01-01

    The U.S. Department of Energy (DOE) is planning the use of the Multi-Purpose Canister (MPC) to handle transportation, storage, and disposal of civilian spent nuclear fuel. This canister, which would be loaded and sealed at the utility site, would remain sealed through waste emplacement in a Mined Geologic Disposal System (MGDS). Two sizes of MPCs are now being considered: large and small rail. The large rail canister has design requirements to be compatible with a 125 ton hook weight crane when fully loaded in the pool with fuel assemblies, water, transportation overpack, and the lifting yoke. The small rail canister under the same conditions weighs less than 75 tons. At present, it is estimated that between four and seventeen reactors will not be able to accommodate either canister. One method of accommodating MPCs at all reactors is the use of Bare Spent nuclear fuel Transfer (BST). In this concept, a small transfer cask is used to move small numbers of assemblies from the spent fuel pool to an external transfer station where a large MPC is loaded. After several of these transfers, the MPC is sealed and either stored on-site, transported to a central storage site, or transported to the MGDS. This paper addresses the total system cost implications of use of BST at 20 sites (31 reactors) which are currently projected to be unable to use the large rail MPC. Results are presented parametrically as a function of the BST capital cost and the time required to load a MPC. This analysis indicates use of BST may be economically favorable if the combination of MPC load times and capital expenditures can be kept to a reasonable level

  16. Neutronic and Logistic Proposal for Transmutation of Plutonium from Spent Nuclear Fuel as Mixed-Oxide Fuel in Existing Light Water Reactors

    International Nuclear Information System (INIS)

    Trellue, Holly R.

    2004-01-01

    The use of light water reactors (LWRs) for the destruction of plutonium and other actinides [especially those in spent nuclear fuel (SNF)] is being examined worldwide. One possibility for transmutation of this material is the use of mixed-oxide (MOX) fuel, which is a combination of uranium and plutonium oxides. MOX fuel is used in nuclear reactors worldwide, so a large experience base for its use already exists. However, to limit implementation of SNF transmutation to only a fraction of the LWRs in the United States with a reasonable number of license extensions, full cores of MOX fuel probably are required. This paper addresses the logistics associated with using LWRs for this mission and the design issues required for full cores of MOX fuel. Given limited design modifications, this paper shows that neutronic safety conditions can be met for full cores of MOX fuel with up to 8.3 wt% of plutonium

  17. 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

  18. 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.

  19. 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.

  20. 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

  1. A continuing success - The United States Foreign Research Reactor Spent Nuclear Fuel Acceptance Program

    International Nuclear Information System (INIS)

    Mustin, Tracy P.; Clapper, Maureen; Reilly, Jill E.

    2000-01-01

    The United States Department of Energy, in consultation with the Department of State, adopted the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel in May 1996. To date, the Foreign Research Reactor (FRR) Spent Nuclear Fuel (SNF) Acceptance Program, established under this policy, has completed 16 spent fuel shipments. 2,651 material test reactor (MTR) assemblies, one Slowpoke core containing less than 1 kilogram of U.S.-origin enriched uranium, 824 Training, Research, Isotope, General Atomic (TRIGA) rods, and 267 TRIGA pins from research reactors around the world have been shipped to the United States so far under this program. As the FRR SNF Acceptance Program progresses into the fifth year of implementation, a second U.S. cross country shipment has been completed, as well as a second overland truck shipment from Canada. Both the cross country shipment and the Canadian shipment were safely and successfully completed, increasing our knowledge and experience in these types of shipments. In addition, two other shipments were completed since last year's RERTR meeting. Other program activities since the last meeting included: taking pre-emptive steps to avoid license amendment pitfalls/showstoppers for spent fuel casks, publication of a revision to the Record of Decision allowing up to 16 casks per ocean going vessel, and the issuance of a cable to 16 of the 41 eligible countries reminding their governments and the reactor operators that the U.S.-origin uranium in their research reactors may be eligible for return to the United States under the Acceptance Program and urging them to begin discussions on shipping schedules. The FRR SNF program has also supported the Department's implementation of the competitive pricing policy for uranium and resumption of shipments of fresh uranium for fabrication into assemblies for research reactors. The United States Foreign Research Reactor Spent Nuclear Fuel Acceptance Program continues

  2. 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

  3. 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

  4. Transportation of failed or damaged foreign research reactor spent nuclear fuel

    International Nuclear Information System (INIS)

    Messick, Charles E.; Mustin, Tracy P.; Massey, Charles D.

    1999-01-01

    Since initiating the Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance Program in 1996, the Program has had to deal with difficult issues associated with the transportation of failed or damaged spent fuel. In several instances, problems with failed or damaged fuel have prevented the acceptance of the fuel at considerable cost to both the Department of Energy and research reactor operators. In response to the problems faced by the Acceptance Program, DOE has undertaken significant steps to better define the spent fuel acceptance criteria. DOE has worked closely with the U.S. Nuclear Regulatory Commission to address failed or damaged research reactor spent fuel causing a degradation of the fuel assembly exposing fuel meat and to identify cask certificate issues which must be resolved by cask owners and foreign regulatory authorities. The specific issues and implementation challenges associated with the transport of MTR type FRR SNF will be discussed. The information presented will include U.S. Nuclear Regulatory Commission regulatory issues, cask certificate issues, technical constraints, implementation status, and lessons learned. Specific information will also be provided on the latest efforts to revise DOE's Appendix B, Transport Package (Cask) Acceptance Criteria. The information presented in this paper will be of interest to foreign research reactor operators, shippers, and cask vendors in evaluating the condition of their fuel to ensure it can be transported in accordance with appropriate cask certificate requirements. (author)

  5. Impacts of SNF burnup credit on the shipment capability of the GA-4 cask

    International Nuclear Information System (INIS)

    Mobasheran, A.S.; Lake, W.; Richardson, J.

    1996-01-01

    Scoping analyses were performed to determine the impacts of two different levels of burnup credit and two different spent fuel pickup rates on the shipment capability and the minimum fleet size of the GA-4 cask. The analyses involved developing loading curves for the GA-4 cask based on the actinide-only and principal-isotope burnup credit considerations. The analyses also involved examination of the spent nuclear fuel assembly population at nine reactor sites and categorization of the assemblies in accordance with the loading restrictions imposed. The results revealed that for the nine sites considered, depending on the level of burnup credit and the pickup rate assumed, the total savings in shipment and cask fleet costs (1994 dollars) can range from $55 million to $74 million

  6. 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.)

  7. 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.)

  8. 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

  9. 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

  10. 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.)

  11. 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.)

  12. 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)

  13. 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

  14. Divergent Evolution of the Transcriptional Network Controlled by Snf1-Interacting Protein Sip4 in Budding Yeasts.

    Directory of Open Access Journals (Sweden)

    Constance Mehlgarten

    Full Text Available Cellular responses to starvation are of ancient origin since nutrient limitation has always been a common challenge to the stability of living systems. Hence, signaling molecules involved in sensing or transducing information about limiting metabolites are highly conserved, whereas transcription factors and the genes they regulate have diverged. In eukaryotes the AMP-activated protein kinase (AMPK functions as a central regulator of cellular energy homeostasis. The yeast AMPK ortholog SNF1 controls the transcriptional network that counteracts carbon starvation conditions by regulating a set of transcription factors. Among those Cat8 and Sip4 have overlapping DNA-binding specificity for so-called carbon source responsive elements and induce target genes upon SNF1 activation. To analyze the evolution of the Cat8-Sip4 controlled transcriptional network we have compared the response to carbon limitation of Saccharomyces cerevisiae to that of Kluyveromyces lactis. In high glucose, S. cerevisiae displays tumor cell-like aerobic fermentation and repression of respiration (Crabtree-positive while K. lactis has a respiratory-fermentative life-style, respiration being regulated by oxygen availability (Crabtree-negative, which is typical for many yeasts and for differentiated higher cells. We demonstrate divergent evolution of the Cat8-Sip4 network and present evidence that a role of Sip4 in controlling anabolic metabolism has been lost in the Saccharomyces lineage. We find that in K. lactis, but not in S. cerevisiae, the Sip4 protein plays an essential role in C2 carbon assimilation including induction of the glyoxylate cycle and the carnitine shuttle genes. Induction of KlSIP4 gene expression by KlCat8 is essential under these growth conditions and a primary function of KlCat8. Both KlCat8 and KlSip4 are involved in the regulation of lactose metabolism in K. lactis. In chromatin-immunoprecipitation experiments we demonstrate binding of both, KlSip4 and

  15. Research reactors

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

    It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

  16. Mediator, SWI/SNF and SAGA complexes regulate Yap8-dependent transcriptional activation of ACR2 in response to arsenate.

    Science.gov (United States)

    Menezes, Regina Andrade; Pimentel, Catarina; Silva, Ana Rita Courelas; Amaral, Catarina; Merhej, Jawad; Devaux, Frédéric; Rodrigues-Pousada, Claudina

    2017-04-01

    Response to arsenic stress in Saccharomyces cerevisiae is orchestrated by the regulatory protein Yap8, which mediates transcriptional activation of ACR2 and ACR3. This study contributes to the state of art knowledge of the molecular mechanisms underlying yeast stress response to arsenate as it provides the genetic and biochemical evidences that Yap8, through cysteine residues 132, 137, and 274, is the sensor of presence of arsenate in the cytosol. Moreover, it is here reported for the first time the essential role of the Mediator complex in the transcriptional activation of ACR2 by Yap8. Based on our data, we propose an order-of-function map to recapitulate the sequence of events taking place in cells injured with arsenate. Modification of the sulfhydryl state of these cysteines converts Yap8 in its activated form, triggering the recruitment of the Mediator complex to the ACR2/ACR3 promoter, through the interaction with the tail subunit Med2. The Mediator complex then transfers the regulatory signals conveyed by Yap8 to the core transcriptional machinery, which culminates with TBP occupancy, ACR2 upregulation and cell adaptation to arsenate stress. Additional co-factors are required for the transcriptional activation of ACR2 by Yap8, particularly the nucleosome remodeling activity of SWI/SNF and SAGA complexes. Copyright © 2017. Published by Elsevier B.V.

  17. Mechanisms involved in the p62-73 idiopeptide-modulated delay of lupus nephritis in SNF(1) mice.

    Science.gov (United States)

    Nyland, J F; Stoll, M L; Jiang, F; Feng, F; Gavalchin, J

    2012-12-01

    The F(1) progeny of the (SWR × NZB) cross develop a lupus-like disease with high serum titers of autoantibodies, and increased frequency and severity of immune complex-mediated glomerulonephritis in females. In previous work, we found that an idiotypic peptide corresponding to aa62-73 (p62-73) of the heavy chain variable region of autoantibody 540 (Id(LN)F(1)) induced the proliferation of p62-73 idiotype-reactive T cell clones. Further, monthly immunization of pre-nephritic SNF(1) female mice with p62-73 resulted in decreased nephritis and prolonged life spans. Here we show that this treatment modulated proliferative responses to Id(LN)F(1) antigen, including a reduction in the population of idiopeptide-presenting antigen-presenting cells (APCs), as early as two weeks after immunization (10 weeks of age). Th1-type cytokine production was increased at 12 weeks of age. The incidence and severity of nephritis was reduced by 14 weeks compared to controls. Clinical indicators of nephritis, specifically histological evidence of glomerulonephritis and urine protein levels, were reduced by 20 weeks. Together these data suggest that events involved in the mechanism(s) whereby p62-73 immunization delayed nephritis occurred early after immunization, and involved modulation of APCs, B and T cell populations.

  18. The risk-rewards structure of using spent nuclear fuel in molten salt reactor - 5513

    International Nuclear Information System (INIS)

    He, X.; Du, Z.; Macian-Juan, R.; Seidl, M.

    2015-01-01

    The molten salt reactor concept naturally lends itself to a re-use of fuel either by online reprocessing or by using spent nuclear fuel as part of the driver fuel. Moreover some well-known safety advantages over traditional LWR designs are promised: the primary loop can be operated at atmospheric pressure, refueling can be done online, only a minimum amount of excess reactivity needs to be stored inside the core and the continuous circulation and inter-mixing of the fuel results in a more homogenous redistribution of fission products. In this paper the feasibility of running a molten salt reactor on spent LWR fuel is discussed in a number of scenarios in order to make the various trade-offs transparent: using SNF in a classic graphite moderated MSR and doing the same for a lead-cooled dual-fluid MSR. From a commercial company's point of view the MSR concept faces already substantial risks even without the use of SNF: licensing concerns due to an enrichment of fissile nuclides typically above 5% of heavy metal mass, limited practical experience with the reliability of proposed MSR materials and almost no experience with online reprocessing. For one thing one could therefore aim for the most conservative design which would rely on the design of ORNL's graphite moderated MSR operated in the sixties. While appearing realistic from a technical perspective, the potential for SNF re-use in the sense of actinide destruction appears limited. On the other hand one can maximize the risk and the potential payoff by concentrating on the most speculative design, i.e. a dual fluid reactor with an ultra-hard neutron spectrum in order to most efficiently burn higher actinides. In this paper the neutronic design calculations for the above described MSR concepts are presented in order to maximize SNF's contribution for the reactors' energy generation and their potential for actinide destruction. Among the optimization parameters are the lattice constants, the type

  19. Achieving the timely receipt of foreign research reactor spent nuclear fuel at the Savannah River site

    International Nuclear Information System (INIS)

    Brizes, C.M.; Clark, W.D; Thomas, J.; Andes, T.

    1998-01-01

    The May 1996 Record of Decision on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel states that the United States will accept spent nuclear fuel containing uranium of U.S.-origin from foreign research reactors through the year 2009. The best information available indicates that approximately 13,000 assemblies of Material Test Reactor (MTR) spent nuclear fuel from 29 countries are expected to be shipped to the Savannah River Site during the 13 years of the program. As of July 1998, 1,371 spent nuclear fuel assemblies from 12 foreign research reactors have been received at the SRS. That is, after more than two years of the FRR program (approximately 15 percent of the program time), 11 percent of the total assemblies have been received at SRS. Current projections show that most of the assemblies can be received by 2009, however if some of the eligible, non-participating countries decide to rejoin the program, a bottleneck would occur at the end of the program. Also adding to the potential for the bottleneck is a trend of shipments being moved out in the timeline. The Savannah River Site is working to be proactive in avoiding a bottleneck at the end of the program, but cooperation is required from all program participants to be successful. Activities currently in progress include inventory/information questionnaires, verifying fuel against cask(s) certificate of compliance (C. of C.), and collecting Appendix A information well in advance of shipping the SNF. The inventory/information sheets have been distributed to a select number of reactor facilities in the past, but work is in progress to refine the process. Information requested in the questionnaire includes inventory numbers, preferred shipping dates, and cask preferences. This information allows for improved shipment planning and helps to ensure that we are working to meet the needs of the reactor facilities. Current plans are to send the questionnaires to

  20. Reactor container

    International Nuclear Information System (INIS)

    Naruse, Yoshihiro.

    1990-01-01

    The thickness of steel shell plates in a reactor container embedded in sand cussions is monitored to recognize the corrosion of the steel shell plates. That is, the reactor pressure vessel is contained in a reactor container shell and the sand cussions are disposed on the lower outside of the reactor container shell to elastically support the shell. A pit is disposed at a position opposing to the sand cussions for measuring the thickness of the reactor container shell plates. The pit is usually closed by a closing member. In the reactor container thus constituted, the closing member can be removed upon periodical inspection to measure the thickness of the shell plates. Accordingly, the corrosion of the steel shell plates can be recognized by the change of the plate thickness. (I.S.)

  1. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

  2. 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

  3. Nuclear reactor

    International Nuclear Information System (INIS)

    Batheja, P.; Huber, R.; Rau, P.

    1985-01-01

    Particularly for nuclear reactors of small output, the reactor pressure vessel contains at least two heat exchangers, which have coolant flowing through them in a circuit through the reactor core. The circuit of at least one heat exchanger is controlled by a slide valve, so that even for low drive forces, particularly in natural circulation, the required even loading of the heat exchanger is possible. (orig./HP) [de

  4. Blue Ribbon Commission, Yucca Mountain Closure, Court Actions - Future of Decommissioned Reactors, Operating Reactors and Nuclear Power - 13249

    International Nuclear Information System (INIS)

    Devgun, Jas S.

    2013-01-01

    Issues related to back-end of the nuclear fuel cycle continue to be difficult for the commercial nuclear power industry and for the decision makers at the national and international level. In the US, the 1982 NWPA required DOE to develop geological repositories for SNF and HLW but in spite of extensive site characterization efforts and over ten billion dollars spent, a repository opening is nowhere in sight. There has been constant litigation against the DOE by the nuclear utilities for breach of the 'standard contract' they signed with the DOE under the NWPA. The SNF inventory continues to rise both in the US and globally and the nuclear industry has turned to dry storage facilities at reactor locations. In US, the Blue Ribbon Commission on America's Nuclear Future issued its report in January 2012 and among other items, it recommends a new, consent-based approach to siting of facilities, prompt efforts to develop one or more geologic disposal facilities, and prompt efforts to develop one or more consolidated storage facilities. In addition, the March 2011 Fukushima Daiichi accident had a severe impact on the future growth of nuclear power. The nuclear industry is focusing on mitigation strategies for beyond design basis events and in the US, the industry is in the process of implementing the recommendations from NRC's Near Term Task Force. (authors)

  5. Heterogeneous reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de; Nair, R.P.K.

    1979-08-01

    The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author) [pt

  6. Minor Actinide Recycle in Sodium Cooled Fast Reactors Using Heterogeneous Targets

    International Nuclear Information System (INIS)

    Bays, Samuel; Medvedev, Pavel; Pope, Michael; Ferrer, Rodolfo; Forget, Benoit; Asgari, Mehdi

    2009-01-01

    This paper investigates the plausible design of transmutation target assemblies for minor actinides (MA) in Sodium Fast Reactors (SFR). A heterogeneous recycling strategy is investigated, whereby after each reactor pass, un-burned MAs from the targets are blended with MAs produced by the driver fuel and additional MAs from Spent Nuclear Fuel (SNF). A design iteration methodology was adopted for customizing the core design, target assembly design and matrix composition design. The overall design was constrained against allowable peak or maximum in-core performances. While respecting these criteria, the overall design was adjusted to reduce the total number of assemblies fabricated per refueling cycle. It was found that an inert metal-hydride MA-Zr-Hx target matrix gave the highest transmutation efficiency, thus allowing for the least number of targets to be fabricated per reactor cycle.

  7. Slurry reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kuerten, H; Zehner, P [BASF A.G., Ludwigshafen am Rhein (Germany, F.R.)

    1979-08-01

    Slurry reactors are designed on the basis of empirical data and model investigations. It is as yet not possible to calculate the flow behavior of such reactors. The swarm of gas bubbles and cluster formations of solid particles and their interaction in industrial reactors are not known. These effects control to a large extent the gas hold-up, the gas-liquid interface and, similarly as in bubble columns, the back-mixing of liquids and solids. These hydrodynamic problems are illustrated in slurry reactors which constructionally may be bubble columns, stirred tanks or jet loop reactors. The expected effects are predicted by means of tests with model systems modified to represent the conditions in industrial hydrogenation reactors. In his book 'Mass Transfer in Heterogeneous Catalysis' (1970) Satterfield complained of the lack of knowledge about the design of slurry reactors and hence of the impossible task of the engineer who has to design a plant according to accepted rules. There have been no fundamental changes since then. This paper presents the problems facing the engineer in designing slurry reactors, and shows new development trends.

  8. Reactor safety

    International Nuclear Information System (INIS)

    Butz, H.P.; Heuser, F.W.; May, H.

    1985-01-01

    The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

  9. Nuclear reactor

    International Nuclear Information System (INIS)

    Mysels, K.J.; Shenoy, A.S.

    1976-01-01

    A nuclear reactor is described in which the core consists of a number of fuel regions through each of which regulated coolant flows. The coolant from neighbouring fuel regions is combined in a manner which results in an averaging of the coolant temperature at the outlet of the core. By this method the presence of hot streaks in the reactor is reduced. (UK)

  10. Reactor container

    International Nuclear Information System (INIS)

    Kato, Masami; Nishio, Masahide.

    1987-01-01

    Purpose: To prevent the rupture of the dry well even when the melted reactor core drops into a reactor pedestal cavity. Constitution: In a reactor container in which a dry well disposed above the reactor pedestal cavity for containing the reactor pressure vessel and a torus type suppression chamber for containing pressure suppression water are connected with each other, the pedestal cavity and the suppression chamber are disposed such that the flow level of the pedestal cavity is lower than the level of the pressure suppression water. Further, a pressure suppression water introduction pipeway for introducing the pressure suppression water into the reactor pedestal cavity is disposed by way of an ON-OFF valve. In case if the melted reactor core should fall into the pedestal cavity, the ON-OFF valve for the pressure suppression water introduction pipeway is opened to introduce the pressure suppression water in the suppression chamber into the pedestal cavity to cool the melted reactor core. (Ikeda, J.)

  11. RA Reactor

    International Nuclear Information System (INIS)

    1989-01-01

    This chapter includes the following: General description of the RA reactor, organization of work, responsibilities of leadership and operators team, regulations concerning operation and behaviour in the reactor building, regulations for performing experiments, regulations and instructions for inserting samples into experimental channels [sr

  12. Reactor physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, H.

    1998-01-01

    Progress in research on reactor physics in 1997 at the Belgian Nuclear Research Centre SCK/CEN is described. Activities in the following four domains are discussed: core physics, ex-core neutron transport, experiments in Materials Testing Reactors, international benchmarks

  13. Radionuclide migration at sites of temporary storage of SNF and RW in North-West Russia - Contribution to regulatory development

    International Nuclear Information System (INIS)

    Sneve, M.K.; Shandala, N.K.; Orlova, E.I.; Titov, A.V.; Kochetkov, O.A.; Smith, G.M.; Barraclough, I.M.

    2007-01-01

    Two technical bases of the Northern Fleet were created in the Russian northwest in the 1960s at Andreeva in the Kola Bay and Gremikha village on the coast of the Barents Sea. They maintained nuclear submarines, performing receipt and storage of radioactive waste and spent nuclear fuel. No further stored material was received after 1985. These technical bases have since been re-categorised as sites of temporary storage. It is necessary to note that, during the storage of RW and SNF, certain conditions arose which resulted in failure of the storage barrier system, resulting in release of radionuclides. Remediation activities at the site focus on reduction of major risks associated with most hazardous radioactive source terms. In addition, the long term management of the sites includes consideration of how to remediate contaminated areas, not only because they affect continuing work at the site, but also because this work will influence final radiological status of the sites. The optimum approach to remediation will be affected by how quickly radionuclides could move, both during the remediation works and, so far as any residual activity is concerned, after the works are completed. Present investigations reported here are directed to determination of sorption-desorption parameters of radionuclides in the studied areas, which will affect their underground migration, with the purpose of accounting for regional peculiarities in optimization process of the STSs remediation. The work is being carried out by the TSO State Research Centre - Institute of Biophysics, of Russian Federation, with assistance from western experts. The work forms part of a regulatory collaboration programme on-going between the Norwegian Radiation Protection Authority and the Federal Medical-Biological Agency which is designed to support the development of norms and standards to be applied in the remediation of these sites of temporary storage. (author)

  14. Reactor core

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

    In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

  15. Reactor container

    International Nuclear Information System (INIS)

    Fukazawa, Masanori.

    1991-01-01

    A system for controlling combustible gases, it has been constituted at present such that the combustible gases are controlled by exhausting them to the wet well of a reactor container. In this system, however, there has been a problem, in a reactor container having plenums in addition to the wet well and the dry well, that the combustible gases in such plenums can not be controlled. In view of the above, in the present invention, suction ports or exhaust ports of the combustible gas control system are disposed to the wet well, the dry well and the plenums to control the combustible gases in the reactor container. Since this can control the combustible gases in the entire reactor container, the integrity of the reactor container can be ensured. (T.M.)

  16. Reactor container

    International Nuclear Information System (INIS)

    Kojima, Yoshihiro; Hosomi, Kenji; Otonari, Jun-ichiro.

    1997-01-01

    In the present invention, a catalyst for oxidizing hydrogen to be disposed in a reactor container upon rupture of pipelines of a reactor primary coolant system is prevented from deposition of water droplets formed from a reactor container spray to suppress elevation of hydrogen concentration in the reactor container. Namely, a catalytic combustion gas concentration control system comprises a catalyst for oxidizing hydrogen and a support thereof. In addition, there is also disposed a water droplet deposition-preventing means for preventing deposition of water droplets in a reactor pressure vessel on the catalyst. Then, the effect of the catalyst upon catalytic oxidation reaction of hydrogen can be kept high. The local elevation of hydrogen concentration can be prevented even upon occurrence of such a phenomenon that various kinds of mobile forces in the container such as dry well cooling system are lost. (I.S.)

  17. Nuclear reactor

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1975-01-01

    A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

  18. Spent Nuclear Fuel Option Study on Hybrid Reactor for Waste Transmutation

    International Nuclear Information System (INIS)

    Hong, Seong Hee; Kim, Myung Hyun

    2016-01-01

    DUPIC nuclear fuel can be used in hybrid reactor by compensation of subcritical level through (U-10Zr) fuel. Energy production performance of Hyb-WT with DUPIC is grateful because it has high EM factor and performs waste transmutation at the same time. However, waste transmutation performance should be improved by different fissile fuel instead of (U-10Zr) fuel. SNF (Spent Nuclear Fuel) disposal is one of the problems in the nuclear industry. FFHR (Fusion-Fission Hybrid Reactor) is one of the most attractive option on reuse of SNF as a waste transmutation system. Because subcritical system like FFHR has some advantages compared to critical system. Subcritical systems have higher safety potential than critical system. Also, there is suppressed excess reactivity at BOC (Beginning of Cycle) in critical system, on the other hand there is no suppressed reactivity in subcritical system. Our research team could have designed FFHR for waste transmutation; Hyb-WT. Various researches have been conducted on fuel and coolant option for optimization of transmutation performance. However, Hyb-WT has technical disadvantage. It is required fusion power (Pfus) which is the key design parameter in FFHR is increased for compensation of decreasing subcritical level. As a result, structure material integrity is damaged under high irradiation condition by increasing Pfus. Also, deep burn of reprocessed SNF is limited by weakened integrity of structure material. Therefore, in this research, SNF option study will be conducted on DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactor) fuel, TRU fuel and DUPIC + TRU mixed fuel for optimization of Hyb-WT performance. Goal of this research is design check for low required fusion power and high waste transmutation. In this paper, neutronic analysis is conducted on Hyb-WT with DUPIC nuclear fuel. When DUPIC nuclear fuel is loaded in fast neutron system, supplement fissile materials need to be loaded together for compensation of low criticality

  19. Spent Nuclear Fuel Option Study on Hybrid Reactor for Waste Transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seong Hee; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

    2016-05-15

    DUPIC nuclear fuel can be used in hybrid reactor by compensation of subcritical level through (U-10Zr) fuel. Energy production performance of Hyb-WT with DUPIC is grateful because it has high EM factor and performs waste transmutation at the same time. However, waste transmutation performance should be improved by different fissile fuel instead of (U-10Zr) fuel. SNF (Spent Nuclear Fuel) disposal is one of the problems in the nuclear industry. FFHR (Fusion-Fission Hybrid Reactor) is one of the most attractive option on reuse of SNF as a waste transmutation system. Because subcritical system like FFHR has some advantages compared to critical system. Subcritical systems have higher safety potential than critical system. Also, there is suppressed excess reactivity at BOC (Beginning of Cycle) in critical system, on the other hand there is no suppressed reactivity in subcritical system. Our research team could have designed FFHR for waste transmutation; Hyb-WT. Various researches have been conducted on fuel and coolant option for optimization of transmutation performance. However, Hyb-WT has technical disadvantage. It is required fusion power (Pfus) which is the key design parameter in FFHR is increased for compensation of decreasing subcritical level. As a result, structure material integrity is damaged under high irradiation condition by increasing Pfus. Also, deep burn of reprocessed SNF is limited by weakened integrity of structure material. Therefore, in this research, SNF option study will be conducted on DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactor) fuel, TRU fuel and DUPIC + TRU mixed fuel for optimization of Hyb-WT performance. Goal of this research is design check for low required fusion power and high waste transmutation. In this paper, neutronic analysis is conducted on Hyb-WT with DUPIC nuclear fuel. When DUPIC nuclear fuel is loaded in fast neutron system, supplement fissile materials need to be loaded together for compensation of low criticality

  20. Metallographic examination of damaged N reactor spent nuclear fuel element SFEC5,4378

    Energy Technology Data Exchange (ETDEWEB)

    Marschman, S.C.; Pyecha, T.D.; Abrefah, J.

    1997-08-01

    N-Reactor spent nuclear fuel (SNF) is currently residing underwater in the K Basins at the Hanford site, in Richland, Washington. This report presents results of the metallographic examination of specimens cut from an SNF element (Mark IV-E) with breached cladding. The element had resided in the K-West (KW) Storage Basin for at least 10 years after it was discharged from the N-Reactor. The storage containers in the KW Basin were nominally closed, isolating the SNF elements from the open pool environment. Seven specimens from this Mark IV-E outer fuel element were examined using an optical metallograph. Included were two specimens that had been subjected to a conditioning process recommended by the Independent Technical Assessment Team, two specimens that had been subjected to a conditioning process recommended in the Integrated Process Strategy Report, and three that were in the as-received, as-cut condition. One of the as-received specimens had been cut from the damaged (or breached) end of the element. All other specimens were cut from the undamaged mid-region of the fuel element. The specimens were visually examined to (1) identify uranium hydride inclusions present in the uranium metal fuel, (2) measure the thickness of the oxide layer formed on the uranium edges and assess the apparent integrity and adhesion of the oxide layer, and (3) look for features in the microstructure that might provide an insight into the various corrosion processes that occurred during underwater storage in the KW Basin. These features included, but were not limited to, the integrity of the cladding and the fuel-to-cladding bond, obvious anomalies in the microstructure, excessive pitting or friability of the fuel matrix, and obvious anomalies in the distribution of uranium hydride or uranium carbide inclusions. Also, the observed metallographic features of the conditioned specimens were compared with those of the as-received (unconditioned) specimens. 11 refs., 93 figs., 2 tabs.

  1. Reactor building

    International Nuclear Information System (INIS)

    Maruyama, Toru; Murata, Ritsuko.

    1996-01-01

    In the present invention, a spent fuel storage pool of a BWR type reactor is formed at an upper portion and enlarged in the size to effectively utilize the space of the building. Namely, a reactor chamber enhouses reactor facilities including a reactor pressure vessel and a reactor container, and further, a spent fuel storage pool is formed thereabove. A second spent fuel storage pool is formed above the auxiliary reactor chamber at the periphery of the reactor chamber. The spent fuel storage pool and the second spent fuel storage pool are disposed in adjacent with each other. A wall between both of them is formed vertically movable. With such a constitution, the storage amount for spent fuels is increased thereby enabling to store the entire spent fuels generated during operation period of the plant. Further, since requirement of the storage for the spent fuels is increased stepwisely during periodical exchange operation, it can be used for other usage during the period when the enlarged portion is not used. (I.S.)

  2. Reactor container

    International Nuclear Information System (INIS)

    Shibata, Satoru; Kawashima, Hiroaki

    1984-01-01

    Purpose: To optimize the temperature distribution of the reactor container so as to moderate the thermal stress distribution on the reactor wall of LMFBR type reactor. Constitution: A good heat conductor (made of Al or Cu) is appended on the outer side of the reactor container wall from below the liquid level to the lower face of a deck plate. Further, heat insulators are disposed to the outside of the good heat conductor. Furthermore, a gas-cooling duct is circumferentially disposed at the contact portion between the good heat conductor and the deck plate around the reactor container. This enables to flow the cold heat from the liquid metal rapidly through the good heat conductor to the cooling duct and allows to maintain the temperature distribution on the reactor wall substantially linear even with the abrupt temperature change in the liquid metal. Further, by appending the good heat conductor covered with inactive metals not only on the outer side but also on the inside of the reactor wall to introduce the heat near the liquid level to the upper portion and escape the same to the cooling layer below the roof slab, the effect can be improved further. (Ikeda, J.)

  3. Nuclear reactor

    International Nuclear Information System (INIS)

    Rau, P.

    1980-01-01

    The reactor core of nuclear reactors usually is composed of individual elongated fuel elements that may be vertically arranged and through which coolant flows in axial direction, preferably from bottom to top. With their lower end the fuel elements gear in an opening of a lower support grid forming part of the core structure. According to the invention a locking is provided there, part of which is a control element that is movable along the fuel element axis. The corresponding locking element is engaged behind a lateral projection in the opening of the support grid. The invention is particularly suitable for breeder or converter reactors. (orig.) [de

  4. The FOX transcription factor Hcm1 regulates oxidative metabolism in response to early nutrient limitation in yeast. Role of Snf1 and Tor1/Sch9 kinases.

    Science.gov (United States)

    Rodríguez-Colman, María José; Sorolla, M Alba; Vall-Llaura, Núria; Tamarit, Jordi; Ros, Joaquim; Cabiscol, Elisa

    2013-08-01

    Within Saccharomyces cerevisiae, Hcm1is a member of the forkhead transcription factor family with a role in chromosome organization. Our group recently described its involvement in mitochondrial biogenesis and stress resistance, and reports here that Hcm1 played a role in adaptation to respiratory metabolism when glucose or nitrogen was decreased. Regulation of Hcm1 activity occurs in at least three ways: i) protein quantity, ii) subcellular localization, and iii) transcriptional activity. Transcriptional activity was measured using a reporter gene fused to a promoter that contains a binding site for Hcm1. We also analyzed the levels of several genes whose expression is known to be regulated by Hcm1 levels and the role of the main kinases known to respond to nutrients. Lack of sucrose-nonfermenting (Snf1) kinase increases cytoplasmic localization of Hcm1, whereas Δtor1 cells showed a mild increase in nuclear Hcm1. In vitro experiments showed that Snf1 clearly phosphorylates Hcm1 while Sch9 exerts a milder phosphorylation. Although in vitroTor1 does not directly phosphorylate Hcm1, in vivo rapamycin treatment increases nuclear Hcm1. We conclude that Hcm1 participates in the adaptation of cells from fermentation to respiratory metabolism during nutrient scarcity. According to our hypothesis, when nutrient levels decrease, Snf1 phosphorylates Hcm1. This results in a shift from the cytoplasm to the nucleus and increased transcriptional activity of genes involved in respiration, use of alternative energy sources, NAD synthesis and oxidative stress resistance. Copyright © 2013 Elsevier B.V. All rights reserved.

  5. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.

    1976-01-01

    A nuclear reactor containment vessel faced internally with a metal liner is provided with thermal insulation for the liner, comprising one or more layers of compressible material such as ceramic fiber, such as would be conventional in an advanced gas-cooled reactor and also a superposed layer of ceramic bricks or tiles in combination with retention means therefor, the retention means (comprising studs projecting from the liner, and bolts or nuts in threaded engagement with the studs) being themselves insulated from the vessel interior so that the coolant temperatures achieved in a High-Temperature Reactor or a Fast Reactor can be tolerated with the vessel. The layer(s) of compressible material is held under a degree of compression either by the ceramic bricks or tiles themselves or by cover plates held on the studs, in which case the bricks or tiles are preferably bedded on a yielding layer (for example of carbon fibers) rather than directly on the cover plates

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Miyashita, Akio.

    1981-01-01

    Purpose: To facilitate and accelerate a leakage test of valves of a main steam pipe by adding a leakage test partition valve thereto. Constitution: A leakage testing partition valve is provided between a pressure vessel for a nuclear reactor and the most upstream side valve of a plurality of valves to be tested for leakage, a testing branch pipe is communicated with the downstream side of the partition valve, and the testing water for preventing leakage is introduced thereto through the branch pipe. Since main steam pipe can be simply isolated by closing the partition valve in the leakage test, the leakage test can be conducted without raising or lowering the water level in the pressure vessel, and since interference with other work in the reactor can be eliminated, the leakage test can be readily conducted parallel with other work in the reactor in a short time. Clean water can be used without using reactor water as the test water. (Yoshihara, H.)

  7. Reactor container

    International Nuclear Information System (INIS)

    Abe, Yoshihito; Sano, Tamotsu; Ueda, Sabuo; Tanaka, Kazuhisa.

    1987-01-01

    Purpose: To improve the liquid surface disturbance in LMFBR type reactors. Constitution: A horizontal flow suppressing mechanism mainly comprising vertical members is suspended near the free liquid surface of coolants in the upper plenum. The horizontal flow of coolants near the free liquid surface is reduced by the suppressing mechanism to effectively reduce the surface disturbance. The reduction in the liquid surface disturbance further prevails to the entire surface region with no particular vertical variations to the free liquid surface to remarkably improve the preventive performance for the liquid surface disturbance. Accordingly, it is also possible to attain the advantageous effects such as prevention for the thermal fatigue in reactor vessel walls, reactor upper mechanisms, etc. and prevention of burning damage to the reactor core due to the reduction of envolved Ar gas. (Kamimura, M.)

  8. REACTOR SHIELD

    Science.gov (United States)

    Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  9. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  10. Foreign research reactor uranium supply program: The Y-12 national security complex process

    International Nuclear Information System (INIS)

    Nelson, T.; Eddy, B.G.

    2010-01-01

    The Foreign Research Reactor (FRR) Uranium Supply Program at the Y-12 National Security Complex supports the nonproliferation objectives of the HEU Disposition Program, the Reduced Enrichment Research and Test Reactors (RERTR) Program, and the United States FRR Spent Nuclear Fuel (SNF) Acceptance Program. The Y-12 National Nuclear Security Administration (NNSA) Y-12 Site Office maintains the prime contracts with foreign governments for the supply of Low-Enriched Uranium (LEU) for their research reactors. The LEU is produced by down blending Highly Enriched Uranium (HEU) that has been declared surplus to the U.S. national defense needs. The down blending and sale of the LEU supports the Surplus HEU Disposition Program Record of Decision to make the HEU non-weapons usable and to recover the economic value of the uranium to the extent feasible. This program supports the important U.S. government and nuclear nonproliferation commitment to serve as a reliable and cost-effective uranium supplier for those foreign research reactors that are converting or have converted to LEU fuel under the guidance of the NNSA RERTR Program. In conjunction with the FRR SNF Acceptance Program which supports the global nonproliferation efforts to disposition U.S.-origin HEU, the Y-12 FRR Uranium Supply Program can provide the LEU for the replacement fuel fabrication. In addition to feedstock for fuel fabrication, Y-12 supplies LEU for target fabrication for medical isotope production. The Y-12 process uses supply forecasting tools, production improvements and efficient delivery preparations to successfully support the global research reactor community

  11. Breeder reactors

    International Nuclear Information System (INIS)

    Gollion, H.

    1977-01-01

    The reasons for the development of fast reactors are briefly reviewed (a propitious neutron balance oriented towards a maximum uranium burnup) and its special requirements (cooling, fissile material density and reprocessing) discussed. The three stages in the French program of fast reactor development are outlined with Rapsodie at Cadarache, Phenix at Marcoule, and Super Phenix at Creys-Malville. The more specific features of the program of research and development are emphasized: kinetics and the core, the fuel and the components [fr

  12. Nuclear reactor

    International Nuclear Information System (INIS)

    Schulze, I.; Gutscher, E.

    1980-01-01

    The core contains a critical mass of UN or U 2 N 3 in the form of a noncritical solution with melted Sn being kept below a N atmosphere. The lining of the reactor core consists of graphite. If fission progresses part of the melted metal solution is removed and cleaned from fission products. The reactor temperatures lie in the range of 300 to 2000 0 C. (Examples and tables). (RW) [de

  13. Reactor technology

    International Nuclear Information System (INIS)

    Erdoes, P.

    1977-01-01

    This is one of a series of articles discussing aspects of nuclear engineering ranging from a survey of various reactor types for static and mobile use to mention of atomic thermo-electric batteries of atomic thermo-electric batteries for cardiac pacemakers. Various statistics are presented on power generation in Europe and U.S.A. and economics are discussed in some detail. Molten salt reactors and research machines are also described. (G.M.E.)

  14. Scenarios for potential radionuclide release from marine reactors dumped in the Kara Sea

    International Nuclear Information System (INIS)

    Lynn, N.; Mount, M.; Gussgard, K.

    1995-01-01

    The largest inventory of radioactive materials dumped in the Kara Sea by the former Soviet Union comes from the spent nuclear fuel (SNF) of seven marine reactors, the current (1994) inventory of which makes a total of approximately 4.7x10 15 Bq. In progressing its work for the International Arctic Seas Assessment Project, under the auspices of the International Atomic Energy Agency, the Source Term Working Group has analysed the Source Term and subsequently developed a number of model scenarios for the potential release patterns of radionuclides into the Kara Sea from the SNF and activated components dumped within the marine reactors.These models are based on the present and future conditions of the barrier materials and their configuration within the dumped objects. They account for progressive corrosion of the outer and inner steel barriers, breakdown of the organic fillers, and degradation and leaching from the SNFs. Annual release rates are predicted to four thousand years into the future. 5 refs., 1 fig., 1 tab

  15. Reactor containment

    International Nuclear Information System (INIS)

    Kawabe, Ryuhei; Yamaki, Rika.

    1990-01-01

    A water vessel is disposed and the gas phase portion of the water vessel is connected to a reactor container by a pipeline having a valve disposed at the midway thereof. A pipe in communication with external air is extended upwardly from the liquid phase portion to a considerable height so as to resist against the back pressure by a waterhead in the pipeline. Accordingly, when the pressure in the container is reduced to a negative level, air passes through the pipeline and uprises through the liquid phase portion in the water vessel in the form of bubbles and then flows into the reactor container. When the pressure inside of the reactor goes higher, since the liquid surface in the water vessel is forced down, water is pushed up into the pipeline. Since the waterhead pressure of a column of water in the pipeline and the pressure of the reactor container are well-balanced, gases in the reactor container are not leaked to the outside. Further, in a case if a great positive pressure is formed in the reactor container, the inner pressure overcomes the waterhead of the column of water, so that the gases containing radioactive aerosol uprise in the pipeline. Since water and the gases flow being in contact with each other, this can provide the effect of removing aerosol. (T.M.)

  16. Fast reactors

    International Nuclear Information System (INIS)

    Vasile, A.

    2001-01-01

    Fast reactors have capacities to spare uranium natural resources by their breeding property and to propose solutions to the management of radioactive wastes by limiting the inventory of heavy nuclei. This article highlights the role that fast reactors could play for reducing the radiotoxicity of wastes. The conversion of 238 U into 239 Pu by neutron capture is more efficient in fast reactors than in light water reactors. In fast reactors multi-recycling of U + Pu leads to fissioning up to 95% of the initial fuel ( 238 U + 235 U). 2 strategies have been studied to burn actinides: - the multi-recycling of heavy nuclei is made inside the fuel element (homogeneous option); - the unique recycling is made in special irradiation targets placed inside the core or at its surroundings (heterogeneous option). Simulations have shown that, for the same amount of energy produced (400 TWhe), the mass of transuranium elements (Pu + Np + Am + Cm) sent to waste disposal is 60,9 Kg in the homogeneous option and 204.4 Kg in the heterogeneous option. Experimental programs are carried out in Phenix and BOR60 reactors in order to study the feasibility of such strategies. (A.C.)

  17. SFCOMPO 2.0 - A relational database of spent fuel isotopic measurements, reactor operational histories, and design data

    Science.gov (United States)

    Michel-Sendis, Franco; Martinez-González, Jesus; Gauld, Ian

    2017-09-01

    SFCOMPO-2.0 is a database of experimental isotopic concentrations measured in destructive radiochemical analysis of spent nuclear fuel (SNF) samples. The database includes corresponding design description of the fuel rods and assemblies, relevant operating conditions and characteristics of the host reactors necessary for modelling and simulation. Aimed at establishing a thorough, reliable, and publicly available resource for code and data validation of safety-related applications, SFCOMPO-2.0 is developed and maintained by the OECD Nuclear Energy Agency (NEA). The SFCOMPO-2.0 database is a Java application which is downloadable from the NEA website.

  18. SFCOMPO 2.0 – A relational database of spent fuel isotopic measurements, reactor operational histories, and design data

    Directory of Open Access Journals (Sweden)

    Michel-Sendis Franco

    2017-01-01

    Full Text Available SFCOMPO-2.0 is a database of experimental isotopic concentrations measured in destructive radiochemical analysis of spent nuclear fuel (SNF samples. The database includes corresponding design description of the fuel rods and assemblies, relevant operating conditions and characteristics of the host reactors necessary for modelling and simulation. Aimed at establishing a thorough, reliable, and publicly available resource for code and data validation of safety-related applications, SFCOMPO-2.0 is developed and maintained by the OECD Nuclear Energy Agency (NEA. The SFCOMPO-2.0 database is a Java application which is downloadable from the NEA website.

  19. Generation IV reactors: reactor concepts

    International Nuclear Information System (INIS)

    Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

    2003-01-01

    Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

  20. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs

  1. Study of the feasibility of distributed cathodic arc as a plasma source for development of the technology for plasma separation of SNF and radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    Amirov, R. Kh.; Vorona, N. A.; Gavrikov, A. V.; Liziakin, G. D.; Polistchook, V. P.; Samoylov, I. S.; Smirnov, V. P.; Usmanov, R. A., E-mail: ravus46@yandex.ru; Yartsev, I. M. [Russian Academy of Sciences, Joint Institute for High Temperatures (Russian Federation)

    2015-12-15

    One of the key problems in the development of plasma separation technology is designing a plasma source which uses condensed spent nuclear fuel (SNF) or nuclear wastes as a raw material. This paper covers the experimental study of the evaporation and ionization of model materials (gadolinium, niobium oxide, and titanium oxide). For these purposes, a vacuum arc with a heated cathode on the studied material was initiated and its parameters in different regimes were studied. During the experiment, the cathode temperature, arc current, arc voltage, and plasma radiation spectra were measured, and also probe measurements were carried out. It was found that the increase in the cathode heating power leads to the decrease in the arc voltage (to 3 V). This fact makes it possible to reduce the electron energy and achieve singly ionized plasma with a high degree of ionization to fulfill one of the requirements for plasma separation of SNF. This finding is supported by the analysis of the plasma radiation spectrum and the results of the probe diagnostics.

  2. DOC1-Dependent Recruitment of NURD Reveals Antagonism with SWI/SNF during Epithelial-Mesenchymal Transition in Oral Cancer Cells

    Directory of Open Access Journals (Sweden)

    Adone Mohd-Sarip

    2017-07-01

    Full Text Available The Nucleosome Remodeling and Deacetylase (NURD complex is a key regulator of cell differentiation that has also been implicated in tumorigenesis. Loss of the NURD subunit Deleted in Oral Cancer 1 (DOC1 is associated with human oral squamous cell carcinomas (OSCCs. Here, we show that restoration of DOC1 expression in OSCC cells leads to a reversal of epithelial-mesenchymal transition (EMT. This is caused by the DOC1-dependent targeting of NURD to repress key transcriptional regulators of EMT. NURD recruitment drives extensive epigenetic reprogramming, including eviction of the SWI/SNF remodeler, formation of inaccessible chromatin, H3K27 deacetylation, and binding of PRC2 and KDM1A, followed by H3K27 methylation and H3K4 demethylation. Strikingly, depletion of SWI/SNF mimics the effects of DOC1 re-expression. Our results suggest that SWI/SNF and NURD function antagonistically to control chromatin state and transcription. We propose that disturbance of this dynamic equilibrium may lead to defects in gene expression that promote oncogenesis.

  3. Nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Prescott, R F; George, B V; Baglin, C J

    1978-05-10

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given.

  4. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1978-01-01

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given. (UK)

  5. 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.

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Scholz, M.

    1976-01-01

    An improvement of the accessibility of that part of a nuclear reactor serving for biological shield is proposed. It is intended to provide within the biological shield, distributed around the circumference of the reactor pressure vessel, several shielding chambers filled with shielding material, which are isolated gastight from the outside by means of glass panes with a given bursting strength. It is advantageous that, on the one hand, inspection and maintenance will be possible without great effort and, on the other, a large relief cross section will be at desposal if required. (UWI) [de

  7. NEUTRONIC REACTOR

    Science.gov (United States)

    Wigner, E.P.; Weinberg, A.W.; Young, G.J.

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

  8. 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

  9. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield

  10. Reactor facility

    International Nuclear Information System (INIS)

    Suzuki, Hiroaki; Murase, Michio; Yokomizo, Osamu.

    1997-01-01

    The present invention provides a BWR type reactor facility capable of suppressing the amount of steams generated by the mutual effect of a failed reactor core and coolants upon occurrence of an imaginal accident, and not requiring spacial countermeasures for enhancing the pressure resistance of the container vessel. Namely, a means for supplying cooling water at a temperature not lower by 30degC than the saturated temperature corresponding to the inner pressure of the containing vessel upon occurrence of an accident is disposed to a lower dry well below the pressure vessel. As a result, upon occurrence of such an accident that the reactor core should be melted and flown downward of the pressure vessel, when cooling water at a temperature not lower than the saturated temperature, for example, cooling water at 100degC or higher is supplied to the lower dry well, abrupt generation of steams by the mutual effect of the failed reactor core and cooling water is scarcely caused compared with a case of supplying cooling water at a temperature lower than the saturation temperature by 30degC or more. Accordingly, the amount of steams to be generated can be suppressed, and special countermeasure is no more necessary for enhancing the pressure resistance of the container vessel is no more necessary. (I.S.)

  11. Nuclear reactor

    International Nuclear Information System (INIS)

    Gilroy, J.E.

    1980-01-01

    An improved cover structure for liquid metal cooled fast breeder type reactors is described which it is claimed reduces the temperature differential across the intermediate grid plate of the core cover structure and thereby reduces its subjection to thermal stresses. (UK)

  12. Reactor licensing

    International Nuclear Information System (INIS)

    Harvie, J.D.

    2002-01-01

    This presentation discusses reactor licensing and includes the legislative basis for licensing, other relevant legislation , the purpose of the Nuclear Safety and Control Act, important regulations, regulatory document, policies, and standards. It also discusses the role of the CNSC, its mandate and safety philosophy

  13. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sekine, Katsuhisa.

    1987-01-01

    Purpose: To decrease the thickness of a reactor container and reduce the height and the height and plate thickness of a roof slab without using mechanical vibration stoppers. Constitution: Earthquake proofness is improved by filling fluids such as liquid metal between a reactor container and a secondary container and connecting the outer surface of the reactor container with the inner surface of the secondary container by means of bellows. That is, for the horizontal seismic vibrations, horizontal loads can be supported by the secondary container without providing mechanical vibration stoppers to the reactor container and the wall thickness can be reduced thereby enabling to simplify thermal insulation structure for the reduction of thermal stresses. Further, for the vertical seismic vibrations, verical loads can be transmitted to the secondary container thereby enabling to reduce the wall thickness in the same manner as for the horizontal load. By the effect of transferring the point of action of the container load applied to the roof slab to the outer circumferential portion, the intended purpose can be attained and, in addition, the radiation dose rate at the upper surface of the roof slab can be decreased. (Kamimura, M.)

  14. Reactor system

    International Nuclear Information System (INIS)

    Miyano, Hiroshi; Narabayashi, Naoshi.

    1990-01-01

    The represent invention concerns a reactor system with improved water injection means to a pressure vessel of a BWR type reactor. A steam pump is connected to a heat removing system pipeline, a high pressure water injection system pipeline and a low pressure water injection system pipeline for injecting water into the pressure vessel. A pump actuation pipeline is disposed being branched from a main steam pump or a steam relieaf pipeline system, through which steams are supplied to actuate the steam pump and supply cooling water into the pressure vessel thereby cooling the reactor core. The steam pump converts the heat energy into the kinetic energy and elevates the pressure of water to a level higher than the pressure of the steams supplied by way of a pressure-elevating diffuser. Cooling water can be supplied to the pressure vessel by the pressure elevation. This can surely inject cooling water into the pressure vessel upon loss of coolant accident or in a case if reactor scram is necessary, without using an additional power source. (I.N.)

  15. Reactor core

    International Nuclear Information System (INIS)

    Matsuura, Tetsuaki; Nomura, Teiji; Tokunaga, Kensuke; Okuda, Shin-ichi

    1990-01-01

    Fuel assemblies in the portions where the gradient of fast neutron fluxes between two opposing faces of a channel box is great are kept loaded at the outermost peripheral position of the reactor core also in the second operation cycle in the order to prevent interference between a control rod and the channel box due to bending deformation of the channel box. Further, the fuel assemblies in the second row from the outer most periphery in the first operation cycle are also kept loaded at the second row in the second operation cycle. Since the gradient of the fast neutrons in the reactor core is especially great at the outer circumference of the reactor core, the channel box at the outer circumference is bent such that the surface facing to the center of the reactor core is convexed and the channel box in the second row is also bent to the identical direction, the insertion of the control rod is not interfered. Further, if the positions for the fuels at the outermost periphery and the fuels in the second row are not altered in the second operation cycle, the gaps are not reduced to prevent the interference between the control rod and the channel box. (N.H.)

  16. New about research reactors

    International Nuclear Information System (INIS)

    Egorenkov, P.M.

    2001-01-01

    The multi-purpose research reactor MAPLE (Canada) and concept of new reactor MAPLE-CNF as will substitute the known Canadian research reactor NRU are described. New reactor will be used as contributor for investigations into materials, neutron beams and further developments for the CANDU type reactor. The Budapest research reactor (BRR) and its application after the last reconstruction are considered also [ru

  17. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

    1994-01-01

    In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

  18. Estimating the Permittivity of Rogers 4003C Substrate at Low Frequencies for Application in a Superdirective First-Order Probe for SNF Measurements

    DEFF Research Database (Denmark)

    Kim, Oleksiy S.

    2016-01-01

    The bulk permittivity of Rogers 4003C substrate is estimated in the lower UHF frequency band by comparing the simulated and measured return loss for a bandpass filter based on a coplanar waveguide and a capacitively loaded loop. The obtained value, which deviates from that specified by Rogers at ...... GHz, is subsequently utilized for accurate design of a new light-weight superdirective first-order probe for spherical near-field (SNF) antenna measurements at low frequencies.......The bulk permittivity of Rogers 4003C substrate is estimated in the lower UHF frequency band by comparing the simulated and measured return loss for a bandpass filter based on a coplanar waveguide and a capacitively loaded loop. The obtained value, which deviates from that specified by Rogers at 10...

  19. Poly(ADP-ribosyl)ation links the chromatin remodeler SMARCA5/SNF2H to RNF168-dependent DNA damage signaling

    DEFF Research Database (Denmark)

    Smeenk, G.; Wiegant, W.W.; Luijsterburg, M.S.

    2013-01-01

    Ionizing radiation (IR)-induced DNA double-strand breaks (DSBs) arising in native chromatin elicit an RNF8/RNF168-dependent ubiquitylation response, which triggers the recruitment of various repair factors. Precisely how this response is regulated in the context of chromatin remains largely...... unexplored. Here, we show that SMARCA5/SNF2H, the catalytic subunit of ISWI chromatin remodeling complexes, is recruited to DSBs in a poly(ADP-ribose) polymerase 1 (PARP1)-dependent manner. Remarkably, PARP activity, although dispensable for the efficient spreading of νH2AX into damaged chromatin......, selectively promotes spreading of SMARCA5, the E3 ubiquitin ligase RNF168, ubiquitin conjugates and the ubiquitin-binding factors RAD18 and the RAP80-BRCA1 complex throughout DSB-flanking chromatin. This suggests that PARP regulates the spatial organization of the RNF168-driven ubiquitin response to DNA...

  20. Nuclear reactor

    International Nuclear Information System (INIS)

    Gibbons, J.F.; McLaughlin, D.J.

    1978-01-01

    In the pressure vessel of the water-cooled nuclear reactor there is provided an internal flange on which the one- or two-part core barrel is hanging by means of an external flange. A cylinder is extending from the reactor vessel closure downwards to a seat on the core cupport structure and serves as compression element for the transmission of the clamping load from the closure head to the core barrel (upper guide structure). With the core barrel, subject to tensile stress, between the vessel internal flange and its seat on one hand and the compression of the cylinder resp. hold-down element between the closure head and the seat on the other a very strong, elastic sprung structure is obtained. (DG) [de

  1. Nuclear reactor

    International Nuclear Information System (INIS)

    Sasaki, Tomozo.

    1987-01-01

    Purpose: To improve the nuclear reactor availability by enabling to continuously exchange fuels in the natural-slightly enriched uranium region during operation. Constitution: A control rod is withdrawn to the midway of a highly enriched uranium region by means of control rod drives and the highly enriched uranium region is burnt to maintain the nuclear reactor always at a critical state. At the same time, fresh uranium-slightly enriched uranium is continuously supplied gravitationally from a fresh fuel reservoir through fuel reservoir to each of fuel pipes in the natural-slightly enriched uranium region. Then, spent fuels reduced with the reactivity by the burn up are successively taken out from the bottom of each of the fuel pipes through an exit duct and a solenoid valve to the inside of a spent fuel reservoir and the burn up in the natural-slightly enriched uranium region is conducted continuously. (Kawakami, Y.)

  2. Nuclear reactor

    International Nuclear Information System (INIS)

    Sakurai, Mikio; Yamauchi, Koki.

    1983-01-01

    Purpose: To improve the channel stability and the reactor core stability in a spontaneous circulation state of coolants. Constitution: A reactor core stabilizing device comprising a differential pressure automatic ON-OFF valve is disposed between each of a plurality of jet pumps arranged on a pump deck. The stabilizing device comprises a piston exerted with a pressure on the lower side of the pump deck by way of a pipeway and a valve for flowing coolants through the bypass opening disposed to the pump deck by the opening and closure of the valve ON-OFF. In a case where the jet pumps are stopped, since the differential pressure between the upper and the lower sides of the pump deck is removed, the valve lowers gravitationally into an opened state, whereby the coolants flow through the bypass opening to increase the spontaneous circulation amount thereby improve the stability. (Yoshino, Y.)

  3. Nuclear reactor

    International Nuclear Information System (INIS)

    Aleite, W.; Bock, H.W.; Struensee, S.

    1976-01-01

    The invention concerns the use of burnable poisons in a nuclear reactor, especially in PWRs, in order to improve the controllability of the reactor. An unsymmetrical arrangement in the lattice is provided, if necessary also by insertion of special rods for these additions. It is proposed to arrange the burnable poisons in fuel elements taken over from a previous burn-up cycle and to distribute them, going out from the side facing the control rods, over not more than 20% of the lenth of the fuel elements. It seems sufficient, for the burnable poisons to bind an initial reactivity of only 0.1% and to become ineffective after normal operation of 3 to 4 months. (ORU) [de

  4. Reactor container

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Saba, Kazuhisa.

    1979-01-01

    Purpose: To improve the earthquake resistance as well as reduce the size of a container for a nuclear reactor with no adverse effects on the decrease of impact shock to the container and shortening of construction step. Constitution: Reinforcing profile steel materials are welded longitudinally and transversely to the inner surface of a container, and inner steel plates are secured to the above profile steel materials while keeping a gap between the materials and the container. Reactor shielding wall planted to the base concrete of the container is mounted to the pressure vessel, and main steam pipeways secured by the transverse beams and led to the outside of container is connected. This can improve the rigidity earthquake strength and the safetiness against the increase in the inside pressure upon failures of the container. (Yoshino, Y.)

  5. Reactor container

    International Nuclear Information System (INIS)

    Oyamada, Osamu; Furukawa, Hideyasu; Uozumi, Hiroto.

    1979-01-01

    Purpose: To lower the position of an intermediate slab within a reactor container and fitting a heat insulating material to the inner wall of said intermediate slab, whereby a space for a control rod exchanging device and thermal stresses of the inner peripheral wall are lowered. Constitution: In the pedestal at the lower part of a reactor pressure vessel there is formed an intermediate slab at a position lower than diaphragm floor slab of the outer periphery of the pedestal thereby to secure a space for providing automatic exchanging device of a control rod driving device. Futhermore, a heat insulating material is fitted to the inner peripheral wall at the upper side of the intermediate slab part, and the temperature gradient in the wall thickness direction at the time of a piping rupture trouble is made gentle, and thermal stresses at the inner peripheral wall are lowered. (Sekiya, K.)

  6. Neutronic reactor

    International Nuclear Information System (INIS)

    Lewis, W.R.

    1978-01-01

    Disclosed is a graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels

  7. Nuclear reactors

    International Nuclear Information System (INIS)

    Humphreys, P.; Davidson, D.F.; Thatcher, G.

    1980-01-01

    The cooling system of a liquid metal cooled fast breeder nuclear reactor of the pool kind is described. It has an intermediate heat exchange module comprising a tube-in-shell heat exchanger and an electromagnetic flow coupler in the base region of the module. Primary coolant is flowed through the heat exchanger being driven by electromagnetic interaction with secondary liquid metal coolant flow effected by a mechanical pump. (author)

  8. Nuclear reactor

    International Nuclear Information System (INIS)

    Jungmann, A.

    1975-01-01

    Between a PWR's reactor pressure vessel made of steel and the biological shield made of concrete there is a gap. This gap is filled up with a heat insulation facting the reactor pressure vessel, for example with insulating concrete segments jacketed with sheet steel and with an additional layer. This layer serves for smooth absorption of compressive forces originating in radial direction from the reactor pressure vessel. It consists of cylinder-segment shaped bricks made of on situ concrete, for instance. The bricks have cooling agent ports in one or several rows which run parallel to the wall of the pressure vessel and in alignment with superposed bricks. Between the layer of bricks and the biological shield or rather the heat insulation, there are joints which are filled, however, with injected mortar. That guarantees a smooth series of connected components resistant tom compression. Besides, a slip foil can be set between the heat insulation and the joining joint filled with mortar for the reduction of the friction at thermal expansions. (TK) [de

  9. Reactor building

    International Nuclear Information System (INIS)

    Ebata, Sakae.

    1990-01-01

    At least one valve rack is disposed in a reactor building, on which pipeways to a main closure valve, valves and bypasses of turbines are placed and contained. The valve rack is fixed to the main body of the building or to a base mat. Since the reactor building is designed as class A earthquake-proofness and for maintaining the S 1 function, the valve rack can be fixed to the building main body or to the base mat. With such a constitution, the portions for maintaining the S 1 function are concentrated to the reactor building. As a result, the dispersion of structures of earthquake-proof portion corresponding to the reference earthquake vibration S 1 can be prevented. Accordingly, the conditions for the earthquake-proof design of the turbine building and the turbine/electric generator supporting rack are defined as only the class B earthquake-proof design conditions. In view of the above, the amount of building materials can be saved and the time for construction can be shortened. (I.S.)

  10. Nuclear reactors

    International Nuclear Information System (INIS)

    Yoshioka, Michiko.

    1985-01-01

    Purpose: To obtain an optimum structural arrangement of IRM having a satisfactory responsibility to the inoperable state of a nuclear reactor and capable of detecting the reactor power in an averaged manner. Constitution: As the structural arrangement of IRM, from 6 to 16 even number of IRM are bisected into equial number so as to belong two trip systems respectively, in which all of the detectors are arranged at an equal pitch along a circumference of a circle with a radius rl having the center at the position of the central control rod in one trip system, while one detector is disposed near the central control rod and other detectors are arranged substantially at an equal pitch along the circumference of a circle with a radius r2 having the center at the position for the central control rod in another trip system. Furthermore, the radius r1 and r2 are set such that r1 = 0.3 R, r2 = 0.5 R in the case where there are 6 IRM and r1 = 0.4 R and R2 = 0.8 R where there are eight IRM where R represents the radius of the reactor core. (Kawakami, Y.)

  11. MLR reactor

    International Nuclear Information System (INIS)

    Ryazantsev, E.P.; Egorenkov, P.M.; Nasonov, V.A.; Smimov, A.M.; Taliev, A.V.; Gromov, B.F.; Kousin, V.V.; Lantsov, M.N.; Radchenko, V.P.; Sharapov, V.N.

    1998-01-01

    The Material Testing Loop Reactor (MLR) development was commenced in 1991 with the aim of updating and widening Russia's experimental base to validate the selected directions of further progress of the nuclear power industry in Russia and to enhance its reliability and safety. The MLR reactor is the pool-type one. As coolant it applies light water and as side reflector beryllium. The direction of water circulation in the core is upward. The core comprises 30 FA arranged as hexagonal lattice with the 90-95 mm pitch. The central materials channel and six loop channels are sited in the core. The reflector includes up to 11 loop channels. The reactor power is 100 MW. The average power density of the core is 0.4 MW/I (maximal value 1.0 MW/l). The maximum neutron flux density is 7.10 14 n/cm 2 s in the core (E>0.1 MeV), and 5.10 14 n/cm 2 s in the reflector (E<0.625 eV). In 1995 due to the lack of funding the MLR designing was suspended. (author)

  12. Nuclear reactor

    International Nuclear Information System (INIS)

    Shirakawa, Toshihisa.

    1979-01-01

    Purpose: To prevent cladding tube injuries due to thermal expansion of each of the pellets by successively extracting each of the control rods loaded in the reactor core from those having less number of notches, as well as facilitate the handling work for the control rods. Constitution: A recycle flow control device is provided to a circulation pump for forcibly circulating coolants in the reactor container and an operational device is provided for receiving each of the signals concerning number of notches for each of the control rods and flow control depending on the xenon poisoning effect obtained from the signals derived from the in-core instrument system connected to the reactor core. The operational device is connected with a control rod drive for moving each of the control rods up and down and a recycle flow control device. The operational device is set with a pattern for the aimed control rod power and the sequence of extraction. Upon extraction of the control rods, they are extracted successively from those having less notch numbers. (Moriyama, K.)

  13. Reactor container

    International Nuclear Information System (INIS)

    Hidaka, Masataka; Hatamiya, Shigeo; Kawasaki, Terufumi; Fukui, Toru; Suzuki, Hiroaki; Kataoka, Yoshiyuki; Kawabe, Ryuhei; Murase, Michio; Naito, Masanori.

    1990-01-01

    In order to suppress the pressure elevation in a reactor container due to high temperature and high pressure steams jetted out upon pipeway rupture accidents in the reactor container, the steams are introduced to a pressure suppression chamber for condensating them in stored coolants. However, the ability for suppressing the pressure elevation and steam coagulation are deteriorated due to the presence of inactive incondensible gases. Then, there are disposed a vent channel for introducing the steams in a dry well to a pressure suppression chamber in the reactor pressure vessel, a closed space disposed at the position lower than a usual liquid level, a first channel having an inlet in the pressure suppression chamber and an exit in the closed space and a second means connected by way of a backflow checking means for preventing the flow directing to the closed space. The first paths are present by plurality, a portion of which constitutes a syphon. The incondensible gases and the steams are discharged to the dry well at high pressure by using the difference of the water head for a long cooling time after the pipeway rupture accident. Then, safety can be improved without using dynamic equipments as driving source. (N.H.)

  14. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1989-01-01

    In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

  15. The Y-12 National Security Complex Foreign Research Reactor Uranium Supply Production

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, T. [Nuclear Technology and Nonproliferation Programs, B and W Y-12, L.L.C., Y-12 National Security Complex, Oak Ridge, Tennessee (United States); Keller, A.P. [Disposition and Supply Programs, B and W Y-12, L.L.C., Y-12 National Security Complex, Oak Ridge, Tennessee (United States)

    2011-07-01

    The Foreign Research Reactor (FRR) Uranium Supply Program at the Y-12 National Security Complex supports the nonproliferation objectives of the National Nuclear Security Administration (NNSA) HEU Disposition, the Reduced Enrichment Research and Test Reactors (RERTR), and the United States (U.S.) FRR Spent Nuclear Fuel (SNF) Acceptance Programs. The FRR Supply Program supports the important U.S. government nuclear nonproliferation commitment to serve as a reliable and cost-effective uranium supplier for those foreign research reactors that are converting or have converted to Low-Enriched Uranium (LEU) fuel under the RERTR Program. The NNSA Y-12 Site Office maintains the prime contracts with foreign government agencies for the supply of LEU for their research reactors. The LEU is produced by down blending Highly Enriched Uranium (HEU) that has been declared surplus to the U.S. national defense needs. The down blending and sale of the LEU supports the Surplus HEU Disposition Program Record of Decision to make the HEU non-weapons usable and to recover the economic value of the uranium to the extent feasible. In addition to uranium metal feedstock for fuel fabrication, Y-12 can produce LEU in different forms to support new fuel development or target fabrication for medical isotope production. With production improvements and efficient delivery preparations, Y-12 continues to successfully support the global research reactor community. (author)

  16. Molten salt reactors: reactor cores

    International Nuclear Information System (INIS)

    1983-01-01

    In this critical analysis of the MSBR I project are examined the problems concerning the reactor core. Advantages of breeding depend essentially upon solutions to technological problems like continuous reprocessing or graphite behavior under neutron irradiation. Graphite deformation, moderator unloading, control rods and core instrumentation require more studies. Neutronics of the core, influence of core geometry and salt composition, fuel evolution, and thermohydraulics are reviewed [fr

  17. 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%

  18. Nuclear research reactors

    International Nuclear Information System (INIS)

    1985-01-01

    It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

  19. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    Baeten, P.

    2006-01-01

    The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

  20. 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

  1. Nuclear reactor

    International Nuclear Information System (INIS)

    Anthony, A.J.; Gruber, E.A.

    1979-01-01

    A nuclear reactor with control rods in channels between fuel assemblies wherein the fuel assemblies incorporate guide rods which protrude outwardly into the control rod channels to prevent the control rods from engaging the fuel elements. The guide rods also extend back into the fuel assembly such that they are relatively rigid members. The guide rods are tied to the fuel assembly end or support plates and serve as structural members which are supported independently of the fuel element. Fuel element spacing and support means may be attached to the guide rods. 9 claims

  2. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1979-01-01

    In a nuclear reactor (e.g. one having coolant down-flow through a core to a hearth below) thermal insulation (e.g. of a floor of the hearth) comprises a layer of bricks and a layer of tiles thereon, with smaller clearances between the tiles than between the bricks but with the bricks being of reduced cross-section immediately adjacent the tiles so as to be surrounded by interconnected passages, of relatively large dimensions, constituting a continuous chamber extending behind the layer of tiles. By this arrangement, lateral coolant flow in the inter-brick clearances is much reduced. The reactor core is preferably formed of hexagonal columns, supported on diamond-shaped plates each supported on a pillar resting on one of the hearth-floor tiles. Each plate has an internal duct, four upper channels connecting the duct with coolant ducts in four core columns supported by the plate, and lower channels connecting the duct to a downwardly-open recess common to three plates, grouped to form a hexagon, at their mutually-adjacent corners. This provides mixing, and temperature-averaging, of coolant from twelve columns

  3. Reactor container

    International Nuclear Information System (INIS)

    Oikawa, Hirohide; Otonari, Jun-ichiro; Tozaki, Yuka.

    1993-01-01

    Partition walls are disposed between a reactor pressure vessel and a suppression chamber to separate a dry well to an upper portion and a lower portion. A communication pipe is disposed to the partition walls. One end of the communication pipe is opened in an upper portion of the dry well at a position higher than a hole disposed to a bent tube of the suppression chamber. When coolants overflow from a depressurization valve by an erroneous operation of an emergency reactor core cooling device, the coolants accumulate in the upper portion of the dry well. When the pipeline is ruptured at the upper portion of the pressure vessel, only the inside of the pressure vessel and the upper portion of the dry well are submerged in water. In this case, the water level of the coolants does not elevate to the opening of the commuication pipe but they flow into the suppression chamber from the hole disposed to the bent tube. Since the coolants do not flow out to the lower portion of the dry well, important equipments such as control rod drives disposed at the lower portion of the dry wall can be prevented from submerging in water. (I.N.)

  4. Reactor monitor

    International Nuclear Information System (INIS)

    Takada, Tamotsu.

    1992-01-01

    The device of the present invention monitors a reactor so that each of the operations for the relocation of fuel assemblies and the withdrawal and the insertion of control rods upon exchange of fuel assemblies and control rods in the reactor. That is, when an operator conducts relocating operation by way of a fuel assembly operation section, the device of the present invention judges whether the operation indication is adequate or not, based on the information of control rod arrangement in a control rod memory section. When the operation indication is wrong, a stop signal is sent to a fuel assembly relocating device. Further, when the operator conducts control rod operation by way of a control rod operation section, the device of the present invention judges in the control rod withdrawal judging section, as to whether the operation indication given by the operator is adequate or not by comparing it with fuel assembly arrangement information. When the operation indication is wrong, a stop signal is sent to control rod drives. With such procedures, increase of nuclear heating upon occurrence of erroneous operation can be prevented. (I.S.)

  5. Nuclear reactors

    International Nuclear Information System (INIS)

    Matheson, J.E.

    1983-01-01

    A nuclear reactor has an upper and a lower grid plate. Protrusions project from the upper grid plate. Fuel assemblies having end fittings fit between the grid plates. An arrangement is provided for accepting axial forces generated during the operation of the nuclear reactor by the flow of the cooling medium and thermal expansion and irradiation-induced growth of the fuel assembly, which comprises rods. Each fuel assembly rests on the lower grid plate and its upper end is elastically supported against the upper grid plate by the above-mentioned arrangement. The arrangement comprises four (for example) torsion springs each having a torsion tube and a torsion bar nested within the torsion tube and connected at one end thereto. The other end of the torsion bar is connected to an associated one of four lever arms. The torsion tube is rigidly connected to the other end fitting and the springs are disposed such that the lever arms are biassed against the protrusions. (author)

  6. Reactor core fuel management

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1976-01-01

    The subject is covered in chapters, entitled: concepts of reactor physics; neutron diffusion; core heat transfer; reactivity; reactor operation; variables of core management; computer code modules; alternative reactor concepts; methods of optimization; general system aspects. (U.K.)

  7. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T [Huntington Beach, CA; Sahimi, Muhammad [Altadena, CA; Fayyaz-Najafi, Babak [Richmond, CA; Harale, Aadesh [Los Angeles, CA; Park, Byoung-Gi [Yeosu, KR; Liu, Paul K. T. [Lafayette Hill, PA

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  8. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-11-01

    This single page document is the November 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the production reactor.

  9. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-10-01

    This single page document is the October 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  10. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-10-15

    This single page document is the October 15, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  11. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-09-15

    This single page document is the September 15, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  12. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-12-15

    This single page document is the December 16, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production reactor.

  13. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-12-01

    This single page document is the December 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production reactor.

  14. Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

    International Nuclear Information System (INIS)

    Henry, A.F.

    1980-01-01

    Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

  15. The CEA research reactors

    International Nuclear Information System (INIS)

    Schwartz, J.P.

    1993-01-01

    Two main research reactors, specifically designed, PEGASE reactor and Laue-Langevin high flux reactor, are presented. The PEGASE reactor was designed at the end of the 50s for the study of the gas cooled reactor fuel element behaviour under irradiation; the HFR reactor, was designed in the late 60s to serve as a high yield and high level neutron source. Historical backgrounds, core and fuel characteristics and design, flux characteristics, etc., are presented. 5 figs

  16. Atomic reactor thermal engineering

    International Nuclear Information System (INIS)

    Kim, Gwang Ryong

    1983-02-01

    This book starts the introduction of atomic reactor thermal engineering including atomic reaction, chemical reaction, nuclear reaction neutron energy and soon. It explains heat transfer, heat production in the atomic reactor, heat transfer of fuel element in atomic reactor, heat transfer and flow of cooler, thermal design of atomic reactor, design of thermodynamics of atomic reactor and various. This deals with the basic knowledge of thermal engineering for atomic reactor.

  17. Nuclear reactor types

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

  18. Reactor safety

    International Nuclear Information System (INIS)

    Meneley, D.A.

    The people of Ontario have begun to receive the benefits of a low cost, assured supply of electrical energy from CANDU nuclear stations. This indigenous energy source also has excellent safety characteristics. Safety has been one of the central themes of the CANDU development program from its very beginning. A great deal of work has been done to establish that public risks are small. However, safety design criteria are now undergoing extensive review, with a real prospect of more stringent requirements being applied in the future. Considering the newness of the technology it is not surprising that a consensus does not yet exist; this makes it imperative to discuss the issues. It is time to examine the policies and practice of reactor safety management in Canada to decide whether or not further restrictions are justified in the light of current knowledge

  19. Nuclear reactor

    International Nuclear Information System (INIS)

    Schabert, H.P.; Weber, R.; Bauer, A.

    1975-01-01

    The refuelling of a PWR power reactor of about 1,200 MWe is performed by a transport pipe in the containment leading from an external to an internal fuel pit. A wagon to transport the fuel elements can go from a vertical loading position to an also vertical deloading position in the inner fuel pit via guide rollers. The necessary horizontal movement is effected by means of a cable line through the transport pipe which is inclined at least 10 0 . Gravity thus helps in the movement to the deloading position. The cable line with winch is fastened outside the containment. Swivelling devices tip the wagon from the horizontal to the vertical position or vice versa. Loading and deloading are done laterally. (TK/LH) [de

  20. Nuclear reactor

    International Nuclear Information System (INIS)

    Schweiger, F.; Glahe, E.

    1976-01-01

    In a nuclear reactor of the kind which is charged with spherical reaction elements and in which control rods are arranged to be thrust directly into the charge, each control rod has at least one screw thread on its external surface so that as the rod is thrust into the charge it is caused to rotate and thus make penetration easier. The length of each control rod may have two distinct portions, a latter portion which carries a screw thread and a lead-in portion which is shorter than the latter portion and which may carry a thread of greater pitch than that on the latter portion or may have a number of axially extending ribs instead of a thread

  1. Reactor container

    International Nuclear Information System (INIS)

    Furukawa, Hideyasu; Oyamada, Osamu; Uozumi, Hiroto.

    1976-01-01

    Purpose: To provide a container for a reactor provided with a pressure suppressing chamber pool which can prevent bubble vibrating load, particularly negative pressure generated at the time of starting to release exhaust from a main steam escape-safety valve from being transmitted to a lower liner plate of the container. Constitution: This arrangement is characterized in that a safety valve exhaust pool for main steam escape, in which a pressure suppressing chamber pool is separated and intercepted from pool water in the pressure suppressing chamber pool, a safety valve exhaust pipe is open into said safety valve exhaust pool, and an isolator member, which isolates the bottom liner plate in the pressure suppressing chamber pool from the pool water, is disposed on the bottom of the safety valve exhaust pool. (Nakamura, S.)

  2. Management of Spent Nuclear Fuel of Nuclear Research Reactor VVR-S at the National Institute of Physics and Nuclear Engineering, Bucharest, Romania

    Science.gov (United States)

    Biro, Lucian

    2009-05-01

    The Nuclear Research Reactor VVR-S (RR-VVR-S) located in Magurele-Bucharest, Romania, was designed for research and radioisotope production. It was commissioned in 1957 and operated without any event or accident for forty years until shut down in 1997. In 2002, by government decree, it was permanently shutdown for decommissioning. The National Institute of Physics and Nuclear Engineering (IFIN-HH) is responsible for decommissioning the RR-VVR-S, the first nuclear decommissioning project in Romania. In this context, IFIN-HH prepared and obtained approval from the Romanian Nuclear Regulatory Body for the Decommissioning Plan. One of the most important aspects for decommissioning the RR-VVR-S is solving the issue of the fresh and spent nuclear fuel (SNF) stored on site in wet storage pools. In the framework of the Russian Research Reactor Fuel Return Program (RRRFR), managed by the U.S. Department of Energy and in cooperation with the International Atomic Energy Agency and the Rosatom State Corporation, Romania repatriated all fresh HEU fuel to the Russian Federation in 2003 and the HEU SNF will be repatriated to Russia in 2009. With the experience and lessons learned from this action and with the financial support of the Romanian Government it will be possible for Romania to also repatriate the LEU SNF to the Russian Federation before starting the dismantling and decontamination of the nuclear facility. [4pt] In collaboration with K. Allen, Idaho National Laboratory, USA; L. Biro, National Commission for Nuclear Activities Control, Romania; and M. Dragusin, National Institute of Physics and Nuclear Engineering, Bucharest-Magurele, Romania.

  3. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.

    1976-01-01

    In the system described the fuel elements are arranged vertically in groups and are supported in such a manner as to tend to tilt them towards the center of the respective group, the fuel elements being urged laterally into abutment with one another. The elements have interlocking bearing pads, whereby lateral movement of adjacent elements is resisted; this improves the stability of the reactor core during refuelling operations. Fuel elements may comprise clusters of parallel fuel pins enclosed in a wrapper of hexagonal cross section, with bearing pads in the form of spline-like ribs located on each side of the wrapper and extending parallel to the longitudinal axis of the fuel element, being interlockable with ribs on pads of adjacent fuel elements. The arrangement is applicable to a reactor core in which fuel elements and control rod guide tubes are arranged in modules each of which comprises a cluster of at least three fuel elements, one of which is rigidly supported whilst the others are resiliently tilted towards the center of the cluster so as to lean on the rigidly supported element. It is also applicable to modules comprising a cluster of six fuel elements, each resiliently tilted towards a central void to form a circular arch. The modules may include additional fuel elements located outside the clusters and also resiliently tilted towards the central voids, the latter being used to accommodate control rod guide tubes. The need for separate structural members to act as leaning posts is thus avoided. Such structural members are liable to irradiation embrittlement, that could lead to core failure. (U.K.)

  4. Nuclear reactor neutron shielding

    Science.gov (United States)

    Speaker, Daniel P; Neeley, Gary W; Inman, James B

    2017-09-12

    A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

  5. Progress of the United States foreign research reactor spent nuclear fuel acceptance program

    International Nuclear Information System (INIS)

    Huizenga, D.G.; Clapper, M.; Thrower, A.W.

    2002-01-01

    The United States Department of Energy (DOE), in consultation with the Department of State (DOS), adopted the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel in May 1996. To date, the Foreign Research Reactor (FRR) Spent Nuclear Fuel (SNF) Acceptance Program has completed 23 shipments. Almost 5000 spent fuel assemblies from eligible research reactors throughout the world have been accepted into the United States under this program. Over the past year, another cross-country shipment of fuel was accomplished, as well as two additional shipments in the fourth quarter of calendar year 2001. These shipments attracted considerable safeguards oversight since they occurred post September 11. Recent guidance from the Nuclear Regulatory Commission (NRC) pertaining to security and safeguards issues deals directly with the transport of nuclear material. Since the Acceptance Program has consistently applied above regulatory safety enhancements in transport of spent nuclear fuel, this guidance did not adversely effect the Program. As the Program draws closer to its termination date, an increased number of requests for program extension are received. Currently, there are no plans to extend the policy beyond its current expiration date; therefore, eligible reactor operators interested in participating in this program are strongly encouraged to evaluate their inventory and plan for future shipments as soon as possible. (author)

  6. U.S. uranium supply to the research and test reactor community

    International Nuclear Information System (INIS)

    Parker, Elaine M.

    2002-01-01

    From the 1950s through the early 1990s, the U.S. Department of Energy (DOE) was the primary supplier of low enriched uranium (LEU) and highly enriched uranium (HEU) to research and test reactors worldwide. The formerly called Y-12 Plant in Oak Ridge, Tennessee, was put into operational stand down in 1994 due to inadequate safety documentation. This paper will discuss the re-start of the Y-12 Plant and its current capabilities. Additionally, the paper will address recent changes within the DOE, with the creation of the National Nuclear Security Administration (NNSA). It will show how the change to NNSA and an organizational re-alignment has improved efficiencies. NNSA is committed to operate its sales program so that it is complementary to, and in support of, the Reduced Enrichment for Research and Test Reactors (RERTR) and Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Return Programs. The NNSA is committed to provide an assurance of competitively-priced, high-quality uranium supply to the research and test reactor community under long-term contracts. This paper will discuss some of NNSA's recent successes in long-term contracting and meeting deliveries. (author)

  7. Fate of the Epsilon Phase in the Oklo Natural Reactors

    International Nuclear Information System (INIS)

    S. Utsunomiya; R.C. Ewing

    2005-01-01

    In spent nuclear fuel (SNF), the micron- to submicron-sized epsilon phase (Mo-Ru-Pd-Tc-Rh) is an important host of 99 Tc which has a long half life (2.13 x 10 5 years) and can be an important contributor to dose in safety assessments of nuclear waste repositories. In addition, Tc is predominantly present as TcO 4 - under oxidizing conditions at wide range of pH, weakly adsorbed onto mineral surfaces, and unlikely to be incorporated into alteration uranyl minerals. In the Oklo natural reactor (2.0 Ga), essentially all of the 99 Tc has decayed to 99 Ru. Thus, this study focuses on Ru and the other metals of the epsilon phase in order to investigate the occurrence and the fate of the epsilon phase during the corrosion of this natural SNF. Samples from reactor zone (RZ)-10 (836, 819, 687); from RZ-13 (864, 910); were investigated using TEM (transmission electron microscopy). Within the UO 2 matrix, a Bi-Pd particle (40-60 nm), fioodite, PdBi 2 , was observed with trace amounts of As, Fe, and Te surrounded by an amorphous Pb-rich area. (Pd,Rh) 2 As, palladodymite or rhodarsenide, was observed (400-500 nm in size). Ruthenarsinite, (Ru,Ni)As, was identified in most samples: with a representative composition of As, 59.9: Co, 2.5: Ni, 5.2; Ru, 18.6; Rh, 8.4; Pd, 3.1; Sb, 2.4 in atomic percent. The particles diameters are a few hundred nanometers and, in most cases, surrounded by a Pb-rich phase (400-500 nm). Typically, the ruthenarsenite does not occur as single particle but an aggregate of ∼200 nm-sized particles. Some Ru-particles revealed a complex phase separation within the grain such as a Ru-particle (600-700 nm) with Pb at the core of the particle and enrichment of Ni, Co, and As at the rim. Some ruthenarsenite crystals were embedded in chlorite immediately adjacent to uraninite. A few particles were still coated by Pb. These results suggest a history for the epsilon phases: (1) The original epsilon phase was transformed to, in most cases, ruthenarsenite. (2) All

  8. FBR type reactor

    International Nuclear Information System (INIS)

    Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

    1994-01-01

    In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

  9. The prototype fast reactor

    International Nuclear Information System (INIS)

    Broomfield, A.M.

    1985-01-01

    The paper concerns the Prototype Fast Reactor (PFR), which is a liquid metal cooled fast reactor power station, situated at Dounreay, Scotland. The principal design features of a Fast Reactor and the PFR are given, along with key points of operating history, and health and safety features. The role of the PFR in the development programme for commercial reactors is discussed. (U.K.)

  10. Department of reactor technology

    International Nuclear Information System (INIS)

    1980-01-01

    The activities of the Department of Reactor Technology at Risoe during 1979 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

  11. NCSU Reactor Sharing Program

    International Nuclear Information System (INIS)

    Perez, P.B.

    1993-01-01

    The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities

  12. Reactor safety method

    International Nuclear Information System (INIS)

    Vachon, L.J.

    1980-01-01

    This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature

  13. Nuclear shape changes are induced by knockdown of the SWI/SNF ATPase BRG1 and are independent of cytoskeletal connections.

    Directory of Open Access Journals (Sweden)

    Karen M Imbalzano

    Full Text Available Changes in nuclear morphology occur during normal development and have been observed during the progression of several diseases. The shape of a nucleus is governed by the balance of forces exerted by nuclear-cytoskeletal contacts and internal forces created by the structure of the chromatin and nuclear envelope. However, factors that regulate the balance of these forces and determine nuclear shape are poorly understood. The SWI/SNF chromatin remodeling enzyme ATPase, BRG1, has been shown to contribute to the regulation of overall cell size and shape. Here we document that immortalized mammary epithelial cells show BRG1-dependent nuclear shape changes. Specifically, knockdown of BRG1 induced grooves in the nuclear periphery that could be documented by cytological and ultrastructural methods. To test the hypothesis that the observed changes in nuclear morphology resulted from altered tension exerted by the cytoskeleton, we disrupted the major cytoskeletal networks and quantified the frequency of BRG1-dependent changes in nuclear morphology. The results demonstrated that disruption of cytoskeletal networks did not change the frequency of BRG1-induced nuclear shape changes. These findings suggest that BRG1 mediates control of nuclear shape by internal nuclear mechanisms that likely control chromatin dynamics.

  14. Minor Actinide Transmutation Physics for Low Conversion Ratio Sodium Fast Reactors

    International Nuclear Information System (INIS)

    Mehdi Asgari; Samuel E. Bays; Benoit Forget; Rodolfo Ferrer

    2007-01-01

    The effects of varying the reprocessing strategy used in the closed cycle of a Sodium Fast Reactor (SNF) prototype are presented in this paper. The isotopic vector from the aqueous separation of transuranic (TRU) elements in Light Water Reactor (LWR) spent nuclear fuel (SNF) is assumed to also vary according to the reprocessing strategy of the closed fuel cycle. The decay heat, gamma energy, and neutron emission of the fuel discharge at equilibrium are found to vary depending on the separation strategy. The SFR core used in this study corresponds to a burner configuration with a conversion ratio of ∼0.5 based on the Super-PRISM design. The reprocessing strategies stemming from the choice of either metal or oxide fuel for the SFR are found to have a large impact on the equilibrium discharge decay heat, gamma energy, and neutron emission. Specifically, metal fuel SFR with pyroprocessing of the discharge produces the largest amount of TRU consumption (166 kg per Effective Full Power Year or EFPY), but also the highest decay heat, gamma energy, and neutron emission. On the other hand, an oxide fuel SFR with PUREX reprocessing minimizes the decay heat and related parameters of interest to a minimum, even when compared to thermal Mixed Oxide (MOX) or Inert Matrix Fuel (IMF) on a per mass basis. On an assembly basis, however, the metal SFR discharge has a lower decay heat than an equivalent oxide SFR assembly for similar minor actinide consumptions (∼160 kg/EFPY.) Another disadvantage in the oxide PUREX reprocessing scenario is that there is no consumption of americium and curium, since PUREX reprocessing separates these minor actinides (MA) and requires them to be disposed of externally

  15. Physics of nuclear reactors

    International Nuclear Information System (INIS)

    Baeten, Peter

    2006-01-01

    This course gives an introduction to Nuclear Reactor Physics. The first chapter explains the most important parameters and concepts in nuclear reactor physics such as fission, cross sections and the effective multiplication factor. Further on, in the second chapter, the flux distributions in a stationary reactor are derived from the diffusion equation. Reactor kinetics, reactor control and reactor dynamics (feedback effects) are described in the following three chapters. The course concludes with a short description of the different types of existing and future reactors. (author)

  16. Reactor core and initially loaded reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Koyama, Jun-ichi; Aoyama, Motoo.

    1989-01-01

    In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

  17. 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.)

  18. Thermonuclear reactor

    International Nuclear Information System (INIS)

    Yasutomi, Yoshiyuki; Nakagawa, Moroo; Sawai, Yuichi; Chiba, Akio; Suzuki, Yasutaka.

    1997-01-01

    Silicon composited with reinforcing metals is used for a divertor cooling substrate having an effect as a cooling tube to provide a silicon base composite material having increased electric resistance and toughness. The blending ratio of reinforcing materials in the form of granules, whiskers or long fibers is controlled in order to control heat conductivity, electric resistivity and mechanical performances. The divertor cooling substrate comprising the silicon base composite material is integrated with a plasma facing material. The production method therefor includes ordinary metal matrix composite forming methods such as powder metallurgy, melting penetration method, high pressure solidification casting method, centrifugal casting method and vacuum casting method. Since the cooling plate is constituted with the light metal and highly electric resistant metal base composite material, sharing force due to eddy current can be reduced, and radiation exposure can be minimized. Accordingly, a cooling structure for a thermonuclear reactor effective for the improvement of environmental problems caused by waste disposal can be attained. (N.H.)

  19. Nuclear reactor

    International Nuclear Information System (INIS)

    Irion, L.; Tautz, J.; Ulrych, G.

    1976-01-01

    This additional patent complements the arrangement of non-return valves to prevent loss of cooling water on fracture of external tubes in the main coolant circuit (according to PS 24 24 427.7) by ensuring that the easily movable valves only operate in case of a fault, but do not flutter in operation, because the direction of flow is not the same at each location where they are installed. The remedy for this undesirable effect consists of allocating 1 non-return valve unit with 5 to 10 valves to each (of several) ducts for the cooling water intake. These units are installed in the annular space between the reactor vessel and the pressure vessel below the inlet of the ducts. Due to flow guidance surfaces in the same space, the incoming cooling water is deflected downwards and as the guiding surfaces are closed at the sides, must pass parallel to the valves of the non-return valve unit. On fracture of the external cooling water inlet pipe concerned, all valves of this unit close due to reversal of flow on the outlet side. (TK) [de

  20. Nuclear reactors

    International Nuclear Information System (INIS)

    Pearson, K.G.

    1977-01-01

    Reference is made to auxiliary means of cooling the nuclear fuel clusters used in light or heavy water cooled nuclear reactors. One method is to provide one or more spray cooling tubes. From holes in the side walls of those tubes coolant water may be sprayed laterally into the cluster against the rods. The flow of main coolant may thus be supplemented or even replaced by the auxiliary coolant. A difficulty, however, is that only those fuel rods close to a spray cooling tube can readily be reached by the auxiliary coolant. In the arrangement described, where the fuel rods are spaced apart by transverse grids, at least one of the interspaces between the grids is provided with an axially extending auxiliary coolant conduit having lateral holes through which an auxiliary coolant is sprayed into the cluster. A deflector is provided that extends from a transverse grid into a position in front of the holes and deflects auxiliary coolant on to parts of the fuel rods otherwise inaccessible to the auxiliary coolant. The construction of the deflector is described. (U.K.)

  1. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

  2. Fast breeder reactors

    International Nuclear Information System (INIS)

    Heinzel, V.

    1975-01-01

    The author gives a survey of 'fast breeder reactors'. In detail the process of breeding, the reasons for the development of fast breeders, the possible breeder reactors, the design criteria, fuels, cladding, coolant, and safety aspects are reported on. Design data of some experimental reactors already in operation are summarized in stabular form. 300 MWe Prototype-Reactors SNR-300 and PFR are explained in detail and data of KWU helium-cooled fast breeder reactors are given. (HR) [de

  3. Reactor Physics Training

    International Nuclear Information System (INIS)

    Baeten, P.

    2007-01-01

    University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

  4. The fast breeder reactor

    International Nuclear Information System (INIS)

    Collier, J.

    1990-01-01

    The arguments for and against the fast breeder reactor are debated. The case for the fast reactor is that the world energy demand will increase due to increasing population over the next forty years and that the damage to the global environment from burning fossil fuels which contribute to the greenhouse effect. Nuclear fission is the only large scale energy source which can achieve a cut in the use of carbon based fuels although energy conservation and renewable sources will also be important. Fast reactors produce more energy from uranium than other types of (thermal) reactors such as AGRs and PWRs. Fast reactors would be important from about 2020 onwards especially as by then many thermal reactors will need to be replaced. Fast reactors are also safer than normal reactors. The arguments against fast reactors are largely economic. The cost, especially the capital cost is very high. The viability of the technology is also questioned. (UK)

  5. Nuclear reactor instrumentation at research reactor renewal

    International Nuclear Information System (INIS)

    Baers, B.; Pellionisz, P.

    1981-10-01

    The paper overviews the state-of-the-art of research reactor renewals. As a case study the instrumentation reconstruction of the Finnish 250 kW TRIGA reactor is described, with particular emphasis on the nuclear control instrumentation and equipment which has been developed and manufactured by the Central Research Institute for Physics, Budapest. Beside the presentation of the nuclear instrument family developed primarily for research reactor reconstructions, the quality assurance policy conducted during the manufacturing process is also discussed. (author)

  6. Safeguarding research reactors

    International Nuclear Information System (INIS)

    Powers, J.A.

    1983-03-01

    The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

  7. Guide to power reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1959-07-15

    The IAEA's major first scientific publication is the Directory of Power Reactors now in operation or under construction in various parts of the world. The purpose of the directory is to present important details of various power projects in such a way as to provide a source of easy reference for anyone interested in the development of the peaceful uses of atomic energy, either at the technical or management level. Six pages have been devoted to each reactor the first of which contains general information, reactor physics data and information about the core. The second and third contain sketches of the fuel element or of the fuel element assembly, and of the horizontal and vertical sections of the reactor. On the fourth page information is grouped under the following heads: fuel element, core heat transfer, control, reactor vessel and over-all dimensions, and fluid flow. The fifth page shows a simplified flow diagram, while the sixth provides information on reflector and shielding, containment and turbo generator. Some information has also been given, when available, on cost estimates and operating staff requirements. Remarks and a bibliography constitute the last part of the description of each reactor. Reactor projects included in this directory are pressurized light water cooled power reactors. Boiling light water cooled power reactors, heavy water cooled power reactors, gas cooled power reactors, organic cooled power reactors liquid metal cooled power reactors and liquid metal cooled power reactors

  8. 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.)

  9. 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.)

  10. Tokamak reactor studies

    International Nuclear Information System (INIS)

    Baker, C.C.

    1981-01-01

    This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features

  11. Links with NEA activities: Nuclear Data services and WPs

    International Nuclear Information System (INIS)

    Cabellos, O.

    2016-01-01

    During the last three years a significant number of international activities have been undertaken on Primary Radiation Damage (PRD): the Expert Group on Primary Radiation Damage belonging to the Working Party on Multi-scale Modelling of Fuels and Structural Materials for Nuclear Systems (WPMM- OECD/NEA/NSC), Coordinated Research Project (CRP) on “Primary Radiation Damage Cross Sections” (IAEA/NDS) and the activities of the European Consortium on Nuclear Data Development and Analysis for Fusion within the Fusion for Energy (F4E) Program. NEA activities related to fuels and structural materials are coordinated by the Working Party on Multi-scale Modelling of Fuels and Structural Materials for Nuclear Systems, WPMM. In May 2015, the Expert Group on “Primary Radiation Damage” reviewed the limitation of the NRT-dpa standard, and recommended a new improved standard of primary damage characteristics, the arc-dpa. Recently, the new Expert Group on Structural Materials Modelling (SMM) has been created to enhance efforts on the multiscale modelling approach applied to structural materials, in order to investigate phenomena that can not be studied experimentally, predicting how the nanostructure and the microchemistry changes under irradiation

  12. Inhibition of Early Stages of HIV-1 Assembly by INI1/hSNF5 Transdominant Negative Mutant S6 ▿

    Science.gov (United States)

    Cano, Jennifer; Kalpana, Ganjam V.

    2011-01-01

    INI1/hSNF5 is an HIV-1 integrase (IN) binding protein specifically incorporated into virions. A truncated mutant of INI1 (S6, amino acids 183 to 294) harboring the minimal IN binding Rpt1 domain potently inhibits HIV-1 particle production in a transdominant manner. The inhibition requires interaction of S6 with IN within Gag-Pol. While INI1 is a nuclear protein and harbors a masked nuclear export signal (NES), the transdominant negative mutant S6 is cytoplasmic, due to the unmasking of NES. Here, we examined the effects of subcellular localization of S6 on HIV-1 inhibition and further investigated the stages of assembly that are affected. We found that targeting a nuclear localization signal-containing S6 variant [NLS-S6(Rpt1)] to the nucleoplasm (but not to the nucleolus) resulted in complete reversal of inhibition of particle production. Electron microscopy indicated that although no electron-dense particles at any stage of assembly were seen in cells expressing S6, virions were produced in cells expressing the rescue mutant NLS-S6(Rpt1) to wild-type levels. Immunofluorescence analysis revealed that p24 exhibited a diffuse pattern of localization within the cytoplasm in cells expressing S6 in contrast to accumulation along the membrane in controls. Pulse-chase analysis indicated that in S6-expressing cells, although Gag(Pr55gag) protein translation was unaffected, processing and release of p24 were defective. Together, these results indicate that expression of S6 in the cytoplasm interferes with trafficking of Gag-Pol/Gag to the membrane and causes a defective processing leading to inhibition of assembly at an early stage prior to particle formation and budding. PMID:21159874

  13. Why Hospitals and Payers are Recommending Home Care Upon Discharge Instead of SNF or Traditional Home Health Services--Alternative Payment Model Hospital Incentives Aligning with Patient Choice.

    Science.gov (United States)

    Luke, Josh

    2016-01-01

    Seniors and other hospital patients in the United States have traditionally had the option of being discharged to a skilled nursing facility (convalescent home) for post-acute services, or home with nursing and therapy services provided in the home setting. Traditionally, these home based services have been referred to as "home health." As more Americans have retired, home health services have expanded and are readily accessible. This growth put tremendous stress on the Medicare fund which pays for senior care services. However, "Home Care," which traditionally has been viewed as non-medical home based services, has also become a booming industry for the cost conscious in recent years as more Americans reach retirement age. With the passing of the Affordable Care Act in 2010, providers and payers are now finding themselves responsible for post-acute care and continuous patient health, so cost efficient solutions for post-acute care are thriving. For the first time in history, American hospitals and Insurers are recognizing Home Care as an effective model that achieves the Triple Aim of Health Care reform. Home Care, which is no longer completely non-medical services, has proven to be an integral part of the care continuum for seniors in recent years and is now becoming a viable solution for keeping patients well, while still honoring their desire to age and heal at home. This paper analyzes the benefits and risks of home care and provides a clear understanding as to why American hospitals are emphasizing SNF Avoidance and skipping home health, opting instead to refer patients directly to home care as the preferred discharge solution in a value based model.

  14. Coffin-Siris syndrome and related disorders involving components of the BAF (mSWI/SNF) complex: historical review and recent advances using next generation sequencing.

    Science.gov (United States)

    Kosho, Tomoki; Miyake, Noriko; Carey, John C

    2014-09-01

    This issue of Seminars in Medical Genetics, American Journal of Medical Genetics Part C investigates the human diseases caused by mutations in the BAF complex (also known as the mammalian SWI/SNF complex) genes, particularly focusing on Coffin-Siris syndrome (CSS). CSS is a rare congenital malformation syndrome characterized by developmental delay or intellectual disability (ID), coarse facial appearance, feeding difficulties, frequent infections, and hypoplasia/aplasia of the fifth fingernails and fifth distal phalanges. In 2012, 42 years after the first description of CSS in 1970, five causative genes (SMARCB1, SMARCE1, SMARCA4, ARID1A, ARID1B), all encoding components of the BAF complex, were identified as being responsible for CSS through whole exome sequencing and pathway-based genetic screening. The identification of two additional causative genes (PHF6, SOX11) followed. Mutations in another BAF complex gene (SMARCA2) and (TBC1D24) were found to cause clinically similar conditions with ID, Nicolaides-Baraitser syndrome and DOORS syndrome, respectively. Also, ADNP was found to be mutated in an autism/ID syndrome. Furthermore, there is growing evidences for germline or somatic mutations in the BAF complex genes to be causal for cancer/cancer predisposition syndromes. These discoveries have highlighted the role of the BAF complex in the human development and cancer formation. The biology of BAF is very complicated and much remains unknown. Ongoing research is required to reveal the whole picture of the BAF complex in human development, and will lead to the development of new targeted therapies for related disorders in the future. © 2014 Wiley Periodicals, Inc.

  15. Reactor Physics Programme

    International Nuclear Information System (INIS)

    De Raedt, C.

    2000-01-01

    The Reactor Physics and 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 on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research 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. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

  16. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

    The Reactor Physics and 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 on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research 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. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

  17. Undergraduate reactor control experiment

    International Nuclear Information System (INIS)

    Edwards, R.M.; Power, M.A.; Bryan, M.

    1992-01-01

    A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise

  18. Reactor System Design

    International Nuclear Information System (INIS)

    Chi, S. K.; Kim, G. K.; Yeo, J. W.

    2006-08-01

    SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

  19. Fusion reactor design studies

    International Nuclear Information System (INIS)

    Emmert, G.A.; Kulcinski, G.L.; Santarius, J.F.

    1990-01-01

    This report discusses the following topics on the ARIES tokamak: systems; plasma power balance; impurity control and fusion ash removal; fusion product ripple loss; energy conversion; reactor fueling; first wall design; shield design; reactor safety; and fuel cost and resources

  20. Nuclear reactors; graphical symbols

    International Nuclear Information System (INIS)

    1987-11-01

    This standard contains graphical symbols that reveal the type of nuclear reactor and is used to design graphical and technical presentations. Distinguishing features for nuclear reactors are laid down in graphical symbols. (orig.) [de

  1. Control for nuclear reactor

    International Nuclear Information System (INIS)

    Ash, E.B.; Bernath, L.; Facha, J.V.

    1980-01-01

    A nuclear reactor is provided with several hydraulically-supported spherical bodies having a high neutron absorption cross section, which fall by gravity into the core region of the reactor when the flow of supporting fluid is shut off. (auth)

  2. Hybrid plasmachemical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lelevkin, V. M., E-mail: lelevkin44@mail.ru; Smirnova, Yu. G.; Tokarev, A. V. [Kyrgyz-Russian Slavic University (Kyrgyzstan)

    2015-04-15

    A hybrid plasmachemical reactor on the basis of a dielectric barrier discharge in a transformer is developed. The characteristics of the reactor as functions of the dielectric barrier discharge parameters are determined.

  3. Ship propulsion reactors technology

    International Nuclear Information System (INIS)

    Fribourg, Ch.

    2002-01-01

    This paper takes the state of the art on ship propulsion reactors technology. The french research programs with the corresponding technological stakes, the reactors specifications and advantages are detailed. (A.L.B.)

  4. Guidebook to nuclear reactors

    International Nuclear Information System (INIS)

    Nero, A.V. Jr.

    1976-05-01

    A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen

  5. continuous stirred tank reactor (CSTR)

    African Journals Online (AJOL)

    AFRICAN JOURNALS ONLINE (AJOL) · Journals · Advanced Search ... stirred tank reactor (CSTR) and the small and large intestines as plug flow reactor (PFR) ... from the two equations are used for the reactor sizing of the modeled reactors.

  6. Conception of transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel of power reactors, which meets the requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism

    International Nuclear Information System (INIS)

    Il'kaev, R.I.; Matveev, V.Z.; Morenko, A.I.; Shapovalov, V.I.; Semenov, A.G.; Sergeyev, V.M.; Orlov, V.K.; Shatalov, V.V.; Gotovchikov, V.T.; Seredenko, V.A.; Haire, Jonathan M.; Forsberg, C.W.

    2004-01-01

    The report is devoted to the problem of creation of a new generation of multi-purpose universal transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel (SNF) of power reactors, which meets all requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism. Meeting all IAEA requirements in terms of safety both in normal operation conditions and accidents, as well as increased stability of transport cask (TC) with SNF under the conditions of beyond-design-basis accidents and acts of terrorism has been achieved in the design of multi-purpose universal TC due to the use of DU (depleted uranium) in it. At that, it is suggested to use DU in TC, which acts as effective gamma shield and constructional material in the form of both metallic depleted uranium and metal-ceramic mixture (cermet), based on stainless or carbon steel and DU dioxide. The metal in the cermet is chosen to optimize cask performance. The use of DU in the design of multi-purpose universal TC enables getting maximum load of the container for spent nuclear fuel when meeting IAEA requirements in terms of safety and providing increased stability of the container with SNF under conditions of beyond-design-basis accident and acts of terrorism

  7. Conception of transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel of power reactors, which meets the requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism

    Energy Technology Data Exchange (ETDEWEB)

    Il' kaev, R.I.; Matveev, V.Z.; Morenko, A.I.; Shapovalov, V.I. [Russian Federal Nuclear Center - All-Russian Research Inst. of Experimental Physics, Sarov (Russian Federation); Semenov, A.G.; Sergeyev, V.M.; Orlov, V.K. [All-Russian Research Inst. of Inorganic Materials, Moscow (Russian Federation); Shatalov, V.V.; Gotovchikov, V.T.; Seredenko, V.A. [All-Russian Research Inst. of Applied Chemistry, Moscow (Russian Federation); Haire, Jonathan M.; Forsberg, C.W. [Oak Ridge National Lab., Oak Ridge (United States)

    2004-07-01

    The report is devoted to the problem of creation of a new generation of multi-purpose universal transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel (SNF) of power reactors, which meets all requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism. Meeting all IAEA requirements in terms of safety both in normal operation conditions and accidents, as well as increased stability of transport cask (TC) with SNF under the conditions of beyond-design-basis accidents and acts of terrorism has been achieved in the design of multi-purpose universal TC due to the use of DU (depleted uranium) in it. At that, it is suggested to use DU in TC, which acts as effective gamma shield and constructional material in the form of both metallic depleted uranium and metal-ceramic mixture (cermet), based on stainless or carbon steel and DU dioxide. The metal in the cermet is chosen to optimize cask performance. The use of DU in the design of multi-purpose universal TC enables getting maximum load of the container for spent nuclear fuel when meeting IAEA requirements in terms of safety and providing increased stability of the container with SNF under conditions of beyond-design-basis accident and acts of terrorism.

  8. Reactor physics aspects of CANDU reactors

    International Nuclear Information System (INIS)

    Critoph, E.

    1980-01-01

    These four lectures are being given at the Winter Course on Nuclear Physics at Trieste during 1978 February. They constitute part of the third week's lectures in Part II: Reactor Theory and Power Reactors. A physical description of CANDU reactors is given, followed by an overview of CANDU characteristics and some of the design options. Basic lattice physics is discussed in terms of zero energy lattice experiments, irradiation effects and analytical methods. Start-up and commissioning experiments in CANDU reactors are reviewed, and some of the more interesting aspects of operation discussed - fuel management, flux mapping and control of the power distribution. Finally, some of the characteristics of advanced fuel cycles that have been proposed for CANDU reactors are summarized. (author)

  9. Molten salt breeder reactor

    International Nuclear Information System (INIS)

    1977-01-01

    MSBR Study Group formed in October 1974 has studied molten salt breeder reactor and its various aspects. Usage of a molten salt fuel, extremely interesting as reactor chemistry, is a great feature to MSBR; there is no need for separate fuel making, reprocessing, waste storage facilities. The group studied the following, and these results are presented: molten salt technology, molten salt fuel chemistry and reprocessing, reactor characteristics, economy, reactor structural materials, etc. (Mori, K.)

  10. Zero energy reactor 'RB'

    Energy Technology Data Exchange (ETDEWEB)

    Popovic, D; Takac, S; Markovic, H; Raisic, N; Zdravkovic, Z; Radanovic, Lj [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

    1959-03-15

    In 1958 the zero energy reactor RB was built with the purpose of enabling critical experiments with various reactor systems to be carried out. The first core assembly built in this reactor consists of heavy water as moderator and natural uranium metal as fuel. In order to be able to obtain very accurate results when measuring the main characteristics of the assembly the reactor was built as a completely bare system. (author)

  11. Reactor utilization, Annex A

    International Nuclear Information System (INIS)

    Martinc, R.; Stanic, A.

    1984-01-01

    Reactor was operated until August 1984 due to prohibition issued by the Ministry since the reactor does not have the emergency cooling system nor special filters in the ventilation system yet. This means that the operation plan was fulfilled by 69%. This annex includes detailed tables containing data about utilization of reactor experimental channels, irradiated samples, as well as interruptions of operation. Detailed data about reactor power during this period are shown as well

  12. PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Masood, Z.

    2016-01-01

    The PUSPATI TRIGA Reactor is the only research reactor in Malaysia. This 1 MW TRIGA Mk II reactor first reached criticality on 28 June 1982 and is located at the Malaysian Nuclear Agency premise in Bangi, Malaysia. This reactor has been mainly utilised for research, training and education and isotope production. Over the years several systems have been refurbished or modernised to overcome ageing and obsolescence problems. Major achievements and milestones will also be elaborated in this paper. (author)

  13. Nuclear reactor shutdown system

    International Nuclear Information System (INIS)

    Mangus, J.D.; Cooper, M.H.

    1982-01-01

    An improved nuclear reactor shutdown system is described comprising a temperature sensitive device connected to control the electric power supply to a magnetic latch holding a body of a neutron absorbing material. The temperature sensitive device is exposed to the reactor coolant so that when the reactor coolant temperature rises above a specific level, the temperature sensitive device will cause deenergization of the magnetic latch to allow the body of neutron absorbing material to enter the reactor core. (author)

  14. The fast reactor

    International Nuclear Information System (INIS)

    1980-02-01

    The subject is discussed as follows: brief description of fast reactors; advantage in conserving uranium resources; experience, in UK and elsewhere, in fast reactor design, construction and operation; safety; production of plutonium, security aspects; consideration of future UK fast reactor programme. (U.K.)

  15. Mirror fusion reactor design

    International Nuclear Information System (INIS)

    Neef, W.S. Jr.; Carlson, G.A.

    1979-01-01

    Recent conceptual reactor designs based on mirror confinement are described. Four components of mirror reactors for which materials considerations and structural mechanics analysis must play an important role in successful design are discussed. The reactor components are: (a) first-wall and thermal conversion blanket, (b) superconducting magnets and their force restraining structure, (c) neutral beam injectors, and (d) plasma direct energy converters

  16. Towards nuclear fusion reactors

    International Nuclear Information System (INIS)

    1993-11-01

    The results of nuclear fusion researches in JAERI are summarized. In this report, following themes are collected: the concept of fusion reactor (including ITER), fusion reactor safety, plasma confinement, fusion reactor equipment, and so on. Includes glossary. (J.P.N.)

  17. Rotating reactors : a review

    NARCIS (Netherlands)

    Visscher, F.; Schaaf, van der J.; Nijhuis, T.A.; Schouten, J.C.

    2013-01-01

    This review-perspective paper describes the current state-of-the-art in the field of rotating reactors. The paper has a focus on rotating reactor technology with applications at lab scale, pilot scale and industrial scale. Rotating reactors are classified and discussed according to their geometry:

  18. Refuelling nuclear reactors

    International Nuclear Information System (INIS)

    Stacey, J.; Webb, J.; White, W.P.; McLaren, N.H.

    1981-01-01

    An improved nuclear reactor refuelling machine is described which can be left in the reactor vault to reduce the off-load refuelling time for the reactor. The system comprises a gripper device rangeable within a tubular chute, the gripper device being movable by a pantograph. (U.K.)

  19. Reactor power measuring device

    International Nuclear Information System (INIS)

    Izumi, Mikio; Sano, Yuji; Seki, Eiji; Yoshida, Toshifumi; Ito, Toshiaki.

    1993-01-01

    The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

  20. Ulysse, mentor reactor

    International Nuclear Information System (INIS)

    Bouquin, B.; Rio, I.; Safieh, J.

    1997-01-01

    On July 23, 1961, the ULYSSE reactor began its first power rise. Designed at that time to train nuclear engineering students and reactor operators, this reactor still remains an indispensable tool for nuclear teaching and a choice instrument for scientists. (author)

  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. Draft environmental impact statement on a proposed nuclear weapons nonproliferation policy concerning foreign research reactor spent nuclear fuel: Volume 2, Appendix E, Evaluation of human health effects of overland transportation

    International Nuclear Information System (INIS)

    1995-03-01

    This Appendix provides an overview of the approach used to assess the human health risks that may result from the overland transportation of foreign research reactor spent nuclear fuel. The Appendix includes discussion of the scope of the assessment, analytical methods used for the risk assessment (i.e., computer models), important assessment assumptions, determination of potential transportation routes, and presents the results of the assessment. In addition, to aid in the understanding and interpretation of the results, specific arm of uncertainty are described, with an emphasis an how the uncertainties may affect comparisons of the alternatives. he approach used in this Appendix is modeled after that used in the Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Draft Environmental Impact Statement (SNF ampersand INEL Draft EIS) (DOE, 1994b). The SNF ampersand INEL Draft EIS did not perform as detailed an analysis on the specific actions taken for foreign research reactor spent nuclear fuel because of the breadth necessary to analyze the entire spent fuel management program. However, the fundamental assumptions used in this analysis are consistent with those used in the SNF ampersand INEL Draft EIS (DOE, 1994b), and the same computer codes and generic release and accident data are used. The risk assessment results are presented in this Appendix in terms of ''Per-shipment'' risk factors, as well as for the total risks associated with each alternative. Per-shipment risk factors provide an estimate of the risk from a single spent nuclear fuel shipment between a specific origin and destination. They are calculated for all possible origin and destination pairs for each spent nuclear fuel type. The total risks for a given alternative are found by multiplying the expected number of shipments by the appropriate per-shipment risk factors. This approach provides maximum flexibility for determining the risks for a large number of potential

  4. 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)

  5. Reactor as furnace and reactor as lamp

    International Nuclear Information System (INIS)

    Goldanskii, V.I.

    1992-01-01

    There are presented general characteristics of the following ways of transforming of nuclear energy released in reactors into chemical : ordinary way (i.e. trough the heat, mechanical energy and electricity); chemonuclear synthesis ; use of high-temperature fuel elements (reactor as furnace); use of the mixed nγ-radiation of reactors; use of the radiation loops; radiation - photochemical synthesis (reactor as lamp). Advantage and disadvantages of all above variants are compared. The yield of the primary product of fixation of nitrogen (nitric oxide NO) in reactor with the high-temperature (above ca. 1900degC) fuel elements (reactor-furnace) can exceed W ∼ 200 kg per gram of burned uranium. For the latter variant (reactor-lamp) the yield of chemical products can reach W ∼ 60 kg. per gram of uranium. Such values of W are close to or even strongly exceed the yields of chemical products for other abovementioned variants and - what is particularly important - are not connected to the necessity of archscrupulous removal of radioactive contamination of products. (author)

  6. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  7. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  8. Fundamentals of reactor chemistry

    International Nuclear Information System (INIS)

    Akatsu, Eiko

    1981-12-01

    In the Nuclear Engineering School of JAERI, many courses are presented for the people working in and around the nuclear reactors. The curricula of the courses contain also the subject material of chemistry. With reference to the foreign curricula, a plan of educational subject material of chemistry in the Nuclear Engineering School of JAERI was considered, and the fundamental part of reactor chemistry was reviewed in this report. Since the students of the Nuclear Engineering School are not chemists, the knowledge necessary in and around the nuclear reactors was emphasized in order to familiarize the students with the reactor chemistry. The teaching experience of the fundamentals of reactor chemistry is also given. (author)

  9. Generation III+ Reactor Portfolio

    International Nuclear Information System (INIS)

    2010-03-01

    While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO 2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPR TM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1 TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENA TM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENA TM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

  10. The fast breeder reactor

    International Nuclear Information System (INIS)

    Davis, D.A.; Baker, M.A.W.; Hall, R.S.

    1990-01-01

    Following submission of written evidence, the Energy Committee members asked questions of three witnesses from the Central Electricity Generating Board and Nuclear Electric (which will be the government owned company running nuclear power stations after privatisation). Both questions and answers are reported verbatim. The points raised include where the responsibility for the future fast reactor programme should lie, with government only or with private enterprise or both and the viability of fast breeder reactors in the future. The case for the fast reactor was stated as essentially strategic not economic. This raised the issue of nuclear cost which has both a construction and a decommissioning element. There was considerable discussion as to the cost of building a European Fast reactor and the cost of the electricity it would generate compared with PWR type reactors. The likely demand for fast reactors will not arrive for 20-30 years and the need to build a fast reactor now is questioned. (UK)

  11. 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.)

  12. Reactor power control device

    International Nuclear Information System (INIS)

    Ishii, Yoshihiko; Arita, Setsuo; Miyamoto, Yoshiyuki; Fukazawa, Yukihisa; Ishii, Kazuhiko

    1998-01-01

    The present invention provides a reactor power control device capable of enhancing an operation efficiency while keeping high reliability and safety in a BWR type nuclear power plant. Namely, the device of the present invention comprises (1) a means for inputting a set value of a generator power and a set value of a reactor power, (2) a means for controlling the reactor power to either smaller one of the reactor power corresponding to the set value of the generator power and the set value of the reactor power. With such procedures, even if the nuclear power plant is set so as to operate it to make the reactor power 100%, when the generator power reaches the upper limit, the reactor power is controlled with a preference given to the upper limit value of the generator power. Accordingly, safety and reliability are not deteriorated. The operation efficiency of the plant can be improved. (I.S.)

  13. Reactor power monitoring device

    International Nuclear Information System (INIS)

    Dogen, Ayumi; Ozawa, Michihiro.

    1983-01-01

    Purpose: To significantly improve the working efficiency of a nuclear reactor by reflecting the control rod history effect on thermal variants required for the monitoring of the reactor operation. Constitution: An incore power distribution calculation section reads the incore neutron fluxes detected by neutron detectors disposed in the reactor to calculate the incore power distribution. A burnup degree distribution calculation section calculates the burnup degree distribution in the reactor based on the thus calculated incore power distribution. A control rod history date store device supplied with the burnup degree distribution renews the stored control rod history data based on the present control rod pattern and the burnup degree distribution. Then, thermal variants of the nuclear reactor are calculated based on the thus renewed control rod history data. Since the control rod history effect is reflected on the thermal variants required for the monitoring of the reactor operation, the working efficiency of the nuclear reactor can be improved significantly. (Seki, T.)

  14. The Maple reactor project

    International Nuclear Information System (INIS)

    Malkoske, G.R.; Labrie, J.-P.

    2003-01-01

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

  15. Fission reactors and materials

    International Nuclear Information System (INIS)

    Frost, B.R.T.

    1981-12-01

    The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions

  16. Medicare Program; Prospective Payment System and Consolidated Billing for Skilled Nursing Facilities for FY 2018, SNF Value-Based Purchasing Program, SNF Quality Reporting Program, Survey Team Composition, and Correction of the Performance Period for the NHSN HCP Influenza Vaccination Immunization Reporting

    Science.gov (United States)

    2017-08-04

    This final rule updates the payment rates used under the prospective payment system (PPS) for skilled nursing facilities (SNFs) for fiscal year (FY) 2018. It also revises and rebases the market basket index by updating the base year from 2010 to 2014, and by adding a new cost category for Installation, Maintenance, and Repair Services. The rule also finalizes revisions to the SNF Quality Reporting Program (QRP), including measure and standardized resident assessment data policies and policies related to public display. In addition, it finalizes policies for the Skilled Nursing Facility Value-Based Purchasing Program that will affect Medicare payment to SNFs beginning in FY 2019. The final rule also clarifies the regulatory requirements for team composition for surveys conducted for investigating a complaint and aligns regulatory provisions for investigation of complaints with the statutory requirements. The final rule also finalizes the performance period for the National Healthcare Safety Network (NHSN) Healthcare Personnel (HCP) Influenza Vaccination Reporting Measure included in the End-Stage Renal Disease (ESRD) Quality Incentive Program (QIP) for Payment Year 2020.

  17. Summary engineering description of underwater fuel storage facility for foreign research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Dahlke, H.J.; Johnson, D.A.; Rawlins, J.K.; Searle, D.K.; Wachs, G.W.

    1994-10-01

    This document is a summary description for an Underwater Fuel Storage Facility (UFSF) for foreign research reactor (FRR) spent nuclear fuel (SNF). A FRR SNF environmental Impact Statement (EIS) is being prepared and will include both wet and dry storage facilities as storage alternatives. For the UFSF presented in this document, a specific site is not chosen. This facility can be sited at any one of the five locations under consideration in the EIS. These locations are the Idaho National Engineering Laboratory, Savannah River Site, Hanford, Oak Ridge National Laboratory, and Nevada Test Site. Generic facility environmental impacts and emissions are provided in this report. A baseline fuel element is defined in Section 2.2, and the results of a fission product analysis are presented. Requirements for a storage facility have been researched and are summarized in Section 3. Section 4 describes three facility options: (1) the Centralized-UFSF, which would store the entire fuel element quantity in a single facility at a single location, (2) the Regionalized Large-UFSF, which would store 75% of the fuel element quantity in some region of the country, and (3) the Regionalized Small-UFSF, which would store 25% of the fuel element quantity, with the possibility of a number of these facilities in various regions throughout the country. The operational philosophy is presented in Section 5, and Section 6 contains a description of the equipment. Section 7 defines the utilities required for the facility. Cost estimates are discussed in Section 8, and detailed cost estimates are included. Impacts to worker safety, public safety, and the environment are discussed in Section 9. Accidental releases are presented in Section 10. Standard Environmental Impact Forms are included in Section 11.

  18. Spent nuclear fuel project recommended reaction rate constants for corrosion of N-Reactor fuel

    International Nuclear Information System (INIS)

    Cooper, T.D.; Pajunen, A.L.

    1998-01-01

    The US Department of Energy (DOE) established the Spent Nuclear Fuel Project (SNF Project) to address safety and environmental concerns associated with deteriorating spent nuclear fuel presently stored in the Hanford Site's K Basins. The SNF Project has been tasked by the DOE with moving the spent N-Reactor fuel from wet storage to contained dry storage in order to reduce operating costs and environmental hazards. The chemical reactivity of the fuel must be understood at each process step and during long-term dry storage. Normally, the first step would be to measure the N-fuel reactivity before attempting thermal-hydraulic transfer calculations; however, because of the accelerated project schedule, the initial modeling was performed using literature values for uranium reactivity. These literature values were typically found for unirradiated, uncorroded metal. It was fully recognized from the beginning that irradiation and corrosion effects could cause N-fuel to exhibit quite different reactivities than those commonly found in the literature. Even for unirradiated, uncorroded uranium metal, many independent variables affect uranium metal reactivity resulting in a wide scatter of data. Despite this wide reactivity range, it is necessary to choose a defensible model and estimate the reactivity range of the N-fuel until actual reactivity can be established by characterization activities. McGillivray, Ritchie, and Condon developed data and/or models that apply for certain samples over limited temperature ranges and/or reaction conditions (McGillivray 1994, Ritchie 1981 and 1986, and Condon 1983). These models are based upon small data sets and have relatively large correlation coefficients

  19. Reactor Vessel Surveillance Program for Advanced Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

    2008-10-15

    This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

  20. Research reactors in Argentina

    International Nuclear Information System (INIS)

    Carlos Ruben Calabrese

    1999-01-01

    Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

  1. Siting of research reactors

    International Nuclear Information System (INIS)

    1987-01-01

    The purpose of this document is to develop criteria for siting and the site-related design basis for research reactors. The concepts presented in this document are intended as recommendations for new reactors and are not suggested for backfitting purposes for facilities already in existence. In siting research reactors serious consideration is given to minimizing the effects of the site on the reactor and the reactor on the site and the potential impact of the reactor on the environment. In this document guidance is first provided on the evaluation of the radiological impact of the installation under normal reactor operation and accident conditions. A classification of research reactors in groups is then proposed, together with a different approach for each group, to take into account the relevant safety problems associated with facilities of different characteristics. Guidance is also provided for both extreme natural events and for man-induced external events which could affect the safe operation of the reactor. Extreme natural events include earthquakes, flooding for river or coastal sites and extreme meteorological phenomena. The feasibility of emergency planning is finally considered for each group of reactors

  2. Reactor core for FBR type reactor

    International Nuclear Information System (INIS)

    Fujita, Tomoko; Watanabe, Hisao; Kasai, Shigeo; Yokoyama, Tsugio; Matsumoto, Hiroshi.

    1996-01-01

    In a gas-sealed assembly for a FBR type reactor, two or more kinds of assemblies having different eigen frequency and a structure for suppressing oscillation of liquid surface are disposed in a reactor core. Coolant introduction channels for introducing coolants from inside and outside are disposed in the inside of structural members of an upper shielding member to form a shielding member-cooling structure in the reactor core. A structure for promoting heat conduction between a sealed gas in the assembly and coolants at the inner side or the outside of the assembly is disposed in the reactor core. A material which generates heat by neutron irradiation is disposed in the assembly to heat the sealed gases positively by radiation heat from the heat generation member also upon occurrence of power elevation-type event to cause temperature expansion. Namely, the coolants flown out from or into the gas sealed-assemblies cause differential fluctuation on the liquid surface, and the change of the capacity of a gas region is also different on every gas-sealed assemblies thereby enabling to suppress fluctuation of the reactor power. Pressure loss is increased by a baffle plate or the like to lower the liquid surface of the sodium coolants or decrease the elevating speed thereof thereby suppressing fluctuation of the reactor power. (N.H.)

  3. Thai research reactor

    International Nuclear Information System (INIS)

    Aramrattana, M.

    1987-01-01

    The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

  4. Nuclear Reactor Physics

    Science.gov (United States)

    Stacey, Weston M.

    2001-02-01

    An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

  5. COMMERCIAL SNF ACCIDENT RELEASE FRACTIONS

    Energy Technology Data Exchange (ETDEWEB)

    S.O. Bader

    1999-10-18

    The purpose of this design analysis is to specify and document the total and respirable fractions for radioactive materials that are released from an accident event at the Monitored Geologic Repository (MGR) involving commercial spent nuclear fuel (CSNF) in a dry environment. The total and respirable release fractions will be used to support the preclosure licensing basis for the MGR. The total release fraction is defined as the fraction of total CSNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. The radionuclides are released from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses. This subset of the total release fraction is referred to as the respirable release fraction. Potential accidents may involve waste forms that are characterized as either bare (unconfined) fuel assemblies or confined fuel assemblies. The confined CSNF assemblies at the MGR are contained in shipping casks, canisters, or disposal containers (waste packages). In contrast to the bare fuel assemblies, the container that confines the fuel assemblies has the potential of providing an additional barrier for diminishing the total release fraction should the fuel rod cladding breach during an accident. However, this analysis will not take credit for this additional bamer and will establish only the total release fractions for bare unconfined CSNF assemblies, which may however be conservatively applied to confined CSNF assemblies.

  6. COMMERCIAL SNF ACCIDENT RELEASE FRACTIONS

    International Nuclear Information System (INIS)

    S.O. Bader

    1999-01-01

    The purpose of this design analysis is to specify and document the total and respirable fractions for radioactive materials that are released from an accident event at the Monitored Geologic Repository (MGR) involving commercial spent nuclear fuel (CSNF) in a dry environment. The total and respirable release fractions will be used to support the preclosure licensing basis for the MGR. The total release fraction is defined as the fraction of total CSNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. The radionuclides are released from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses. This subset of the total release fraction is referred to as the respirable release fraction. Potential accidents may involve waste forms that are characterized as either bare (unconfined) fuel assemblies or confined fuel assemblies. The confined CSNF assemblies at the MGR are contained in shipping casks, canisters, or disposal containers (waste packages). In contrast to the bare fuel assemblies, the container that confines the fuel assemblies has the potential of providing an additional barrier for diminishing the total release fraction should the fuel rod cladding breach during an accident. However, this analysis will not take credit for this additional bamer and will establish only the total release fractions for bare unconfined CSNF assemblies, which may however be conservatively applied to confined CSNF assemblies

  7. Reactor containment and reactor safety in the United States

    International Nuclear Information System (INIS)

    Kouts, H.

    1986-01-01

    The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.) [pt

  8. ATM and SIRT6/SNF2H Mediate Transient H2AX Stabilization When DSBs Form by Blocking HUWE1 to Allow Efficient γH2AX Foci Formation

    Directory of Open Access Journals (Sweden)

    Yuko Atsumi

    2015-12-01

    Full Text Available In response to DNA double-strand breaks (DSBs, H2AX is rapidly phosphorylated at Ser139 to promote DSB repair. Here we show that H2AX is rapidly stabilized in response to DSBs to efficiently generate γH2AX foci. This mechanism operated even in quiescent cells that barely expressed H2AX. H2AX stabilization resulted from the inhibition of proteasome-mediated degradation. Synthesized H2AX ordinarily underwent degradation through poly-ubiquitination mediated by the E3 ligase HUWE1; however, H2AX ubiquitination was transiently halted upon DSB formation. Such rapid H2AX stabilization by DSBs was associated with chromatin incorporation of H2AX and halting of its poly-ubiquitination mediated by the ATM kinase, the sirtuin protein SIRT6, and the chromatin remodeler SNF2H. H2AX Ser139, the ATM phosphorylation site, was essential for H2AX stabilization upon DSB formation. Our results reveal a pathway controlled by ATM, SIRT6, and SNF2H to block HUWE1, which stabilizes H2AX and induces its incorporation into chromatin only when cells are damaged.

  9. HAB1–SWI3B Interaction Reveals a Link between Abscisic Acid Signaling and Putative SWI/SNF Chromatin-Remodeling Complexes in Arabidopsis[C][W

    Science.gov (United States)

    Saez, Angela; Rodrigues, Americo; Santiago, Julia; Rubio, Silvia; Rodriguez, Pedro L.

    2008-01-01

    Abscisic acid (ABA) has an important role for plant growth, development, and stress adaptation. HYPERSENSITIVE TO ABA1 (HAB1) is a protein phosphatase type 2C that plays a key role as a negative regulator of ABA signaling; however, the molecular details of HAB1 action in this process are not known. A two-hybrid screen revealed that SWI3B, an Arabidopsis thaliana homolog of the yeast SWI3 subunit of SWI/SNF chromatin-remodeling complexes, is a prevalent interacting partner of HAB1. The interaction mapped to the N-terminal half of SWI3B and required an intact protein phosphatase catalytic domain. Bimolecular fluorescence complementation and coimmunoprecipitation assays confirmed the interaction of HAB1 and SWI3B in the nucleus of plant cells. swi3b mutants showed a reduced sensitivity to ABA-mediated inhibition of seed germination and growth and reduced expression of the ABA-responsive genes RAB18 and RD29B. Chromatin immunoprecipitation experiments showed that the presence of HAB1 in the vicinity of RD29B and RAB18 promoters was abolished by ABA, which suggests a direct involvement of HAB1 in the regulation of ABA-induced transcription. Additionally, our results uncover SWI3B as a novel positive regulator of ABA signaling and suggest that HAB1 modulates ABA response through the regulation of a putative SWI/SNF chromatin-remodeling complex. PMID:19033529

  10. Multipurpose research reactors

    International Nuclear Information System (INIS)

    1988-01-01

    The international symposium on the utilization of multipurpose research reactors and related international co-operation was organized by the IAEA to provide for information exchange on current uses of research reactors and international co-operative projects. The symposium was attended by about 140 participants from 36 countries and two international organizations. There were 49 oral presentations of papers and 24 poster presentations. The presentations were divided into 7 sessions devoted to the following topics: neutron beam research and applications of neutron scattering (6 papers and 1 poster), reactor engineering (6 papers and 5 posters), irradiation testing of fuel and material for fission and fusion reactors (6 papers and 10 posters), research reactor utilization programmes (13 papers and 4 posters), neutron capture therapy (4 papers), neutron activation analysis (3 papers and 4 posters), application of small reactors in research and training (11 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

  11. TRIGA reactor main systems

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

    This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

  12. 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...

  13. Upgradation of Apsara reactor

    International Nuclear Information System (INIS)

    Mammen, S.; Mukherjee, P.; Bhatnagar, A.; Sasidharan, K.; Raina, V.K.

    2009-01-01

    Apsara is a 1 MW swimming pool type research reactor using high enriched uranium as fuel with light water as coolant and moderator. The reactor is in operation for more than five decades and has been extensively used for basic research, radioisotope production, neutron radiography, detector testing, shielding experiments etc. In view of its long service period, it is planned to carry out refurbishment of the reactor to extend its useful life. During refurbishment, it is also planned to upgrade the reactor to a 2 MW reactor to improve its utilization and to upgrade the structure, system and components in line with the current safety standards. This paper gives a brief account of the design features and safety aspects of the upgraded Apsara reactor. (author)

  14. Methodology and techniques of early detection anomalies 85Kr and 3H in near bottom layer of sea water by water infiltration in spent nuclear fuel in dumpsites of the ship and submarine reactors

    International Nuclear Information System (INIS)

    Soyfer, V.N.; Goryachev, V.A.; Andreev, D.V.; Barabanov, I.R.; Yanovich, E.A.

    2004-01-01

    Assessment of the future radioactive environmental contamination threat by releases from the dumped submarine and ship reactors with spent nuclear fuel (SNF) is a project goal. The methodology of search and detection of traces of noble gases ( 85 Kr, 222 Rn) and 3 H (which leaking the first from the SNF) on them plume on a bottom in dumpsites of ship nuclear reactors is advanced. Investigation of releases from the dumped ship reactors with SNF of icebreaker enin in Kara Sea is launched (September 2003). Research has multi-plane character: experiments with SNF, creation of techniques and expeditions. Preliminary observation of current field and hydrological characteristics in zone of marked dumpsites on the shelf has been fulfilled. For an estimation of a gas fission products release (including Kr-85) from the irradiated fuel during the corrosion in conditions a near-bottom layer of the sea in Arctic the special installation has been created in the RRC 'Kurchatov Institute' hot laboratory. Then the following experiment (modeling the conditions a near-bottom layer of the sea in Arctic) was made: First, samples of irradiated VVER-1000 type reactor fuel (UO 2 ) are loaded into capsules filled with pure sea water or sea water with the silt sampled in Kara Sea. The fuel samples have been selected due to similarity of its burnup level (62 MWd/kg) and other parameters to the one of icebreaker 'Lenin' reactor, that had been buried (1965) in Kara sea. The average sample activity was about 1 GBq for 137 Cs. Second, the capsules and the vacuum system are pumped out and checked on tightness by the helium leak tester. Then capsules are filled with the artificial gas mixture without traces of 85 Kr. Temperature of capsules are maintained in a range 0;1 deg C. Each 1-2 month gas samples are taken from the capsules into the previously pumped out flask, then capsules are refilled by the gas mixture. The measurements of 85 Kr concentration in the samples are carried out on low

  15. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1980-09-01

    Research activities in the Division of Reactor Engineering in fiscal 1979 are described. The work of the Division is closely related to development of multi-purpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committees on Reactor Physics and on Decomissioning of Nuclear Facilities. (author)

  16. 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

  17. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    1985-08-01

    Research and development activities in the Department of Reactor Engineering in fiscal 1984 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, safeguards technology, and activities of the Committee on Reactor Physics. (author)

  18. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Matsuura, Shojiro; Nakahara, Yasuaki; Takano, Hideki

    1982-09-01

    Research and development activities in the Division of Reactor Engineering in fiscal 1981 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)

  19. Reactor Fuel Isotopics and Code Validation for Nuclear Applications

    Energy Technology Data Exchange (ETDEWEB)

    Francis, Matthew W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Weber, Charles F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pigni, Marco T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-02-01

    Experimentally measured isotopic concentrations of well characterized spent nuclear fuel (SNF) samples have been collected and analyzed by previous researchers. These sets of experimental data have been used extensively to validate the accuracy of depletion code predictions for given sets of burnups, initial enrichments, and varying power histories for different reactor types. The purpose of this report is to present the diversity of data in a concise manner and summarize the current accuracy of depletion modeling. All calculations performed for this report were done using the Oak Ridge Isotope GENeration (ORIGEN) code, an internationally used irradiation and decay code solver within the SCALE comprehensive modeling and simulation code. The diversity of data given in this report includes key actinides, stable fission products, and radioactive fission products. In general, when using the current ENDF/B-VII.0 nuclear data libraries in SCALE, the major actinides are predicted to within 5% of the measured values. Large improvements were seen for several of the curium isotopes when using improved cross section data found in evaluated nuclear data file ENDF/B-VII.0 as compared to ENDF/B-V-based results. The impact of the flux spectrum on the plutonium isotope concentrations as a function of burnup was also shown. The general accuracy noted for the actinide samples for reactor types with burnups greater than 5,000 MWd/MTU was not observed for the low-burnup Hanford B samples. More work is needed in understanding these large discrepancies. The stable neodymium and samarium isotopes were predicted to within a few percent of the measured values. Large improvements were seen in prediction for a few of the samarium isotopes when using the ENDF/B-VII.0 libraries compared to results obtained with ENDF/B-V libraries. Very accurate predictions were obtained for 133Cs and 153Eu. However, the predicted values for the stable ruthenium and rhodium isotopes varied

  20. Power reactors operational diagnosis

    International Nuclear Information System (INIS)

    Dach, K.; Pecinka, L.

    1976-01-01

    The definition of reactor operational diagnostics is presented and the fundamental trends of research are determined. The possible sources of power reactor malfunctions, the methods of defect detection, the data evaluation and the analysis of the results are discussed in detail. In view of scarcity of a theoretical basis and of insufficient in-core instrumentation, operational diagnostics cannot be as yet incorporated in a computer-aided reactor control system. (author)

  1. Fast reactors worldwide

    International Nuclear Information System (INIS)

    Hall, R.S.; Vignon, D.

    1985-01-01

    The paper concerns the evolution of fast reactors over the past 30 years, and their present status. Fast reactor development in different countries is described, and the present position, with emphasis on cost reduction and collaboration, is examined. The French development of the fast breeder type reactor is reviewed, and includes: the acquisition of technical skills, the search for competitive costs and the spx2 project, and more advanced designs. Future prospects are also discussed. (U.K.)

  2. Pressurised water reactor operation

    International Nuclear Information System (INIS)

    Birnie, S.; Lamonby, J.K.

    1987-01-01

    The operation of a pressurized water reactor (PWR) is described with respect to the procedure for a unit start-up. The systems details and numerical data are for a four loop PWR station of the design proposed for Sizewell-'B', United Kingdom. A description is given of: the initial conditions, filling the reactor coolant system (RCS), heat-up and pressurisation of the RCS, secondary system preparations, reactor start-up, and reactivity control at power. (UK)

  3. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    Arien, B.

    1998-01-01

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of four main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents, the development of an expert system for the aid to diagnosis; the development and application of a probabilistic reactor dynamics method. Main achievements in 1999 are reported

  4. Seals in nuclear reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The aim of this invention is the provision of improved seals for reactor vessels in which fuel assemblies are located together with inlets and outlets for the circulation of a coolant. The object is to provide a seal arrangement for the rotatable plugs of nuclear reactor closure heads which has good sealing capacities over a wide gap during operation of the reactor but which also permits uninhibited rotation of the plugs for maintenance. (U.K.)

  5. Reactor BR2

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2000-07-01

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported.

  6. The Dragon reactor experiment

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    The concept on which the Dragon Reactor Experiment was based was evolved at the Atomic Energy Research Establishment at Harwell in 1956, and in February of that year a High Temperature Gas- cooled Reactor Project Group was set up to study the feasibility of a helium-cooled reactor with a graphite or beryllium moderator, and with the emphasis on the thorium fuel cycle [af

  7. The replacement research reactor

    International Nuclear Information System (INIS)

    Cameron, R.

    1999-01-01

    As a consequences of the government decision in September 1997. ANSTO established a replacement research reactor project to manage the procurement of the replacement reactor through the necessary approval, tendering and contract management stages This paper provides an update of the status of the project including the completion of the Environmental Impact Statement. Prequalification and Public Works Committee processes. The aims of the project, management organisation, reactor type and expected capabilities are also described

  8. TRIGA reactor characteristics

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

    This module describes the general design, characteristics and parameters of TRIGA reactors and fuels. It is recommended that most of this information should be incorporated into any reactor operator training program and, in many cases, the facility Safety Analysis Report. It is oriented to teach the basics of the physics and mechanical design of the TRIGA fuel as well as its unique operational characteristics and the differences between TRIGA fuels and others more traditional reactor fuels. (nevyjel)

  9. Integral fast reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1989-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics

  10. Reactor BR2

    International Nuclear Information System (INIS)

    Gubel, P.

    2000-01-01

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported

  11. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1988-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab

  12. Mirror fusion reactors

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    Conceptual design studies were made of fusion reactors based on the three current mirror-confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fuel for fission reactors. We have designed a large commercial hybrid and a small pilot-plant hybrid based on standard mirror confinement. Tandem mirror designs include a commercial 1000-MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single-cell pilot plant

  13. HOMOGENEOUS NUCLEAR POWER REACTOR

    Science.gov (United States)

    King, L.D.P.

    1959-09-01

    A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

  14. Australian research reactor studies

    International Nuclear Information System (INIS)

    McCulloch, D.B.

    1978-01-01

    The Australian AEC has two research reactors at the Lucas Heights Research Establishment, a 10 HW DIDO class materials testing reactor, HIFAR, and a smaller 100kW reactor MOATA, which was recently upgraded from 10kW power level. Because of the HIFAR being some 20 years old, major renewal and repair programmes are necessary to keep it operational. To enable meeting projected increases in demand for radioisotopes, plans for a new reactor to replace the HIFAR have been made and the design criteria are described in the paper. (author)

  15. Mirror fusion reactors

    International Nuclear Information System (INIS)

    Carlson, G.A.; Moir, R.W.

    1978-01-01

    We have carried out conceptual design studies of fusion reactors based on the three current mirror confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fission fuel for fission reactors. We have designed a large commercial hybrid based on standard mirror confinement, and also a small pilot plant hybrid. Tandem mirror designs include a commercial 1000 MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single cell pilot plant

  16. Reactor feedwater system

    International Nuclear Information System (INIS)

    Hikabe, Katsumi.

    1978-01-01

    Purpose: In order to prevent thermal stresses of a core of PWR type reactor, described has been a method for feeding heated recirculating water to the core in the case of the reactor start-up or shut-down. Constitution: A recirculating water is degassed, cleaned up and heated in the steam condensers, and then feeds the water to the reactor, characterized in that heaters are provided in the bypasses of the turbine, so that heated water is constantly supplied to the reactor. (Nakamura, S.)

  17. Remote Reactor Monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Bernstein, Adam [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dazeley, Steve [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dobie, Doug [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marleau, Peter [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Brennan, Jim [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gerling, Mark [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sumner, Matthew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sweany, Melinda [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-10-21

    The overall goal of the WATCHMAN project is to experimentally demonstrate the potential of water Cerenkov antineutrino detectors as a tool for remote monitoring of nuclear reactors. In particular, the project seeks to field a large prototype gadolinium-doped, water-based antineutrino detector to demonstrate sensitivity to a power reactor at ~10 kilometer standoff using a kiloton scale detector. The technology under development, when fully realized at large scale, could provide remote near-real-time information about reactor existence and operational status for small operating nuclear reactors out to distances of many hundreds of kilometers.

  18. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    Decreton, M.

    2002-01-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed

  19. Trench reactor: an overview

    International Nuclear Information System (INIS)

    Spinrad, B.I.; Rohach, A.F.; Razzaque, M.M.; Sankoorikal, J.T.; Schmidt, R.S.; Lofshult, J.; Ramin, T.; Sokmen, N.; Lin, L.C.

    1988-01-01

    Recent fast, sodium-cooled reactor designs reflect new conditions. In nuclear energy these conditions are (a) emphasis on maintainability and operability, (b) design for more transparent safety, and (c) a surplus of uranium and enrichment availability that eases concerns about light water reactor fueling costs. In utility practice the demand is for less capital exposure, short construction time, smaller new unit sizes, and low capital cost. The PRISM, SAFR, and integral fast reactor (IFR) concepts are responses to these conditions. Fast reactors will not soon be deployed commercially, so more radical designs can be considered. The trench reactor is the product of such thinking. Its concepts are intended as contributions to the literature, which may be picked up by one of the existing programs or used in a new experimental project. The trench reactor is a thin-slab, pool-type reactor operated at very low power density and- for sodium-modest temperature. The thin slab is repeated in the sodium tank and the reactor core. The low power density permits a longer than conventional core height and a large-diameter fuel pin. Control is by borated steel slabs that can be lowered between the core and lateral sodium reflector. Shutdown is by semaphore slabs that can be swung into place just outside the control slabs. The paper presents major characteristics of the trench reactor that have been changed since the last report

  20. New reactor concepts

    International Nuclear Information System (INIS)

    Meskens, G.; Govaerts, P.; Baugnet, J.-M.; Delbrassine, A.

    1998-11-01

    The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost